ML20215G630
| ML20215G630 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/16/1987 |
| From: | Bitter S, Burdick T, Damon D, Hare S, Lennartz J, Munro J, Reidinger T, Shepard D, Sunderland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215G626 | List: |
| References | |
| 50-454-OL-87-01, 50-454-OL-87-1, NUDOCS 8706230325 | |
| Download: ML20215G630 (122) | |
Text
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U.S. NUCLEAR. REGULATORY COMMISSION REGION III ,
- Rep' ort. No. 50-454/0L-87 Docket No. 50-454/455 ~ License No. NPF-37
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Licensee: Commonwealth Edison '
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Byron Nuclear Station j 4450 N. German Church Rd.' a 1
Byron, IL :61010 j l
Facility Name: Byron Nuclear Station j
. Examination' Administered At: Byron i
Examination Conducted: Week of May 4 and 11, 1987
' \x"' 0 w aC hg, 6//c/p, Examiners: in .hy D. Reiding(;
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'h. h h Approved By:' Thomas M. Burdick, Chief C/M [E Operating Licenses Section Date' Examination Summary Examination administered on weeks of May 4 and 11, 1987 (Report No. 50-454/0L-87-01)) _
Areas Inspected: Written and/or operating exams were administered to four reactor. operators.and nine senior reactor operators.
.Results: Four reactor operators and seven senior reactor operators passed
,: Ithe examinations, y
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DETAILS l l
- 1. Examiners j
T. Reidinger, NRC J. Lennartz, NRC D. Damon, NRC l
i D. Shepard, NRC P. Sunderland, NRC S. Hare, NRC S. Bitter, NRC J. Munro, NRC
- 2. Examiner Observations During the administration of the operating exams, the examiners observed both generic deficiencies and strengths on the part of the Senior Reactor Operator and Reactor Operator candidates. In addition, one concern was identified by the examiners regarding the maintenance of the key log when ,
radiation area keys are logged out. During the simulator portion of the operating exam, the following deficiency in the candidates' performance was observed:
- The Reactor Operator and Balance of Plant Operator were not made aware of CAUTIONS and NOTES in the Emergency Procedures as the Senior Reactor Operator did not verbalize them during the execution of the Emergency Procedures. This resulted in inattentiveness being displayed towards critical plant parameters and specific plant equipment.
During the simulator portion of the operating exam, the following generic strengths in the candidates' performance were observed:
- The candidates demonstrated good crew interaction and communicated well between each other.
- The Senior Reactor Operator candidates demonstrated good supervisory and leadership skills.
- The candidates conducted themselves with a high sense of professionalism while executing their assigned duties.
One general concern over the licensee's maintenance of the key log was identified by the examiners. In numerous instances when an "R" key (radiation key) was logged out, the log was not properly completed in accordance with Byron Administrative Procedure No. 900-6. It was recommended that better control be taken when filling out the key log, and that more frequent reviews of the log be done to ensure that it is being filled out properly.
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, !' N 37 ExitfNeeting
- f~ An'exitimeeting was held following the examinations to discuss the c' (aforementioned observed weaknesses, strengths, and concerns on~May,12, 1987, wit;h facility management and training staff representatives.
.NRC Representatives in, attendance werei 1 J. Lennartz, Examiner ? '
D. Shepard, Examiner S. Hare, Examiner '/ '
P.:Sundgrland, Examiner?3 G. Nejfel,1, Examiner- ,
R. Eaton, NRR-m, c' Facility' Representatives in attendance were:
$ R. Querio,' Station Manager p' R. Pleniewicz, Production Superintendent f A. Chernick; Training Supervisor b A. Snow, Assistant Training Supervisor a J. Printz; Training Center Instructor l The' facility'msnagement acknowledged the examiners observations discussed inSection2ot(thisreport.
- 4. Simulator Performance
/ During.the examinations, the,fq110 wing proolems with the simulator Jl/ occurred: y
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>b UnabletorunIC-29(10Epower,PeakXenon. Burnout,E0L).
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' f. b. Delta Flux program of',' process computer did not work when one Power
. Range Instrument was out of service,
- c. Steam Dumps failed to close one time when interlock (P-12) was met.
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- d. Off Gas pump had dual indication when in stop.
- e. Condenser. emergency overflow valve open annunciator came in at 30" indicated. The BAR setpoint for this is 38". '
, l 7 f. Had to. leave the H2 co'oler wate'r valve, 1WS-106, controller in j manual. . Auto on" controller would not control temperature in the green band. j N g. IC-14 (50% Power, Eq. Xenon, 790 ppm, 580 MW, BWGP 100-3 Step 54) had. {
to be re-initialized to get stable readings on varicys meters. '
- h. IC-15 (50% Power, Xenon changing after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 12-3-6-3 load cycle) locked pp the machine.'
- i. Simulatcr instrumentation drifts while in freeze, and did not return to tns appropriate values when return to run.
- j. Red on' containment mon!/crs for the RM-11 do not give a containment ldnt$lation isolatim annuncictor. "
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- k. Man anation TUR-13 (Tovbine trip failure) did not work. l
- 1. Simulator re-initialized by itself. I
- m. N 1 functions CRF-10 (DRPI-Loss of Vcitage), and EPS-1 (Loss of I i W/f-sitR power) 'p icked up the simulator.
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- n. Malfunction EPS-1 (Loss of off-site power) did not operate as described in the malfunction book. (There was no time delay.)
These problems were discussed with Production Training Center personnel j at a meeting held on May 13, 1987.
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ATTACHMENT
'At the conclusion'of the' written examinations the facility is given a copy of the examination and the answer key for both the Senior. Reactor Operator and
-the Reactor Operator examinations. .The facility then.has five working days in which to provide written comments concerning the examinations'to the NRC..
The following paragraphs contain the. facility comments concerning the examinations.followed by the NRC' response.
REACTOR OPERATOR EXAMINATION COMMENTS AND RESOLUTIONS Question'1,14 (Part C)-
Byron Response: False. A: leak through a SG PORV would be reflected in the calculation of the calorimetric and in addition, this change would be reflected by a' change in the power range instruments. Reactor' power-is dependent on steam demand.
An actual change in steam demand will affect both the calculation and indication equally; therefore, the answer would be ' false.'
'NRC Response: Concur. The answer key has been modified to reflect the correct answer.
Question 1.19 Byron Response: Water hammer caused by water entrained in steam lines can also be eliminated (prevented) by the following:
(1) Drain water from the steam lines prior to heatup.
(Reference 1BGP 100-1, Revision 52, Section F Main Body, Note and Step 7 on Page 7.)
(2) Admit steam slowly. (Reference 1BGP 100-1, Revision 52, Section D Precautions, Step 7 on Page 4.)
At Byron, Main Steam line Low Point Drain Pots / Drip Legs are used in place of steam traps; therefore, this should also be an acceptable answer. (Reference 1BGP 100-1, Revision 52, Section F Main Body, note on Page 7.)
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. 'NRC Response: Concur. The answers originally provided came directly-from Commonwealth Edison s PWR FUNDAMENTALS, Fluid Flow-(Chapter 1, Page 49). However, there are certainly 1
additional acceptable answers.' " Admitting steam slowly" and "using drain pots" are examples. The answer key has been' modified to reflect these alternate answers.
Question 2.01?
Byron Response: Chapter 57, Braidwood System Description Lesson Plan TP0 No.:2, Page 1, lists the five Fire Protection Subsystem flowpaths as:
(1) Sprinkler system (2) Deluge system l (3) Foam. System (4) Cardox (C0 )' system (5) Halon.syst$m Byron proposes that the above responses also be accepted as correct answers.
NRC Response: Do not' concur. The objectives furnished to the NRC do'not match those provided in the. form of ATTACHMENT 3 to the Byron comments. The NRC was furnished with objectives for Chapter 57 that are dated April 1986, Revision 2.
Attachment 3 referred to by BYRON in the Byron comments contains objectives that are dated August 19, 1985, Revision 1.
Because the version originally supplied to the NRC is the latest version, it will be used. No points will be awarded for answers that do not answer the question as stated. The facility is urged to ensure that the students,
~ instructors, and the training text all have objectives that are consistent.
Question 2.02b.
Byron Response: The only remaining interlock to start the RCP is RCP Oil lift. pump discharge pressure greater than 600 psig.
6E-1-4030'RC01'and RC05 indicate that only the 600 psig (
interlock need be met. Byron proposes that RCP oil lift >
pressure of 600 psig be sufficient for full credit on this .
portion of Question 2.02. (
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NRC Response: Concur. However, on Page 13-30 of the Braidwood System Description, there is listed an additional interlock:
(c) RCP oil lift pump running." The logic diagram in the same tert does not match the interlock explanation. The facility is urged to ensure that the body of the text is consistent with the logic diagrams prior to the next scheduled examination. The answer key has been modified as suggested by the facility.
Question 2.03 Byron Response: The diesel-driven Aux Feed Pump engine heat exchanger should be added to the above list. (Jacket Water Heat Exchanger).
NRC Response: Concur. The answer key has been modified to reflect this. l Question 2.06 ;
Byron Response: The question could be interpreted to mean "On a loss of all AC, what does 250 VDC supply?"
The main condenser vacuum breaker valve is still supplied by 250 VDC, even if all AC is lost, and this should not be considered as an incorrect answer.
NRC Response: Concur. Although answer key has been changed to accept above answer no change to candidate exams were made since pass / fail results are unaffected.
Question 2.07 Byron Response: There are several other responses considered correct, in addition to the above list. These are:
- 1. Accidental opening of transfer canal gate would lower {
water level only 6 feet. !
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- 2. There are no drains installed in the pool.
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Also, for Part 3), answer key, the parenthesis are around the wrong part. The acceptable answer should be: "The spent fuel liner has continuous weld seam drains."
NRC Response: Concur. The confusion stems from the text on Page 51-33 of the Braidwood System Description. This portion of the ,
system description states what appears to be "the three l ways in which the spent fuel pool is protected. . . ." '
However, there are more than three possible answers.
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15, and 29 of the Braidwood System Description. It should.
be.noted further, that.no Byron License Lesson Plans were--
provided as part.of the examination preparation material.
~In summary, the answer key has been modified to accept the p alternate answers as suggested _by the facility.
Question 2!08:
L ' Byron' Response: The NRC; answer is correct, but'some of the candidates may have listed.the key lock valve as 0WX-353 or Release Tank outlet. isolation valve. Byron. proposes that these responses also be considered correct.
NRC Response: Concur. Each of the postulated answers.is certainly-equivalent to that provided in the answer key. This will be indicated on the answer key.
Question 2.09
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' Byron ~ Response: The first three parts of the NRC answer are legitimate. j However, the question asks for interlocks / actions to manually open the valve which implies the control switch istakentoogen. Byron proposes that Part 4,'"MCB switch taken to open not be a required response.for full credit.
NRC Response: Concur. The word " actions" that immediately followed the slash mark was intended to elicit the response "MCB switch taken to open." However, this appears to be too subtle.
Therefore, "MCB switch taken to open" will be deleted as a required response. .The answer key has been modified to-reflect this.
' Question'2.11 Byron Response: The response in Part b) "if low steamline pressure SI is not blocked" should not be required for full credit on b).
The proper response in Part c) adequately demonstrates that the examinee recognizes that the low steamline pressure SI protective function must be unblocked.
NRC' Response: Do not concur. It is not obvious.that by answering Part c) correctly, the examinee recognizes the existence of the interlock dealing with low steamline pressure. The interlock must be stated explicitly to receive full i credi t'. No change to the answer key is warranted. I 4
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- Question 2.13 Byron-Response:. The NRC answer is correct. However Part 5 may be listed by some candidates.as individual switches.. Byron. proposes
'that the following individual switches which wil1~cause a SSPS General Warning also.be considered as correct :
answers:
. Input Error Inhibit Switch Logic A Switch Permissive _. Switch 1 Memories Switch' ,
Mode Selector Switch- ~l NRC Response: '
Concur. The switch positions derived from BAR 1-4-83,
-Revision 51 were not listed originally because they were not: listed in the system description. However, listing.
any of these. switches ~in the. test position as a' response
'is certainly correct. The answer key has been modified to
. reflect these alternate answers.
Question 2.14 Byron Response: The following are also powered or capable of being powered from bus 142, and should be added to the above list:
4150V/480 transformer to bus 132Z Bus-144 Bus 242 NRC Response: Concur.= The original answer was taken directly from Page 43 of Chapter 4 (not 42 as. indicated on the answer key) of the Braidwood System Description. The facility is urged to update the training material prior to the next scheduled examination. The answer key has been modified to reflect the alternate answers as suggested by the facility.
Question 2.15 Byron Response: The reference used by the NRC is incorrect. A similar question was asked on the July 16,.1986 NRC Exam and Examination Report No. 50-454/0L-86-01 references SPDS Revision BY3-0, March 25, 1986 from the CECO Computer Group. This revision indicates that the five inputs should be:
- 1. Pressurizer Pressure
- 2. Wide Range RCS Pressure
- 3. Containment Pressure
- 4. Containment Hi Range rad monitors
- 5. Incore Thermocouples !
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Additionally, the term "be specific" doesn't provide adequate guidance to the examinee. The requirement for channel' numbers or how many channels is not apparent in the question, so the above list of five answers should be considered adequate.
NRC Response: Concur. The: reference'used by the NRC is incorrect i because.the reference supplied to the NRC is incorrect.
Thefacilityisurgedtoupdatethetrainingmaterialina' more timely. fashion. Futhermore, the term . specific" was .
used'to prevent the listing of vague responses such as
" pressurizer pressure." However, this 1s now a moot point because the new answers.-(as. suggested by.the.
facility) are not specific due to the complex nature of the subcooling margin monitor. The. answer key has been
, modified to reflect the correct inputs as suggested by the .
facility.
Question 2.19 Byron Response: System descriptions Chapters 42 and 20 describe the response of RCFC's and support equipment. Byron proposes the following be considered as correct responses:
- 1. Operating RCFC's shift to slow speed.
- 2. Non-Operating RCFC's start in slow speed.
- 3. SX is bypassed to containment chillers resulting in chiller trip.
NRC Response:- Concur. The answer key has been modified to reflect this. :
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. Question 2.20 Byron Response: Byron Station Nuclear' Station Operators (NS0's) are not involved in performing fuel movement. Persons with SR0 licenses are required to supervise fuel movement. Byron proposes that this question be removed from the exam.
NRC Response: Do'not concur. Terminal Performance Objective No. 8 as provided states: Identify by sight each of the following fuel handling tools. No change to the answer key is warranted.
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Question 3.05 Byron Response: . Technical Specifications Bases 3/4.1.3 for, Movable Control Assemblies'on Page B3/4:1-3 also lists three purposes:
(1) Acceptable power. distribution limits are maintained, (2) The minimum Shutdown-Margin is maintained, and (3) The potential-effects of rod misalignment on associated accident analyses are. limited.
NRC Response:' Concur. The three purposes listed in the Technical Specification Bases are equivalent to those given in the answer key. The answer key has been modified to reflect this.
Question 3.08' Byron Response: See attachment from Braidwood RC System
Description:
.d.2 "a" &'"b" are all things of which TAVE is an input.
.0nly 2 of these are actually' protection signals (OT Delta T and OP Delta T). The answer key should accept all of these as correct answers or only require 2 answers.
NRC Response: Do not concur. It is true that Tavg is an input to all of the items referred to as d.2 "a" and "b" of ATTACHMENT 14 to the Byron comments. However, all four items of the answer provided by the NRC are protection (as opposed to control) signals. No change to the question or to the answer key is warranted.
Question 3.09 Byron Response: According to BAR 1-10-A7, there are eight conditions that would cause this alarm:
- 1. 2% radial flux tilt
- 2. Any rod > 12 steps from any other rod in its bank
- 3. Any S/D rod <225 steps after S/D Bank E >225 steps
- 4. Any bank >228 steps
- 5. C/BD >225 steps
- 6. Rod sequence or overlap error >l step
- 7. 25% Axial Flux Tilt
- 8. DRPI vs Demand position > 12 steps i The question only asks for five conditions; therefore, any j five from the above list should be acceptable.
NRC Response: Concur. However, the facility is urged to ensure that the System Description is consistent with the current procedures prior to the next scheduled examination. The answer key has been modified to accept the alternate answers.
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t' Question 3.10 Byron Response: No. 3 should accept either oscillator. failure or pulser-failure as acceptable. Braidwood-System Description drawings refer to it as." Pulser-& Oscillator" or
" Pulser." This is just a nomenclature difference.
NRC Response: Concur. The answer key has been modified to reflect this.
Question 3.14c:
Byron Response: Braidwood-System Description says. light is ON.
.NRC Response: Concur. The answer; key has been modified to reflect this. l
-Question 4.03b Byron Response: Question asks when the limit can be exceeded. It does not-ask'the candidate to specify what the limit can be raised-to. Therefore, an acceptable answer should be: "The limit can be exceeded provided that . . . ."
NRC' Response: . Concur. The answer key has been modified to reflect the correct. answer for the question'as wr,itten, a
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SENIOR REACTOR OPERATOR EXAMINATION COMMENTS AND RESOLUTIONS Question 5.01b.
Byron Response: Decrease Tyincreases as power increases should be an acceptable l answer. 4 NRC Response: Concur. Either answer is acceptable. No answer key modification is required, as the answers are equivalent. [
Question 5.12a.
Byron Response: It becomes more negative . . .
The terms "more negative" and "less positive" are used interchangeably. Therefore, "more negative" is also an I acceptable answer. !
NRC Response: Concur. Either answer will be accepted. No answer key modification is required, as the answers are equivalent.
Question 6.02 Byron Response: The NRC answer i., correct according to Byron Units 1 and 2 Differences, Revision 4/1/86. However, Figure 9-1 shows the Hi-Hi alarm setpoint to be 78%. In the past, Byron Training has accepted setpoints rounded to the nearest whole number. Byron proposes the following be the acceptable answers for Question 6.02:
i Unit 1 - 81.4% (81)
Unit 2 - 78.1% (78)
NRC Response: The setpoint rounded to the nearest whole number will be acceptable. The answer key has been modified.
Question 6.11 Byron Response: Part A asks only for eight indications, not the specific channel numbers indicated in the answer key. Additionally, S/G 1A and 1B WR level and pressure are individually !
accepted. However, RCS WR Hot leg and Cold leg temperatures are grouped in a general fashion. Byron ;
proposes that any eight of the following list be accepted as correct answers for Part A:
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-4. RCS WR Hot leg Temperature Loop D
- 9. Pressurizer Pressure
- 10. Pressurizer Level
- 11. S/G 1A Pressure
- 12. S/G 1A WR level
- 13. S/G 1B Pressure ,
- 14. S/G 1B WR level
- 15. Source Range
- 16. Power Range and for Part B.,
- 1. Source Range
- 2. Power Range The power range instrument is labeled on the Fire Hazards-control panel on Unit 1 and OPM02J as intermediate range.
Listing the Power Range instrument as intermediate range should be considered a correct answer, because the readout extends into both intermediate and power ranges.
NRC Response: Part A: Partially concur. Specific instrument numbers of channel numbers are not required for full credit.
However, RCS WR Hot Leg Temperature and RCS WR Cold Leg Temperature will each count as a single response. The answer key has been modified. '
Part B: Concur. Either answer will be accepted. The answer key has been modified.
Question 6.12 Byron Response: The function of the light on 0GW014 does not necessarily mean that a gas discharge is in progress. The following '
conditions must be met to energize the white light-
- 1. 0GW014 control switch in "open" position
- 2. High Rad Signal not present
- 3. Valve Controller not set to 0 (Demand from Controller)
NRC Response: Concur. The originally proposed NRC answer or Byron Responses 1 and 3 together with response No. 2 will be acceptable. No answer key modification is required. j i
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Question 6.13 Byron Response: NRC Answers 1, 2, and 3 are incorrect. For Answer 1, there are not liquid release related monitors on the RM-23. For Answer 2, there is no local alarm on 0GW02J associated with a liquid release. For Answer 3, the Aux.
Bldg. Vent Stacks are not used for liquid release.
Answer 4 should accept either 0WX896 or 0WX353 valve position indication (both are listed in B0A RAD-1 to close.)
Other acceptable answers to this question could be:
- a. " Liquid release in progress" placard placed when release ~is initiated. (BCP 400 TSO, p. 10)
- b. Loss of flow on Release Tank Flow recorder (System Description Table 48a-2)
- c. Loss of flow on Release Tank Flow integrator (System Description Table 48a-2)
NRC Response: Partially Concur: NRC Answers 1 and 3 have been deleted, and Number 2 was modified to include 2nly the local alarm i at the Liquid Radwaste Control Panel. NRC answer Number 4 was modified"to include both OWX 353 and 0WX 896 valve position indication. Loss of Flow on the Release Tank Flow Recorder or the Release Tank Flow Integrator will also be accepted as a single correct response Placing the " Liquid release in progress" placard on the control board is an administrative procedure requirement and not a positive indication of a release or that the setpoints have been exceeded, and therefore, will not be accepted.
The answer key has been modified. I 1
Question 6.14 Byron Response: The question solicits a description of the detection of an electrical fault in the stator. A fault in the generator stator winding therefore is implied, and should not be required as a part of the answer. The question also doesn't ask how the diesel will respond to the fault. l Therefore, shutdown of the diesel should not be required.
"The fault will cause a high current flow through the !
neutral ground transformer (neutral ground resistor) which will actuate the neutral ground protection relay" should be accepted. ;
NRC Response: Concur. The answer key has been modified.
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' Question 6.15 Byron Response: The following Technical Specification monitors are located on the RM-23:
- 1. Wide Range Gas Monitor (WRGM) (PR030's)
T.S. 3/4'3.3.3.6
- 2. Containment High Range (AR011,~12) T.S. 3/4 3.3.3.6
- 3. Main Steam Line (AR022's, 23's) T.S. 3/4 3.3.3.6
'4. Containment Fuel Handling Incident.(AR011,12)
- 5. ' Fuel Building Fuel Handling Incident (AR055, 56)
- 6. Control Room Outside Air Intake (PR031, 32, 33, 34)
NRC Response: Concur. The question was written from the reference material supplied from the utility with revision date of 10/85. The answer key has been-modified to reflect the material with revision date of 11/86. The utility is urged to supply the most recent material to the NRC for '
future. examination preparation.
Question 6.16 Byron Response: Tref is the actual input to the meter. PT505 or 506 could '
supply this input dependent upon which is selected. Byron proposes the following answer be acceptable: 1 Tave/ Tref chart recorder - Auct Hi Tave and Tref (Turbine Impulse Pressure) (PT 505) !
Tref /Tave meter - Auct Lo Tave and Tref (Turbine Impulse Pressure) (PT 505)
NRC Response: Partially concur. Tref will be accepted provided there is an explanation of where Tref is generated from. Tref ;
alone will not be accepted. The answer key has been modified.
Question 8.06 Byron Response: Per BAP 110-13, Page 2, Steps 5-9, the following. answers )
should be acceptable: j 1
Functional checks are acceptable only if: i (1) They ensure that the entire portion of the system which was affected by previous actions is tested ,
(2) They won't compromise plant safety (3) Indication utilized is neither null nor zero
-(4) The reading is directly observed and recorded by the observer I
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Status monitoring is acceptable only if:
(1) It's accomplished by observation of control room lights, ESD panel indications, valve position indicators, etc.
(2) Indication utilized is neither null nor zero (3) The reading is directly observed and recorded by the observer NRC Response: Do not concur. The question specifically asks for conditions in which functional checks and status monitoring would be acceptable, and not for requirements to perform or methods of accomplishing them. No answer key modification is required.
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U. S.: NUCLEAR-REGULATORY COMMISSION l s, . REACTOR OPERATOR LICENSE EXAMINATION FACILITY - _gyBQN_1_________________ ')
REACTOR. TYPE: _PWB-WEQ$____;___________
a .
DATE-ADMINISTERED: _gZ/92/93________________.
EXAMINER: _g1IIESig________________- )
CANDIDATE: _________________________
INDIBWGI19NS_I9.98ND19@151 Use separate paper-for the-answers. Write answers on one side only.
- Stcple. question sheet Hon top of the answer sheets. ' Points for each qusstion ~are indicated in parentheses af ter the question. The passing .
gecdeirequires at least 70% in each category . and a final grade of at-
,:12 cst.80%.L Examination papers.will be picked (4) six (6) hours after:
'th2 examinationLstarts.
l
- 7. OF
, CATEGORY % OF CANDIDATE'S. CATEGORY'
__Y869E_ _I9106 ___EGQBE___ _Y8LUE__ ______________G81EQQBY_____________
r
_2Es99 _ _2Et11 ___________ ________ 1. PRINCIPLES:OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HCAT TRANSFER AND FLUID FLOW
_29s99__ _2kA21 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS .
_2kzE9__ _2}z22' ___________ ________'3. INSTRUMENTS AND CONTROLS
________ 4. PROCEDURES - NORMAL, ABNORMAL,
_22iO9__ _23 22 ___________
ENERGENCY AND RADIOLOGICAL CONTROL 111sE9__ ___________ __._____% Totals Final Grade AII work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
e a
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS s
During the administration of this examination the following rules apply:
1.' Cheating on the examination means an automatic denial of your application cnd 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.
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- 3. .Use black ink or dark penci l gely to f acilitate 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.
- 7. Print your name in the upper right-hand corner of the first page of ggch cection of the answer sheet.
B. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ogw page, write gely 90 gng gidg of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and n' umber, for example, 1.4, 6.3.
- 10. Skip at least th t gg lines between each answer. -
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in f acility liigtetutg.
- 1
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide f or .the depth of answer required.
- 14. Show al1~ calculations, met h od s', 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 DUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the g>;gmi ott onl y.
- 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.
W
\
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- l. - you shall LEl. WhsnLyou complete' pour examination,
^
c.
Assemble your examination as f ollows:
E, (1) Exam questions on top. .
Exam aids . figures, tables, etc.
(2) ,
(3)
Answer pages including. figures which are part of the answer. !
Turn in-your copy of the examination and all pages used to answer
='
6.-
the examination questions.
the paper that you did' gg; scrap paper and the balance Turn in' all answering the questions, of c.-
not use f or as defined by the examiner. Ifisafter still d.
~leaving, Leave the examination area,you are foundorinrevoked.
denied this area while the examination .jlj -
-in progress, your, license may be S 0
f fl '
k
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0 1
a PAGE 2 l
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I t;
DUESTION 1.01 (1.00)
A reactor is shutdown with a boron concentration of 1100 ppm and a SDM of 2 percent delta K per K. The source range channel indicates 10 cps. The boron concentration is then Choose reduced by 100 ppm.- Assume that DBW is -10 pcm/ ppm.
Keff.
l the answer that is closest to the final l
l o) Keff = .980 i b)'Keff = .985 c) Keff.= .990 d) Keff-= .995 i
{'
i DUESTION 1.02 (1.00)
= 1.00000) at Assume that the reactor If + 5 pcm 'is added to the core, choose the is critical (Keff 10exp-8 amps.
ONE statement from those below that BEST describes reactor behavior.
a.) Neutron population and core power will continue to increase
' indefinitely until the operator inserts -5 pcm by inserting .
control rods or boron.
will increase by an amount ,
b.) Neutron population and core power equal to 0.005 percent.
This will add -
c.) Neutron population and core power will increase. The power defect will negative reactivity due Thetocore power defect.
power wi11' level out by itself at a cancel.the +5 pens. i slightly higher power.
Neutron population and core power will increase. This w211 add d.) The power defect negative reactivity due to power defect.
will cancel the +5 pcm. The core power will return to the original power.
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1.03 (1.00)
QUESTION .
During a reactor startup, just. prior to reaching respond to a ;
criticality,"the'SUR meter indication will(choose the best answer) i given amount of rod withdrawal by: !
Rising slowly, then slowly fal. ling off_to zero B
a.)
b.) Rising rapidly, then slowly 'f alling of f to zero I
c.) Rising slowly, then rapidly falling off to zero d.) Rising ' rapidly, then rapidly f alling of f to zero ~
t QUESTION 1.04 , (1.00)
The A reactor has been at 100 percent power for 6 months.
decay heat generation rate immediately f ollowing the reactor
). Fill in the blank trip is approximately ( l using the best answer f rom those listed below. . 4 1
a) 0.7 percent rated thermal power -
b) 1 percent rated thermal power c) 6 percent rated. thermal power d) 14 percent rated thermal power e
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QUESTION 1.05 (1.00)
,LA reactor S/U was commenced five-(5) hours after a trip from full power equilibrium. A O.5 percent / min. ramp to full s
power was used. How would the resulting xenon transient vary, if-instead, a 2 percent / min. ramp were used?
a.) The minimum point in the xeno'n dip-for the 2 y percent / min. ramp would. occur' sooner and the magnitude.of the. dip would be smaller.
b.)- The minimum point in the xenon dip f or the 2 percent / min. ramp-would occur later and the magnitude of the dip would be smaller. ,
c.) The minimum point in. the xenon; dip f or1the 2 percent / min. ramp would occur' sooner and,the magnitude of 'the; dip would be larger.
J d.) The minimum point in' the- xenon dip f or the 2 j j
percent / min. ramp would occur later and the magnitude.of the dip would be larger. .l I
t DUESTION 1.06 (1.00)
A system containing saturated vapor at 585 psig is leaking via a' crack in a pipe wall. The pressure downstream of the crack is 335 psig. What is the condition of the fluid downstream of the crack? Choose the most correct answer.
a.) Wet vapor with a very high quality (quali ty i s near .
100 percent) -
1 b.) Wet vapor with a quality of approximately 60 percent.
c.) Superheated vapor with the amount of superheat less than 2 degrees F.
d.) Superheated vapor with the amount of superheat greater.than 2 degrees F.
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E, 1.07 (1.00)
DUESTION Temperature is 500 degrees F. What-Pressure is 1985 psig.
is the amount of subcooling? (Choose the closest answer) i a.) 36 degrees F b,) 136 degrees F c.) 137 degrees F d.) 170 degrees F IE-DUESTION 1.08 (1.00) b Each has a flow rate "m".
The All four RCPs are operating.
combined flow rate (corechoose flow) is "M". From the f our
.possible answers below, the ONE ANSWER ~THAT BEST FITS if one RCP is secured. ,
f-o) b) The The core flow (M) core flow (M) and will increase.
the flow of each of the operating RCPs (m) will increase. decrease and the flow of each of c) The core flow RCPs the operat2ng (M) will(m) will increase.
will not change because the remain 2ng !
d) The core flow (M)
RCPs will increase their indsvadual flows (m).
1.09 (1.00)
QUESTION Which DNE of the following will cause thenegative)?
(less Axial Flux Difference to become more positive a.) A power increase with' power defect compensated for by dilution only. l; A power increase with power defect compensated for !
b.)
by rod withdrawal only.
the core, f The buildup of xenon in the top portion of i c.) ;
xenon in the bottom portion of the j d.) The burnup of core.
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j 4
e QUESTION 1.10 (1.50)
Indicate whether EACH of the f ollowing INCREASES, DECREASES, or DOES NOT. AFFECT the available NPSH of a centrifugal pumps Consider each statement independently of'the others.
a) The pump discharge valve is throttled shut.
b) The temperature of the pump suction. fluid decreases.
.c) The nitrogen' blanket pressure on the suction side supply tank increases.
QUESTION 1.11 (2.00)
An estimated critical position has been calculated for a reactor startup that is to be performed 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip from a 100 day f ull power run. For EACH of the-f ollowing events or conditions, state whether ACTUAL critical rod position is GREATER THAN, LOWER THAN, or DOES NOT DIFFER FROM the ESTIMATED critical, rod position.
a) -Steam dump pressare setpnint is decreased by 100 psig. .
b) Startup is delayed for appronimately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
c) The present boron concentration is 40 ppm lower than -
that used in the ECP calculation.
d) Condenser vacuum decreases by 2 inches Hg. ,
OUESTION 1.12 (2.50)
Indicate how EACH of the following changes in plant condition would effect DNBR. Consider each case Use the terms INCREASES, independently of the others.
DECREASES, or NO EFFECT. No explanation is required.
a).Tavg decreases b) Rods are withdrawn without changing turbine load c) Reactor coolant flow increases d) Pressuri er pressure increases e) Reactor power increases
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.PAGE 7 iz__EBING1ELEE_QE_N9GLE8B_EQWEB_E'L8NI_QEEBeI1QNa
.4 .IBEBdQDXU8d1GEi_bE8I_IB9NEEEB 0ND ELWID_ELQW QUESTION 1.13 (2.00)
Assuming that steam is undergoing an IDEAL throttling process, indicate whether each of the following thermodynamic parameters will INCREASE, DECREASE, or REMAIN THE SAME.
a.) Enthalpy b.) Pressure c.) Entropy d.) Temperature DUESTION 1.14 (1.50)
Assume that the power range channels have.been adjusted based on a calculated calorimetric. Answer EACH of the-f ollowing three statements, TRUE or FALSE.
c) .If the blowdown flow had been ignored in calculating the calorimetric, then actual reactor power would be higher than indicated reactor power. ,
b) If the f eedwater temperature used in calculating the -
calorimetric had been 10 degrees lower than actual feedwater temperature, then actual reactor power would be higher than indicated reactor power.
c)'If a main steam atmospheric relief valve had been leaking.
during the data-taking portion of the calorimetric, then actual reactor power would be higher than indicated reactor-power.
DUESTION 1.15 (1.50)
With respect to reactor thermal limits, indicate whether EACH of the f ollowing statements is TRUE or FALSE.
a) The enthalpy rise hot channel factor limit prevents DNBR f rom decreasing below 1.3.
b) The purpose of the limit on the heat flux hot channel factor is to insure that fuel clad temperature does not exceed 2200 deg F during NORMAL operations.
c) Nucleate boiling is not permitted at anytime in the core.
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OUESTION 1.16 (1.50)
Indicate whether EACH of -the f ollowing statements is TRUE or
. FALSE.
c) DBW becomes more negative as moderator temperature increases. l J
b) MTC becomes less negative with increasing boron concentration.
c) During an increase in power from equilibrium xenon conditions, xenon concentration initially decreases.
QUESTION 1.17 (2.00)
Attachment B of ES-0.1, Reactor Trip Response, states five conditions that support or indicate natural circulation flow. One'of these is: "RCS subcooling-acceptable value."
State the other four conditions. ( exact wording is not required) e QUESTION 1.18 (1.00)
A sufficiently low temperature is one general requirement *
'for brittle fracture. State the OTHER TWO general requirements for brittle fracture.
QUESTION 1.19 (1.00)
Water hammer can be caused by having water entrained in steam lines. One method of eliminating THIS TYPE OF WATER HAMMER is to bleed the steem lines. State TWO OTHER methods used to eliminate (prevent) THIS TYPE OF WATER HAMMER.
QUESTION 1.20 (1.00)
Etate 2 reasons why 10exp-8 emps i s chosen as a standard reference for obtaining critical rod height date.
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l DUESTION. 1.21 (1.50)
I' Indicate the most significant: type of heat transfer that is taking place in each of the f ollowing conditions. (i . e.
Conduction, Convection or Radiation)
L l' a) Nucleate boiling on the cladding surface of the f uel assembly.
b) Accident. condition in which high quality' steam is passing through the coolant channel s.
1 l c) Heat f r om fission across a f uel pellet.
\
t 4 'b .
e
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'OUESTION 2.01 (1.00)
List the FIVE subsystems of the Plant Fire Protection .
System.
QUESTION 2.02 (2.00)
In order to'close an RCP breaker, several interlocks must be
- met. One of these is: the loop isolation valve interlock must be satisfied.
a) . State preci sel y all conditions / interlocks, etc. that must be met to satisf y the loop isolation valve interlock ,
referred to above.
b) State the remaining two interlocks that must be satisfied to close the RCP breaker. Include setpoints.
4
. QUESTION 2.03 (2.00)
List FOUR Essential Service Water loads that are not cubicle coolers or oil coolers .Do not use backup supplies / emergency _
sources of water as answers. (Duplicate credit will not be given f or listing the same component twice merely with a di f f er ent tr ain. )
QUESTION 2.04 (1.00)
List the f our core bypass flows.
QUESTION 2.05 (1.50)
List five automatic actions that occur if a boric acid flow deviation condition exi sts f or greater than 30 seconds.
QUE3 TION 2.06 (1.00)
Lis loads supplied by the 250 VDC Distribution System
$2"_v- WHEN ALL SOURCES OF AC ARE LOST.
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1
$2__ELeBI_DESIEN_INGLWDINE_E9EEIY_BND_EbEBEENGY_EYEIEdf PAGE- 11 QUESTION '2 07 (1.50)
. gchole7 ~
List bbe 3' ways in which the spent f uel pool is protected.
from a complete loss of. water.
QUESTION 2.08 (2.00).
1).Name the ONE component in the liquid radwaste system
'that prevents the inadvertent discharge of potentially radioactive water to the river.
- 2) List the components (places, etc.) from which the Reactor Coolant. Drain Tank (RCDT) collects water.
.OUESTION 2.09 (2.00)
. State all interlocks / actions (do not include administrative cetions) necessary to manually open containment sump suction E valve SI-8811A. Use valve numbers as shown on Figure 2-1.
DUESTION '2.10 (2.50)
The circulating water system pumps have many.
design / operating f eatures that provide protection f or the
- pumps and pump motors. Answer each of the following:
- a. What is the' purpose of the 6-inch recirculation line located across each circulating water pump?.
- b. List four conditicas/actione that automatically trip a circulating water pump. (Setpoints are not needed)
- c. List the. interlocks that must be satisfied to start a c2 r cul ating water pump. (Setpoints ARE required)
DUEST ION 2.11 (2.00) l 1
List ALL mhin steem isoletion signals with their accompanying setpoints/coinc2 dences/ interlocks.
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<*- 3. ;
s . v. ,
OUES110N 2.12 ( f. Ed'6 -
E . 't ,
a) ; List f the'dtenel generat.or AUTOMATIC trips that are U(i
[' enabl $dwhdnusinglthetsanual emergency starting mode.
!i - .
- g. i b) Which contrhl' mode'will the diesel generator be in'when j the'manbe19 emergency startingimode has been used? !
1 1 .i j
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x; I
DUESTION 2.13 (1.00) -
/ LLiot the fhve (5) co'aditions that cause an SSPS Genural. '
l
/ /
f(Narning. '
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- > ' ,/
c
,< ?
sOUESTION Ta14 -(1.00) ,
I y :
t .4 4 ,
List 10 loh,ds' supplied by, bus 100
' 'I,
) !
OUESTIQN. b.15 31.50) \
List the ' inputs "td .tNe subcoo12 ng margin monitor. .Be
-specific-inipou- description of the inputs. -Instrument ranges are-NOT required.
> n !
o' -
DUESTION 2.16 (r.OO) ;
e
^Each pressuri:er spray" control valve is provided with a bypass valve to allow a small continuous flow. One reason. ;
for needing this continuous flow is to prevent excess cooldown of jhe spray systen.. State the OTHER TWO reasons.
/ <
QUESTION 2.17 (2.00) q i .
- 1. State the purpose of ,the RCS loop stop, idol wtion valves i l': ($5IVs).
- d. State the interlocks that must be sot 1Efted.and the actions necessary to open the hot leg LGIV on an i sol ated loop. Do not include administratave requ2renents and actions. 4 t'
- 3. SL6te the purpose of the LSIV interlock.
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l GENCY SY@TEd@
N .
Etetil_DEG10N_1UR69D100_EGEEI _eUD_EUER
. t E
(1.50)
STION 2.18 lectric-driven auxiliary .
How cre the pump seals for the e f acdwater pump cooled?
diesel-driven au::iliary
)
How cre the pump seals f or the n.ot or . =
fcxdwater pump cooled? the 1 A Auxiliary Feedwater Pump l List three automatic trips d) i f or D 3)
(S tpoints are not requ re l (1.50) 100% power. Describe If
!UESTION 2.19 Cooler (RCFC) '
1 is operating normally atthe Reactor Containment Fan A33uma that Byron Unit tion of an SI signal is received. .
what happens to the operacyctcm and supporting systems when I
- c i.
(1.00) l; QUESTION 2.20 ious f uel handling tools.
th the name of the tool s.
Figures 2-2 through 2-5 depict var Complete the statements below wi i.
a) F1gure 2-2 is a ...
b)
Figure 2-3 is a ...
is a ...
c) Figure 2-4 1 d) Figure 2-5 is a l I
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(****4 END OF CATEGORY O2
2[__lNElB90gyIg_AUQ_gpNT6QLD ,
PAGE 14 [
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t QUESTION 3.01 (1.00) 7 With the pressurizer level control selector switch in the -
channel 459 position, a f ailure causes the f ollowing plant events. '(assume no operator actions taken)
- 1) charging flow reduces to minimum
- 2) pressuri er level decreases
- 3) letdown secured and heaters off
- 4) level increases until high level trip l
Which one of the f ollowing f ailures occurred?
a) Level channel 459 failed high b) Level channel 459. failed low c) Reference pressuri:er level stuck in the no load position d) Auctioneered Tave f ailed H1 due to a f ailed RTD t
QUESTION 3.02 (2.00)
Indicate whether the f ollowing statements regarding the bank "
)
overlap unit (BOU) are TRUE or FALSE. {
I a) The BOU determines which master cycler'gets the step up i l
or step down orders from the slave cycler.
l b) By selecting which banks receive the signal, the BOU can be used to provide the desired overlap f or control rod movement.
c) The DOU receives counts f roni the slave cycler. It counts once f or each step of rod movement for both groups in one bank.
d) The BOU only f unctions with the bank selector switch in the automatic or manual positions. When individual banks are sel ected , the BOU i s loci:ed out.
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.. PAGE 15
'3z__INg1BQdgylg_8Up_QQUIBQLg j
DUESTION 3.03 (1.50) R Azcume that'both of the Intermediate Range (IR) Nucl ear A reactor trip
. Instruments are' operating undercompensated.
is in the intermediate range.
hon occurred and reactor power Ancwer the f ollowing, TRUE or FALSE.
=
a). IR amps i'ndications are higher than they should be.
b) IR SUR indications are greater than they should be.
I, c) P-6 may not become unblocked automatically.
DUESTION 3.04 (2.00)
Indicate wheth'er the f ollowing statements regarding pressurtzer (P2R) pressure and level control are TRUE or FALSE.
- a. The signal generated by the pres.sure control system will control heater operation,, spray valve modulation and saf ety valve actuation. _
- b. P2R pressure control Ch 455 or 457 supplies the PZR pressure master controller which controls both spray valves, variable heaters, PORV PCV-455A, HP al arm bistable, LP alarm and backup heater energize .
bistable,
- c. The two spray valves operate together on the same control signal from the master controller.
- d. The spray valve control signal is Proportional Integral Differential (PID) controller compensated with the rate time constant adjusted to zero.
-QUESTION 3.05 (2.00) a) List the three (3) purposes for using rod insertion limits (RILs).
b) How and why do RILs sary with an increase in reactor power?
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GA__INEISWdgNIg_eNp_QQNIBQLg; PAGE 16 l
QUESTION' 3.06 (2.00)
There are f our '(4) types of low flow trips that_ protect the core f ollowing .a loss of ' coolant accident (LOCA). One of these-
-is called UNDERFREQUENCY.
a) Name the other three (3) _ trips.
b) Explain the philosophy (purpose) of having.an UNDERFREQUENCY trip AND explain how it works (include logic, coincidence, setpoints, interlocks, etc.)
-QUESTION .3.07 (1.00)-
Answer each of the f ollowing questions with one of three possible answers: OP Delta T,_OT Delta T or BOTH OP Delta
-T AND OT Delta T.
a) Which of-the delta T trips' protects against DNB7 b) Which of the delta T trips' setpoint can be either increased or decreased?
c) Which of the delta T trips does not nave a pressure -
input?
d) Which of the delta T traps' setpoint-can be reduced if the flux pattern is skewed toward the top of the
- core?
DUESTION 0.08 (1.00)
List the FOUR (4) PROTECTION signals that the Tave instrument provides.
QUESTION 3.09 (1.50)
List the five (5) possible conditions eeth of which can cause a COMPUTER ALARM / ROD DEVIATION SEQUENCE /NIS P.R. TILT alarm.
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OUESTION 3.10- (1.00) f Li st- the f our -(4) types of f ailures in a' logic cabinet that-gi ve' a ROD CONTROL URGENT FAILURE al arm. !
I OUESTION 3.11 (3.50)
Answer.each of the f ollowings a) If an RTD f ails OPEN, what happens to indicated temperature?
' b) If an RTD f ails due - to a SHORT CIRCUIT, what happens to indicated temperature?
c) If a thermocouple f ails OPEN, what happens to indicated temperature?
d) If a thermocouple f ails due to a 'SHORT CIRCUIT, what happcns to' indicated temperature?
e) Which of the two temperature detection elements, RTD or thermocouple, has the faster response time? ,
f) Which of the two temperature detection elements, RTD or thermocouple, develops its own electrical output?
g) Which of the two-temperature detection' elements, RTD or thermocouple, produces the more linear response? .
QUESTION 3.12 (1.00)
Describe'the purpose (protection) for the following reactor trips:
a) Power range high neutron flux trip (high setpoint).
b) Power range high neutron flux trip (Iow setpoint).
c) -Power range high neutron flux positive rate trip.
d)~ Power range high neutron flux negative rate trip.
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iUSIBUdENIS_6SD_G9UIB963 .
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(1.50) ts CI STION 3.13 the Power Range Instrumen !
is lost to one of will (11) percent, ntrol R) Powernuclear instruments. is at eleven If the powertrip? level the (9) percent will
- c. mi the reactor is at nine If the power level b.
reactor trip? tion at the PR drawern gge i
Describe / explain the indicathat control power has bee
- c. (r eveals) i that shows r lost, (2.50) 3.14 the Intermediate Range DUESTION is lost to one of wer Inatrument powerinstruments, (IR) nuclear the power level is below P-10 will h in BYPASS, andthe the IR dra _
- a. If has reactor its LEVEL trip? TRIP switc lux is above P-10 and the IR fthe reactor trip?
If the power level will -
- b. trip has been blocked, (3) indi c atint ons / al arms loses the three (describe t
- c. Name that are received when an IR instrume !
l:
instrument power, (1.00) ated on 3.15 (4) ,,
DUESTION tary Level Trip switch locfour SS Placing the two-position ro the BYPASS position doesit causes a SR the Source Range drawer inthese is that things. One of the f ront panel ,
al ar m.
ennunci ator Ref er to figure 3-1, (2) others.
List twothe Source Range drawer.
of NEXT PAGE
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. OUESTION 3.16. (2[00) I
'A reactor trip always causes a turbine trip. The tripping, of-the turbine would result in a generator trip if there wsre no time delay. There are two (2) saf ety-rel ated .
raasons for having a generator trip time delay. Name.and explain the two (2) reasons. '
t o
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w . .
EGi__EB9GEDWBE5_:_NDBdebt_ GEN 9BdeLx_EdEB@gNQy_@NQ PAGE 20 E09196001G86_G9BIBDL .
I QUESTION 4.01 -(1.00)
Fill in the blanks f or each of the following statements.
- a. The UNIT 1 Lo-Lo S/G water' level setpoint is
_________- percent.
- b. The UNIT 2 Lo-Lo S/G water level setpoint is percent.
- c. The UNIT 1 Hi-Hi S/G water level setpoint is percent.
- d. The UNIT 2 Hi-Hi S/G water level setpoint is .
_________ percent.
QUESTION 4.02 (2.00) l
~
List the immediate actions f or an ATWS per 1BFR-S.1, Response to Nuclear Power Generation /ATWS. (No detail as to how each step is performed is required)
GUESTION 4.03 (1.00) l
- a. What is the 10CFR2O exposure limit f or the whole
- body.
- b. When can the limit specified above be exceeded.
b QUESTION 4.04 (1.50)
List the Minimum number of channels operable and the number of channels needed to TRIP for each of the following reactor trip functions:
- a. Power range neutron flux h2gh setpoint (Mode 1)
- b. Inter. mediate range neutron flux (Mode 1)
- c. Source range neutron flux (Mode 3)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
7 p
- !z__EBDGEDWBEE_:_NQBdeLi_etNQBdebt_EMESQgNQy,8NQ- .PAGE 21
- BSD196901G86_GQNIBQL *
- l it
!OUESTION 14.05. (2.00)
BOPIRC-1, Startup of an RCP, contains the RCP starting g , duties. . LIST ALL RCP starting duties.
. QUESTION ~ 4.06 - ( 1. 00) 1BEP ES-0.2, Natural Circul ati on Cooldown Unit .1, contains a etatement in Attachment A that calls for the operator to
" Check Containment'for ADVERSE conditions."- State the-criteri a f or having an. ADVERSE containment. Be precise and list setpoints.-
QUESTION 4.07 (2.50)
List precisely the RCP trip requirements as stated on the ,
Operator Action Summary f or 1BEP-O Series Procedures. Include-
- adverse containment'setpoints as applicable.
r DUESTION' 4.08 (2.00)
LA reactor startup is being perf ormed in accordance- with
.BGP-100-2. ,
a '. If the reactor goes critical below the ROD BANK LOW INSERTION limit, what must be done?-
Hb . If the reactor goes critical below the ROD BANK LOW-LOW INSERTION li'mi t, what must be done?
- c. Enplain the basis for perf orming the actions of part a) and b).
(*244* CATEGORY 04 CONTINUED ON NEXT PAGE *****)
'.4t'__EBQGEDL' BEE _:_N9Bdebz_BENQBUSLg,gdgBQgNGY_OND PAGE 22
.60D196991GOL_G9 BIB 96 . .
l QUESTION 4.09 (2.50)
A plant heatup is being conducted in accordance with BGP-100-1,. Plant-Heatup.
Initial' Conditions ares j
- SI signal'is present (it was not i manually initiated)
-Feedwater system it still isolated
-No P-14 signal is present
- a. State, in the correct sequence, the steps necessary to' reset the feedwater isolation signal.
- b. If, as an initial condition, a P-14 signal had.been ,
present, what operator actions would be required to clear this signal?
OUESTION 4.10 (1.50) _
Byron Station Radiation Protection Standards, BRP-1000-A1 identifies 2 types of Radiation Work Pgrmits (RWPs),
- a. State the conditions that require a TYFE 1 RWP. _
- b. State how l or.g a TYPE 2 RWP i s vali d.
QUESTION 4.11 (1.00)
What is used f or stabilizing RCS pressure f ollowing an SI, per 1BEP ES-1.1, SI Termination Unit 1,.if pressurizer, normal spray, and auxiliary spray are not available.
QUESTION 4.12 (1.00)
.What is done, per 1BEP ES-0.1, Reactor Trip Response, Unit 1, if two (2) or more control rods are not fully inserted after a reector trip? A detailed explanation as to how the action is to be performed is NOT necessary.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l
. *d2__EB9CEDWBEE_:_NQBbeLt_8ENQBdeLi_EDEB9ENRY_6ND PAGE 23 :
. BOD 196991G96_G9 NIB 96 , ,
.y-QUESTION 4.13 (1.00)
- c. The' Functional Restoration Procedures (BFRs) be can entered f rom either ofi 2 ways. State the 2 ways. .
- b. Normally, actions called f or by the BFRs take precedence over those actions called for by non-BFR procedures.
However,.one plant condition (casualty, event, etc.),
covered.by a Contingency Action Procedure, is an exception to thi s rul e. (i . e. , its actions take priority L over those of all BFRs.)
~
State the plant condition that is covered by this Contingency Action Procedure (BEP). .
QUESTION 4.14 (1.00)
State the conditions per BOA Section 5, WS System -
Malfunction, that require tripping the reactor.
OUESTION 4.15 (1.00) -
Concerning the steam generator l evel control system:
- a. What is the 100% program level for UNIT 17 ,
- b. What is the 100% program level for UNIT 27 OUESTION 4.16 (2.00)
Step 4 of 1BGP 100-6, Refueling Outage, directs the operator to verify / borate all filled portions of the RCS and the
. refueling canal sufficiently to ensure that the more restrictive of two (2) plant conditions is met.
- a. State the two (2) plant conditions.
- b. Prior to step 4, there is a CAUTION that limits the MAXIMUM boron concentration. State the MAXIMUM boron concentration allowed and e:: plain why a MAXIMUM limit i s imposed.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
di__EB9GEDWBED_:_NQBd8Li_8DN9Bd862_EDEBOENRX 8ND PAGE 24 68D196991986_G9 NIB 96 -
.e ,e . .
QUESTION 4.17 (1.50)
Step 4 of 1BCA-0.0, Loss'of All AC Power Unit 1, directs the operator to verify that the RCS is isolated.
Name the valves.that the operator' checks closed.
(Valve numbers are not needed; however, a description of the valve and the quantity of valves fitting that description IS needed) l DUESTION 4.18 (1.50)
Define the followings (Answers need not be verbatim from-Technical Specifications) a.- Channel Check
- b. Axial Fl u:: Difference _
- c. Duadrant Power Tilt Ratio.
i l
)
I I
i l
(***** END OF CATEGORY 04 *****) l
(************* END OF EXAMINATION ***************) l i
l 1
~ w:,e
.~..-
2 w a og!
s a V,t
- 1/2 at g , ac :
KE = 1/2 av' a * (Vr = to )/t A
- 3 m, ,; . S=A,e PE = agn Vf = Y, + at * =.e/t 1=
an2/t1/2 = 0.693/t1/2 W = v AP A* nD 2 1
1/2'fI*II'1III)) b 4 -
[(t)fg) + (ty))
{
t.E = 931 am Q ,= ph m = Va,Ao g . g , .tx j Q = aCpat i 6 = UA4 T I = I g e'"*
Pwr = d fah I = 1, 10**/ M TVL = 1.3/u 5
P = P,10 "'I*) HVL = -0.693/v P = P,e*/
SUR = 26.06/T ,. SCR = S/(1 - K,ff)
CR, = 5/(1 - K,ff ,) !
SUR = 26e/t= + (s - o)T CR;(1 - K,ffj) = CR2 II ~ keff2)
T = (t=/o) + ((s - o)/Io] M = 1/(1 - K,f,) = CR j/CR, i T = s/(o - s) M = (1 - K,ffo)/(1 - K,ffj)
T = (8 - e)/(Is) SDM = ( - K,ff)/K,f, I o = (K,ff-1)/K,ff = AK,ff/K,ff t' = 10 secones I = 0.1 seconds )
o = [(t*/(T K,ff)) + [i,ff (1 + IT))
/
d =1d Ij Id
) 2 ,2 gd 2 P = (seV)/(3 x 1010) jj 22 l 2
I = oN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g) ,
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x.1010ap3 1 gal.=3.78 liters 1 kg = 2.21 lem .
I ft< a 7.48 gal. 1 np = 2.54 x 103 8tu/nr .
Density = 62.4 14/ft3
~
1 rar = 3.41 x 106 8tu/hr Density = 1 9m/cm3 lin = 2.54 cm Heat of vaporiration = 970 8tu/lem =F = 9/5=C + 32 Heat of fusion = 144 8tu/lba .
'C=5/91*f-32) 1 Ata = 14.7 psi = 29.9 in. Hg. 1 8TU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in. .
. e " 2.718_:. .
74, , 3 h ,"
[f , _' h :;
u%; -
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0 7
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4 FIGURE 7 -2,
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FIGURE 2-3 ,
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FIGURE 2-9 i
4 e
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e SAFETY SCREW 3! OPERATING HANDLES o
l / h kW
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/
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. " HH GS t
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' 3 e, *. #
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/
R O
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R U
u S e O E
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e * .
R T e *= F O
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/ \
A R
E P
c*"
o e s
- c c*
.*a
- a O _
S L _
/ \ O * " e. E
+ s + .
l a
l I.
N
. A
= P _
e s L _
S == s s
T N
T L 8
- - P , v a
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R I
8,-
8 V 1l lI R e 0 T F R
O1 0 T
e 'r. 4 '.
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PAGE 25
, $1z_iEBING1ELEE_QE_NWGLEBB_EQWEB_ELONI_QEEBBI19Ns.
. ~ IBEBdQQ1Ned1 gem _SE61_IB8BEEEB_GUQ_ELUIQ_ELQW
' ANSWERS,-- BYRON 1
-87/05/04-BITTER,S 3
. ANSWER. 1.01 (1.00) c) Keff = .990'(1.0) g ' REFERENCE Commonwealth Edison, PWR Fundamentals, Reactor Physics, Chs.
'5,6,8 LOO 1000A404 192OO2K110 192OO2K114 ...(KA'S).
ANSWER 1.02 (1.00)
'c.) (1.0)
REFERENCE Commwealth Edison, PWR Fundamemtals, Rear'.or Physics, Ch.
-7, 8 . . .
192OO8K110 192OOOK114 ...(KA'S) o ANSWER :1.03 (1.00) b) ( 1'. 0 )
REFERENCE
- Commonwealth Edison, PWR Fundamentals, Reactor Physics, Chs.
7,8-015000A102 015000K506 192OOOK105 ...(KA'S)
ANSWER 1.04 (1.00) ,
c) (1.0)
REFERENCE Commonwealth Edison, PWR Fundamentals, Nuclear Theory, Ch.
3.
001OOOK528 192OOBK127. ...(KA'S)
- 1'z __ EBI N91 ELE E_9E_NUGL E 88_ E9 WEB _EL8NI_ QEgB8I1QUt PAGE 26 IdEBD99YU8019Et_dE8I_IB9NSEEB_0ND_ELylp_ELQW
- ANSWERS -- BYRON.1- -87/05/04-BITTER,5 ANSWER 1.05 (1.00) c) ( 1 '. 0 )
REFERENCE Commonwealth' Edison, PWR Fundamentals', Reactor Theory, Ch.4 192OO6K107 ...(KA'S)-
i j
ANSWER 1.06 (1.00) c) (1.0)
REFERENCE Commonwealth Edison, PWR Fundamentals, Thermodynamics, Ch.3 193OO3K125 193OO4K115 ...(KA'S)
ANSWER 1.07. (1.00) ,,
b) (1.0)
REFERENCE Commonwealth Edison, PWR Fundamentals, Thermodynamics, Ch.2 OO2OOOK507 193OO3K125 ...(KA'S) -
ANSWER 1.08 (1.00) c) (1.0)
REFERENCE Commonwealth Edison, PWR Fundamentals, Fluid Flow, Ch. 1.
OO'OOOK505 191004K109 191004K114 ...(KA'S)
ANSWER 1.09 (1.00) b) (1.0) i l
..
- 1E__EBING1ELEE_QE_ NUCLE 88_EQWEB_ELBUI_QEEB811QUt PAGE 27
'IUEBdQQYU001GEt_SE01_166UEEEB_6UD_ELUID_ELQW ANSWERS ---BYRON 1 -87/05/04-DITTER,5 7
REFERENCE
. Commonwealth Edison, PWR Fundamentals, Reactor Theory, Ch .' 9 OO1000K506 ...(KA'S)
ANSWER 1.10 (1.50) s) INCREASES ( 0.' 5 )
b). INCREASES-(0.5) c) INCREASES (0.5)
REFERENCE Commonwealth Edison, PWR Fundamentals, Fluid Flow, Ch. 1 191004K106 191004K115 ...(KA'S)
ANSWER 1.11- (2.00)
.a) LOWER THAN (0.5) .
b) LOWER'THAN (0.5) c) LOWER'THAN (0.5) ,
d) DOES NOT DIFFER FROM (0.5)
REFERENCE
-Commonwealth Edison, PWR Fundamentals, Reactor Theory, Ch. 7 192OO6K107 192OO6K110 192OO8K101 192OO8K102 ...(KA'S)
ANSWER 1.12 (2.50) a) INCREASES (0.5) b) DECREASES (0.5) c)1 INCREASES (0.5) d) INCREASES (0.5) e) DECREASES (0.5)
REFERENCE Commonwealth Edison, PWR Fundamentals, Heat Transfer, Ch.9 193OO8K105 ...(KA'S)
- '__EBINGIELEE_QE_UWGLEGB_EQWEB_ELONI_QEgB811QNg 1z PAGE ...
- ISEBdQDXUBd1GEt_UEBI_1BBUEEEB_GUD_ELWID_ELQU
- ANSWERS - . BYRON 1 -87/05/04-BITTER,5 ANSWER 1.13 (2.00) a.). Remain the same (0.50) b.) Decrease [O.503 c.) Increase [0.503 d.) Decrease [0.503
-REFERENCE' Commonwealth Edison, PWR Fundamentals, Thermodynamics, Ch. 3 193OO4K115 ...(KS)
ANSWER 1.14 (1.50) l a) TRUE (0.5) b) FALSE (0.5) c) 44WE (0.5)
F/iLS& wer/u/t1
~
REFERENCE :
-Commonwealth Edison, PWR Fundamentals, Thermodynamics, Ch. _
I 3.
015000A101 015000R504 1930071:.106 193OO7K108 ...(KA'EJ ANSWER, 1.15 (1.50) )
a) 'TRUE (0.5) b) FALSE (0.5) s l
c) FALSE (0.5)
REFERENCE Commonwealth Edison, PWR Fundamentals, Heat Transfer, Ch. q 9 j REFERENCE 1930091:105 ...(KA'S) j l
_ j i
.. ~I.._EBING1ELES_QE_NWGLE88_EQWER_ELONI_QEE68119Nt 1 PAGE 29 TUEBbQDXNed1GSt_ NEAT TRANSEgB_AND_ELVID_ELOW
, ANSWERS -- BYRON 1_ -87/05/04-BITTER,5 i
ANSWER. 1 .~ 16 (1.50) o) FALSE (0.5) b) TRUE (0.5) c) TRUE (0.5)
REFERENCE l COMMONWEALTH EDISON,.PWR Fundamentals, Reactor Theory, Chs. !
~
.3,4,5 .
192OO4K106 192OO4K110 192OO6K111 ...(KA'S) .
I ANSWER 1.17 (2.00) l s) SG pressure - stable or' decreasing (0.5)
I b) RCS hot leg temperature - stable or decreasing (0.5) j i
c) Average of ten highest core exit 7 /Cs - stable or decreasing (0.5) -
d)'RCS cold leg temperature - at saturation temperature for l SG pressure (0.5)
REFERENCE BEP, Byron Emergency Operating Procedures, ES-0.1 193OOBK122 ...(KA*S) l l
ANSWER 1.18 (1.00) c) Tensile stress (0.5) ;
l b) Existing flaw (0.5) l REFERENCE Commonwealth Edison, PWR Fundamentals. Heat Transfer, Ch. 6 REFERENCE 193010K101 ...(KA'S) l 1
PAGE 30
.. .it__EB1NG1 ELE 5_QE_NYGLE85_EQWEB_E68WI_QEEB8119Nt ISEBdQ2XW801GEt_ SEAT TRANSEEB_6ND_ELY1D_EL9W
, ANSWERS -- BYRON 1 ~B7/05/04-BITTER,S n,, m a 1.19 ANSWER (1.00) u we nsaa sala G < ge-) w (/2F/b7 a) Use steam traps (0.5) b) U e drains on steam lines (0.5)
$ ? fl >M 'l*h Commonwealth Edison, PWR Fundamentals, Fluid Flow, Ch.1 193OO6K104 ...(KA'S)
ANSWER 1.20 (1.00) 1.) 10exp-8 amps is well above the source range. (0.5)
(Source neutrons can be neglected.)
2.) 10exp-8 amps is below the POAH. (0.5) (Doppler and moderator temperature effects are not present.)
REFERENCE Commonwealth Edison, PWR Fundamentals,' Reactor Theory, Ch 9 192OOOK112 ...(KA'S)
~
l ANSWER 1.21 (1.50) a) Convection (0.5) b) Radiation (0.5) c) Conduccion (0.5)
REFERENCE Commonwealth Edison, PWR Fundamentals, Heat Transfer Ch 1 TPOs 9,11.
193OO7K101 193OO7K104 ...(KA'5)
i 4
.PAGE 31
.. '8, [__ELONI_ DES 10N_INGLWDIND_E9EEIY_BND_EDEBEENGY_fYEIEd5
, . ANSWERS -- BYRON 1 -57/05/04-BITTER / S 1
ANSWER 2.01' (1.00)
- 1) Water Fire Protection. System (sprinklers, foam, and deluge) (0. 2) -
- 2) Carbon-Dioxide. Extinguishing Subsystem (0. 2 ) ;
- 3) Halon'1301 Subsystem (0.2)
- 4) Fire Detection Subsystem (0.2)
- 5) Control Room. Emergency Breathing Air Subsystem-(0.2)
REFERENCE )
Commonwealth Edison, braidwood System Description, Ch.57, Terminal. Performance Objective 2- .
OB6000 GOO 7 ...(KA'S)
ANSWER 2.02. (2.00) a) hot leg val've open (0.15) and (0.15) cold' leg valve open (0.15)
OR (0.10) cold _ leg valve shut (0.15) and (0.15) bypass valve open
- (0.15) 1 559 5/2s/e7 l b) 1. 9CP ril liet p .,; p .2 sp .:. ting "' . 5 ) ;
- 2. RCP oil lift pump (+r25) is at-j least 600 psig -(C .gischarge
_SE pressure 6* % s/se/s7 j
@6 spe[d j REFERENCE Commonwealth Edison, Braidwood System Description, Ch.13, Terminal Performance Objective 10a.
OO3OOOG013 OO3OOOK113 OO3OOOK411 ...(KA'S) l l
i l,
I i
,4 - Et__ELONI_ DESIGN INGLWDING_feEEIY_eND EDEBEENGY_EYEIEUS PAGE 32
- s. -
. ANSWERS'-- BYRON.1'
-57/05/04-BITTER;S ANS'WER 2.03 (2.00) 5fK 4 i)*hA ,
Any .f our of the 44se listed below (0.5 apiece)
- 1) DG heat exchangers
- 2) CCW heat exchangers
- 3) Control Room Refrigeration Room Condensers
- 4) Reactor Containment Fan Coolers 5)
L) h*esel-dr*wn Aun,hary feed famfPrimary Engiw He Containment RefSyff'rigeration Y2 Ye7 Unitb. Eu h REFERENCE Commonwealth Edison,.Braidwood System Description, Ch . 20, Terminal Performance Objective 3 076000K119 ...(MA'S)
ANSWER 2.04 (1.00)
- 1) no::le bypass flow (0.25)
- 2) control rod / instrument thimble bypass flow (0.25)
- 3) baffle wall bypass flow (0.25)
- 4) head cooling bypass flow (0.25)
REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 10,
- Terminal Perf ormance Objective 3 OO2OOOGOO7 ...(MA'5)
ANSWER 2.05 (1.50)
- 1) alarm (0.3)
- 2) CV-110B shuts (0.3)
- 3) CV-111B shuts (0. 3 )
- 4) BATP trips (0.3)
- 5) makeup control trips off (0.3)
REFERENCE Commonwealth Edison, Braadwood System Description, Ch. 15b OO4000 GOO 7 ...(UA'S) j
- m. - - -
WW. p , gg GI__ELONIiDESIEN_1NO,lJDIN4d20EEIY_869_EUE69ENGY_EYSIEdf PAGE 03
\ J , .
p ANSWERS -- BYRON 1 F -87/05/04-BITTER;S Vp +{t .. 1 t, \j y- L ; [- p , .i jtj v / jt 3 y) + s
.,j (
,r ANSWER 2.06 (1.00)/ 4 9 . i i
1
- 1) computer inverter N NY .a
- 2) main turbine DC emergency oil pump '^ . N
- 3) generat.iv emergency air side seal oil pump i v . 2L aDI-
.Cr ) SG feedspump turbine emergency bearin oil pumps 4G. 23) @
- 5) No : C % .nm (kenaue \/o lse. And g .15 %
REFERENCO '; vocox
- Com'nonweaf t h : Edi son , Braidwood Plant Description, Ch. Ob, Terminal F&rfprmance Objective 1 4
063OOOK20) . ...(KA'S)
I'9 ['
h8
'e sr/zy%
' ;a~'NNSWER 4 >\\ se *$l)a su
.2.07 (1.50)
A41% ef ik fvilessy fQe (C,5 haf 'te) 60$ f,j,7 ctSY,J f ,
' t ; ri) The spent' fuel pit cooling pump's suction penetrates y)fvowAcciderd4[ of the poci near the normal water level so.that gravity t cn4ia. ,, te,. o draining due to a line break is prevented, M wt jmq w
- i
- 2) The return lines contain antisiphon h' oles. #+rft' #
k19f ACsd pt[h
- 3) The spent fuel pool liner has continuous weld seam c)7?n ~ , .
lei -//2 a
[ drains.K (Any leakage would drain to a central collection g)IF/J wig g c point f or leak detection ) F r-54 (section in parenthesis no't. required
$7 for full credit) g7'g.s ec48 REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 51, gg yg /
Terminal Performance Objective 51 033OOOK401 ...(KA'S)
ANSWER 2.08 (2.00)
- 1) key-lock valve 0.50) j g 3,g3 (gf,,g T4 lt
- 2) excess letdown (0.25)- Odid lsdM reactor vessel head vent (0.25) g /
/g Yaf(/p) accumulator drain / vent (0.25) spent f uel pit iilter (0.25) loop header drains (0.25) 4t2 seal on RCPs (0.25)
REFERENCE Commonwealth Edison,Braidwood System Description, Ch. 48a, Terminal Performance Objective 10
~
c 1 [ C,.'__EL6NI_DEEIGN_INGLWDINE_EeEEIY_6ND_EDERQEUQy_Qy@ Igd @ PAGE 34
' , ANSWERS -- BYRON 1 -87/05/04-BITTER,.S
~
060000K107 ...(KA'5) c __
ANSWER 2.09 (2.00)
- 1) Either 8701A or 8701B must be shut. '
(o.u) 6 <A @
3; l' 2) 8812A must be closed. F*,4+-(0,97) T 2 9 8 -
- 3) CSOO1A must be closed. 4 x =. w g .~ . - -
0,60
, - (, g REFERENCE Commonweal th' Edi son, Braidwood System Description, Ch.18, Terminal Performance Objective 6c OO6000K409 ...(KA'S)
ANSWER 2.10 (2.50)
- s. The recirculation line limits the hydraulic f orces across the. pump when it_is started with its discharge valve closed. (0.5) -
b.
Any foar 4 %* %Ilowsm (0,LT ef*ceD
- 2. Phase overcurrent .
- 3. Ground overcurrent
- 4. Field monitor relay F. Ob ' y el e cio ed h
- c. 1. discharge valve must be closed (0.5) -
- 2. non-essential service water pressure must be at least 11 psig (0.5)
REFERENCE Commortwealth Edison, Braidwood System Description, Ch 39, Terminal Performance Objectives 3,4 075000K401 ...(KA'S)
I
- i
2 z._ _EL ONI_ DE S10N _ING LU DIN @ _@ @ E gly _8N p_g d g BQ g N QY _@y @lg d @ PAGE 35
- ANSWERS -- BYRON 1 -87/05/04-BITTER,5 ANSWER 2.11 (2.00)
- 1) M a n u a l --- (1/2 switches on MCB) (0.5)
- 2) Auto--
a) High-2 containment pressure-8.2 psi g, 2/3 detectors. (0.5) b) Low steamline pressure-640 psig, 2/3 detectors in 1/4 steamlines if low steamline pressure SI is not blocked. (0.5) c) Steam pressure rate- negative 100 psig/50 seconds, 2/3 detectors in 1/4 steamline, only if l low steamline pressure SI is blocked. (0.5) l REFERENCE l Commonwealth Edison, Braidwood System Description, Ch. 23, l Terminal Performance Objectives 5,1 039000K405 ...(KA'S)
ANSWER 2.12 (1.00) a) 1. generator dif f erential (0.25) -
- 2. (engine) overspeed (0.25) b) i sochr onous (0,5)
REFERENCE "
Commonwealth Edison, Braidwood System Description, Ch. 9, Terminal Performance Objective 5 064000K402 ...(KA'S) j d
' ~
bh [
An ig o hollcair
- 1. Loss of a power st ply w.,
0.1a((Ch .
,p ,
- 2. Removal of a printed circuit card F+rG+ 7) U#3 A b" ' 1 nct en CFF-A bypass breaker for a train being closed & +riiH
- 4. Loss of continuity through the ground return fuse (+.2) ng m (FF
- 5. Pl acement oi switches within a train to positions used q g, ,,,3 {.,g/
during testing W) nt,p g pgp REFERENCE ,
Commonwealth Edison. Braidwood System Description, Ch. 60s, W 6 N n'Y in Terminal Performance Objective 6 Of6M76-
9.__ELONI_DEH10N_INGLUDING_E8EEII_8UD_EdEEGENGY_EYSIEd5 PAGE 36 ANSWERS -- BYRON 1 -87/05/04-B1TTER,S i
ANSWER 2.14 (1.00) -
Any 10 of the f ollowings (0.1 apiece) c) 41'60V/480 transformer to bus'132X m) 84 s (W g b) centrifugal charging pump 1B c) safety injection pump 1B h) 8us 2VL s'/2B/t'7/ -
d) residual heat- removal pump 1B a) control room ref rigeration unit OB f) containment spray pump 1B g) component cooling water pump 1B h) feed to component cooling water pump O i) essential service water pump 1B j) auxiliary building ventilation supply f an OB ,
k) auxiliary building ventilation exhaust fan OB g7 l'
E) 4I40V/%0 +mnsfeneer 10 ha3 l32.2 REFERENCE
[f6'fftejiQ 4 cff)
Commonwealth Edison, Braidwood System Description, Ch .hs,
-Terminal Performance Objective 12 062OOOK201 ...(KA'S)
ANSWER 2.15 (1.50)
- 1) pressuri:er pressure (C.D .13 ivr ecco os 1.n e 4-i
) PT-4"s -
b) c)
'T 56 457 f2h7s PT-
^
t- :rpt:b!r 'ter: tr: 211 prc :uricer pr;;sure c t. -o iw iM
- 2) RCS loop wide range r (03)
G.13 iv ==J. wi th; 2'
'ik
! :: -- p + + 1 = m1+.cna : _ _ . . .i; m e c.,ve pressure cnannelsi
- 3) -- - ' '" ~ t e .. , y = , o ;. w . . 'O W ' Con Ya t an d b ka vp N# " 8 h*
+t+ 4v. crT t e m n = r m + i i r .- <^ & L ,, %g
?gy 4>
S y ;pr;;r Lor \a s um d fNssatY (0. 3) ( Wp gpgjgp R FERENCE Commonwealth Edison, Braidwood System Description, Ch. 346, Terminal Performance Objective 6 074000K101 ...(KA'S)
ht _E(6MI_QEgigN_1NQ(QQ10@_E8EEIY_6NQ_EdEB@ENQY_@YSIEd5 PAGE 37
. ANSWERS -- BYRON 1. -87/OS/04-B2TTER iS ANSWER, 2.16 (1.00)
- 1) to minimize 1 thermal shock to the-lines (0.5)
REFERENCE Commonwealth Edison, Braidwood System' Description, Ch. 14, Terminal Performance Objective 9' 010000A101. 010000K401 010000K603 ...(KA'S) 1 ANSWER 2.17 (2.00)
- 1) They serve to isolate -a loop in the event that the loop requires maintenance. (0.5)
- 2) The cold leg LSIV must be closed (0.5) and the control switch must be placed in the OPEN position. (0.5)
- 3) The LSIV interlock ensures that, when an isolated loop is returned to service, a cold water accident does not -
occur. (0.5)
REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 12
- OO2OOOK409 ...(KA'S)
ANSWER 2.18 (1.50) 1
- 1) The pump seals are cooled by allowing a small amount of AFW to flow along the shaft of the pump. (0.25)
- 2) The pump seals are cooled as described in the first answer above. (0.25) (
f 3). a) undervoltage (on bus 141) (0.33) {
b) overcurrent (0.33) '
l c) (low-low) pump suction pressure (0.33)
REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 26, )
Terminal Performance Objectives 4,13 1 061000 GOO 4 ...(KA'S) 1
. i.__ELGUI_DEE198_10GLUDIUQ_EBEEIY_86D_EdEBEEUGY_EYSIEUS PAGE 30 l
ANSWERS -- BYRON.l' -87/05/d4-B1TTER,S
-ANSWER 2.19 (1.50)
- 1. Operating RCFCs shift to slow speed. S:.5,
& t.ack .373
'2. Non-operating RCFCs start in slow speed. '^5)---
- 3. Chill d e t er 'l=tw all RCFC cnilleru w e i. m m i. 13 reca ;d. '^ E?
5)( is fyfasstk fc cosf}74l g m& ch'e fjty'5 YeSul$1h in cMller wi 2 i
REFERENCE 4, SY valvej8cFC for sf. . 4, is b'a gb c /*J!
Commonwealth Edison, Braidwood System Description, C 42,
, h(f.57<9/m67]
i(k Terminal Performance Objectives.6,Ba 022OOOGOO4 022OOOK104 ...(KA'S)
ANSWER 2.20 (1.00) s a) thimble plug. handling _ tool (0.25) b) burnable poison rod assembly handling tool (0.25) c) irradiation sampling handling tool (0.25) d) new. fuel handling tool (0.25)
REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 52,- .
Terminal Performance Objective 8 e
i a
i l
_ i
.A 4
f
']
PAGE- 39 07s_i1NEIBWdENIE_8ND_GQNIBQLE .
..,L A N S W E R S ; - - B Y R O N - 1 ' -87 /05 /04-B I T TER ,- S
' ANSWER 3.01 -(1.00)
'a;'(1.O)'
< REFERENCE'.
g
- ' Commonwealth Edison, Braidwood System Description, Ch 14
~
Terminal Perf ormance Objective 21 011000A201~ ...(KA'S)
,* ANSWER 3.02 (2.00) a)Lfalse (0.5) b) true (0. 5) '
c) false (0.5) d) true (0.5)
REFERENCE Commonwcalth Edison, Braidwood' System Description, Ch. 28, Terminal Performance Objective 5 OO1000K403 ....(KA'S) ..
ANSWER 3.03 .(1.50) a) . true (0,5) b) false-(0.5) c) true (0.5)
REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 32,
. Terminal Performance Objective 3 015000K501 015000K502 ...(KA'S)
ANSWER 3.04 (2.00)
- a. FALSE (0.5 each)
- b. TRUE
- c. TRUE
- d. .TRUE REFERENCE Commonwealth Edison, Braidwood System Description, Ch.14 Terminal Performance Objective 20.
010000K602 ...(KA'5)
,. 'Il__INHI6WMENIE_eND_GQNI6QLE! , PAGE 40.
4
~
-: ANSWERS -- BYRON 1 -87/05/b4-BITTER,S-ANSWER 3.05. (2.00) a)-1, to maintain adequate SDM (0.5)
- 2. .to maintain nuclear peaking f actors within. limits [to vhaidain b-(0.50)
- 3. to. minimite the ef f ects of an ejected rod,(0.5) accef[k)
Pouw 0ad<dahnt Lo a,rik
'(+c 1 % f th le 1N %.fs c Ok miso.b orm;. m
'b) RILs increase with 2ncreasing reac(ok p*ow)er ecaus the Y powr'r defect increases with increasing reactor power.
(0.5)
- fd DJ)
REFERENCE Commonweal th Edison, Braidwood System Description, Ch.'28,
-Terminal Performance Gbjective 19c OO1000K508 192OO5K115 ...(KA*S)
ANSWER- 3.06 (2.00) a) 1. UNDERVOLTAGE (0.25)
- 2. LOW FLOW (0.25)
- 3. RCP BREAKER OPEN (0.25) +
b) The underfrequency trip provides reactor protection -
f ollowing a major network f requency disturbance. By
. tripping.the RCP breakers, a minimum coastdown time is ensured (0.5). All RCP breakers trip and the reactor trips if an UF condition -below 57.0 H:- exists on 2 of.4 RCP -
buses -(0. 75) .
REFERENCE Commonwealth Edison, Braidwood System Description, CH. 60b.
Terminal performance Objective 4
-OO3OOOK304 OO3OOOK502 012OOOK402 ...(KA*S)
ANSWER' 3.67 (1.00)
- a. OT Delta T (0.25 each)
- b. OT Delta T
- c. OP Delta T
- d. OT Delta T REFERENCE Comneonweal th Edi son, Braidwood System Description, CH 12, 60 012OOOK611 ...(KA'S)
~-
. .j L.__INEIBWt1ENIS_BNQ_GQUISQLS PAGE 4 l'
' ANSWERS -- BYRON 1 -87/05/d4-BITTER,S ANSWER 3.08 (1.00) ,
- a. OT Delta T (0.25 each)
- b. OP Delta T
- c. Low Tave
- d. Low-Low Tave REFERENCE Commonwealth Edison,_Braidwood System Description, Ch.12, Terminal Performance Objective 6 015000K601 ...(KA'S)
ANSWER 3.09 (1.50) h Any ~ five ch hepI fleg*q
- a. 2% radial f l u>:
(03 afsi'd tYit.
/
- 30. 2 :;c;.?
g.) CS P gre4fer fl%
- b. Rod program wrong sequence. (overlap e,wn) h,}4225" Sfef5,Q ,gg
- c. Shutdown bank is less than 220 steps. -
- d. One rod is greater than 12 steps from the bank.
- e. One rod is greater than 12 steps from other rods. i,.) AAy S# PM /ess G. 25 fened axistk Slwy %lt , 3& " kn us-REFERENCE sfeff offer'sp.
Commonwealth Edison, Braidwood System Description, Ch. 28 , (,,d 6 is 9res/<r' Terminal Performance Objective 13 -ti% 2.25'sfe/I, OO1050A201 OO1050A202 OO1050A204 OO1050A206 ...(KA'S)
ANSWER 3.10 (1.00)
- 1. Missing card (0.25 each)
- 2. Slave Cycler failure .
- 3. Oscillator iailure ( fulve fg g lure (Occilldr 4!ser fall /41t
- 4. SCDE circuit failure REFERENCE Commonwealth Edison, Braidwood System Description, CH. 28 Terminal Performance Objective 13 I
C
, .) __lNEIBWENIE 6ND_GQUlBQLE l
-87/05/64-BITTER,S- l
- ANSWERS - EBYRON'1 i e 1.
t.
g ANSWER 3.11 (3.50) o) indicated T f ails high (0.5) b)1 indicated T fails low (0.5) c) indicated T f ails low (0.5)
'd) indicated T' fails low (0.5) .
o) T/C (0.5) f) T/C-(0.5) g) RTD (0.5) I; REFERENCE' Commonwealth Edison, PWR Fundamentals, Instrumentation and ,
Control, Ch . ' 1 , .
191002K113 191002K114 ...(KA'S) i.
ANSWER- 3.12 (1.00)
- a. . Protects against. excessive heat flux leading to DNB caused by fast reactivity excursions. (0.25)
- 6. Protects against excessive power. excursions caused by high startup rates or. react)vity additions.(0.25) .
- c. Protects against an ejected rod accident (0.25)
- d. Protects.the core from the high flux. peaking (0.25) l resulting f rom one or more cropped rods.
1 REFERENCE Commonwealth Edison, Braidwood System Description, Ch. 6Ob, Terminal Perf ormance Objective
...(KA'S) 012OOOGOO4 l f
)
ANSWER 3.13 (1.50) l
/
- a. No (0.5)
- b. No (0.5) is OFF (0,5)
Control wer DN li ht a t e sese</h Aq%po% gk-t PR &&
c.
NOTE: n drawer /
REFERENCE 6 07 Commonwealth Edison, Braidwood System Description, Ch. 33 j' 2 Terminal Perf ormance Objective 8.
015000A201 ...(KA'S) !
l l
esas l
- . .e2I__INSISWdENIE_0ND_GQNISQLE PAGE 43 ANSWERS.-- BYRON 1 -87/05/OI-BITTERS f
. ANSWER 3.14 (2.50) ,
- a. NO -(0.5)
- b. NO: (0.5)
- c. 1. INSTRUMENT POWER ON ' light on IR drawer is OFF (0.5)
- 2. LOSS OF DETECTOR VOLTAGE' light on IR drawer is W!5 Ot)-eFF- (0.5)
I r/2pf7 3. IR HIGH VOLT FAILURE annunciator at MCB (0.5)
REFERENCE '
. Commonwealth Edison, Braidwood System Description, Ch. 32,
'. Terminal Performance Objective 6. .
015000A201- ...(KA'S)
ANSWER. 3.15 -(1.00)
Any 2 of the f ollowing 3 (0.5 each)
- a. LEVEL TRIP BYPASS lamp on the drawer lights.
- b. OPERATION SELECTOR switch is enabled.
- c. (118 VAC) signal is provided to prevent a reactor -
trip.
REFERENCE Commonwealth Edison, Braidwood System Description, Ch 31,
- I Terminal Performance Objective 5 015000A403 015000K405 015000K406 ...(KA'S).
i ANSWER 3.16 (2.00) b
- 1. Certain reactor trips coincident with a f ailure of auto bus transfer could result in a loss of RCP flow. But, if pumps were not lost until after a time delay (when the {
generator trips), the loss of flow would not be as serious !
I because the reactor would have been shutdown f or the duration of the time delay. (1.0)
- 2. Pump overspeed can occur on a major LOCA. This can result in a missile hazard. The chances of getting an RCP ;
overspeed can be minimi:ed by loci:ing in -the pump speed at 60 H: frequency until the generetor trips. (1.0)
REFERENCE.
Commonwealth Edison, Braidwood System Description, Ch. 60b Terminal Performance Objective 6
PAGE 44 J __INSIBUdEUIE_0ND_GQUIB9L@
ANSWERS -- BYRON 1 -87/05/d4-BITTER,S OOOOO7KO10 OO2OOOK514~ ...(KA'5) 4 4
4%
==
8
-.--Y
V, VA2__BBQGEDUBEE_:_NQBd86t_8EN9Bd8Li_EdEBQEUGY_8UD PAGE' 40
^%. B8D1QL991G66_G9NIBRL ,
' ANSWERS - . BYRON 1 -87/05/04-BITTER,S ANSWER. 4.01 (1.00)
- a. 40.8 (0.25 each)
- b. 17.0.
c.- 81.4
- d. 78.1-NOTE: rounding.to.the nearest whole number is acceptable for full credit.
,fREFERENCE Precautions, Limitations and Setpoints f or NSSS, BNPS Units-1 .
and.2, REV 7 Sept 1986.
035010K402 ...(KA'S) .
I ANSWER 4.02 (2.00)-
- 1. .Verif y reactor trip
- 2. Verify turbine trip ,
- 3. Check AF pumps running
- 4. Initiate emergency boration of RCS ;
I REFERENCE l Commonwealth Edison, BNPS 1BFR-5.1 and Training Dept., j Procedure LPs and HOs,.BFRs, TPO 4 *
{
j OOOO29G010 ...(KA'S)
~'
1 ANSWER 4.03 (1.00) l a.- 1.25 rem /qtr (0.5) gd5cf j j
The limit can be raised t; ~ ---'mi=&--
- b. 'O.2? 7 pr:cid:d th:t the total lifetime whole body g exposure limit of 5EN-183 is not exceeded (... and - - -
an NRC Form 4 (or Equivalent) is on file R+rt)
@9 8/8h REFERENCE Commonwealth Edison, Byron Training Dept., Procedure l esson Plans, Module-BRPs, Terminal Performance Objective 6.
1940016'103
- ...(KA'S) l w._ .
F .
'Si__EBgGEDUBEE_ _NQBdebu_6ENQBdebt_EdEBEENGX_8ND PAGE 46
. ' BOD 196901GOL_GQUIB96 4 ANSWERS -- BYRON'1 -87/05/04-BITTER,5 s.
' ANSWER 4.04 (1.50) a.. 3,2
- b. 2,1
- c. 2,1 (.25 each answer)
REFERENCE BNPS, Technical Specifications, Table 3.3-1 p. 3/4 3-2 015000K101 015000K301 ...(KA'S) 1 ANSWER 4.05 (2.00)
- 1. Two (2) successive starts are allowed provided the motor is initially started from ambient temperatures. (0.5)
- 2. Subsequent starts may be made provided that the motor has run for at least 20 minutes (.5) or provided:
that the motor has been idle *for at least 45 minutoc. (.5)
- 3. RCP starts should not everage more than six (6) per day over the life of the motor. (0.5)
REFERENCE BNPS, BOP RC-1 and Commonwealth Edison, Braidwood System -
Description, Ch. 13, Terminal Performance Objectives 9, 10.
003000G010 OO3OOOK614 ...(KA'S)
ANSWER 4.06 (1.00) 1.- Containment pressure greater than 5 psig (0.4) or (0.2)
- 2. . Containment radiation greater than 10 exp 5 R/hr (0.4)
REFERENCE 1BEP ES-0.2 and BNPS, Byron Training Dept., Procedure LPs and HOs, BEP-O series, TPO 1 1030004101 ...(KA'C)
PAGE 47
.Th2__EBQGEDQ8EE_:_NQBd8Ls_GENQBdeksEDEBEENQY_6ND
. ., lB6D1QLQQ1GOL_GQNISQL . .
-87/05/04-BITTER,S
' ANSWERS -- BYRON 1 ANSWER 4.07 (2.50)
CC water to RCP lost (affected pump only) (0.4) i 1.
) OR (0.15)
L 2. ~ CNMT Phase B actuated (0.4)
OR (0.15) l 3. ALL of the f ollowing conditions e::ist:
-a) Controlled RCS cooldown NOT in progress (0.4)
AND (0.1) b). RCS pressure < 1370 PSIG (1670 psig-for adverse CNMT) (0.4) .
AND (0.1) c) CENT CHG pump (s) >2OO gpm or SI pump (s) POSITIVE FLOW (0.4)
REFERENCE BNPS, Byron Training Dept., Procedure LPs and HOs, BEP-O Ssries, TPO 3 000009G010 OOOOO9G011 OOOOO9G012 ...(KA'S)
OOOOO9 GOO 7 0
ANSWER 4.08 (2.00)
- a. Initiate boration (0.5) b.- 1. Initiate emergency boration (100 ppm) (0.5) *
- 2. Insert all control banks (0.5)
- 3. (Inform SE)
The required SDM may:not be met. (T heref or e, to c.
ensure adequate SDM, the rods are inserted, and i a boration is commenced.) (0.5) 'l NOTE: answer in parenthesis not required f or f ull credit {
il REFERENCE Commonwealth Edison, BNPS, BGP-100-2 step E and Byron l Training Department, Procedure LPs and-HOs for BGP' s , j BGP-100-2, Training Perf or mance Objective Sa. l OO1000A204 ...(KA'S) l i
1 1
1 l
PAGE 48 da.__E89EEDUBEE 2 89Bd6bs._6EUQ806ba._EdgB@gNQY_0dp
,. 60D196991G66_G9dI696 . .
-87/05/04-BITTER,S
- ANSWERS -- BYRON 1 ANSWER 4.09 (2.50)
- a. 1. Reset (both trains) SI (0.5 for each of the
'four in the correct sequence.)
- 2. Cycle the Rx trip breakers
- 3. Reset FW isolation l 4. Reset FW isolation auxiliary relays Drain the S/G's below the P-14 setpoint (0. 5 )
b.
REFERENCE Commonwealth Edison, BGP-100-1, step 16, Byron Training 4
Dept, Procedure LPs and HOs for BGPs, BGP-100-1, T.P.O.
059000A411 059000K103 059000K105 ...(KA'S)
ANSWER 4.10 (1.50)
- a. A TYPE 1 RWP is required f or ALL routine access (.25) or work in radiologically controlled areas (.25) where personnel are NOT expected to exceed a whole body dose (.25) of 50 mrem / day. (.25)
- b. A TYPE 2 is valid for the length of the job. (0.5) .
REFERENCE Commonwealth Edison, Byron Training Department Procedure Lesson Plans, BNPS Training Document BRP 1000-A1 rev 3 -
Radiation Work Permit TYPE 1 AND Radiation Work Permit TYPE 2, Terminal Performance Objective 4 194001K104 ...(KA'S)
ANSWER 4.11 (1.00)
- 1. pressuri::er PORV (1.0)
REFERENCE Commonwealth Edison, BNPS 1BEP ES-1.1 and Byron Training Dept., Procedure LPs and HOs, BEPs, TPOs 18<5.
010000A201 010000A202 ...(KA'S)
c . 'eset *v jel__EB9GEDWBEE_:_N9Bd862_9EN9BdBLtEdEBQENGY_eUD BBD19690lGBL_G9dIB96- . .
j A. -87/05/04-BITTER,5 ,
i ANSWERS --_ BYRON 1 I
4.12 (1.00)cs4 g ANSWER Emergency borate 100 ppm GeG64cr each control rod not fully incerted. 40.237-h,$$ =!
REFERENCE BNSP 1BEP ES-0.1 000005A203 ...(KA'S)
E!
ANSWER 4.13 (1.00)
(0.25) p
- a. 1. Byron Status Trees (BSTs) (BEP) steps (0.25)- ;
- 2. . Byron Emergency Procedure Loss of All AC Power (0.5) b.
l l4 Commonwealth Edison, Byron Training Dept., Procedure Lesson REFERENCE i
Plans, Modules BFRs.. Terminal Performance Objective 1 !
194001A102 ...(KA'S)
ANSWER. 4.14 (1.00) i
]
- 1. If WS header pressure is less than 80 psig and trip the thereactor. (.25)
)
standby WS pump cannot be started,(.25) 2.
If all WS pumps require shutdown, trip the reactor. (0.5)
I I
Commonwealth Edison, BNPS, 1 BOA Section 5 and Byron Training REFERENCE Sec Series, TPO 3 f Dept., Procedure Los and HOs for BOAS, I 012OOOG014 ...(KA'S) f 012OOOG013 i ANSWER 4.15 (1.00)
- a. 66 percent (0.5) l
- b. 50 percent (0.5) l REFERENCE l BNPS, 1BFR-C.1 I i
hp
s.
r.
m A__ESQGEDWBEE_:_NQBdeLi_eENQBdeL,_EdEBGENGy._0ND PAGEl 50 Q; : BOD 1QLQQ1GOL_GQUISQL: ,
1
-87/05/04-BITTER,5
.c-ANSWERS --JBYRON.1 ANSWER- 4.16. (2.00)
- a. 1)' Keff of 0.95 or less (0.,4)
OR (0. 2) .
2). Boron concentration-of greater than.or equal to ,
I 1 2OOOppm (0.4)
_b. . A 2150 ' ppm maximum boron concentration limit is imposed 1
-(0.5) to prevent unnecessary reduction in the response of the SR neutron flux monitors. (0.5) 4.
REFERENCE BN@S, Byron. Training Dept, Procedure LPs and HOs, BGP 100-6, TPOs 1,3 and 1BGP 100-6 034000 GOO 1 ...(KA'S) 1 -ANSWER 4.17 (1.50) .
a
- 1. both PZR.PORVs (0.3)
- 2. three letdown orifice isolation valves (0.45) -
- 3. two letdown line isolation valves (0.30)
- 4. 'three excess letdown isolation valves (0.45)
REFERENCE-
ANSWER 4.18 (1.50)
- a. qualitative assessment of channel behavior during operation by observation. (0.5)
- b. dif f erence in normalized flux signals between the top and bottom halves of a two section excore nuclear detector. (0.5)
- c. ratio (.05) of maximum upper (lower) encore detector calibrated output (.2) to the average of the upper (l ower )
excore detector calibrated outputs, (.2) which ever is greater. (.05)
REFERENCE BNPS, TECHNICAL SPECIFICATIONS 015000A104 015000G014 015000G15 015020K508 ...(KA'S)
n ..:
uquw wev COPY U. S.-NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _DXBgN_1_________________
REACTOR TYPE: _EWB-WEQ3________________.
DATE ADMINISTERED: _@ZZggf93________________
EXAMINER: _LENNABTZ 3 _J3__________,__
CANDIDATE: ____________________,.,____
INSIBucIJgus_Ig_gAUDJp61Ei L* 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 areLindicated in parentheses after the questi on. The passing grade requires at least 70% in each category and a final grade of at least 50%. Examination papers will be picked up.six (6) hours af ter -
the examination starts.
's. OF
' CATEGORY % OF. CANDIDATE'S CATEGORY y
_..Y669L _19I06 ___SCOBE __ YGLUL_ _ ___ _ __ _ __ __ _ _CeIE GgB L._ _ _. ,_. .. _._
_2Dt99__ 2Es99 _ _ _ _ _ . _ _ . . _______ O. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2Et99__._2Ergg ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2Es99__ _2ErgQ ___________ ________ 7. PROCEDUFjES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL f _.______ B. ADMIN)STRATIVE PROCEDURES,
_2E.t99__ _2Mx99 _______ ___
CONDITIONS, AND LIMITATIONS
'192:09.._ __.,________
__.,_____% Totals Final Grade All' work done on this examination is my own. I have neither givcn nor received aid.
Candidate's Signature i
i
- .1 1 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the admniintration of this examination the f ollowing rules. apply
- l. l Cheating on the examination means~an automatic denial af your-application and could result in.more severe penalties.
- 2. - Restroom1 trips are to be limited and onl y .one candi date at a tims 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 gn1_y to f acilitate legible reproductions.
4.. ' Print your name in the b' lank provided on the cover sheet of the enemination..
- 5. Fill in'the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print.your-name in the upper right-hand corner of the first page of ggch
.section of the answer sheet.
' O. Consecutively number each answer sheet, write "End of Cat egor y __" as appropriate, start each category on a,new page, write only 90 . gag gido of the paper, and write "Last Page" on the last answer sheet.
- 9. Nuthber each answer as to category and number, for example, 1.4 3 6. 3.
tio. Skip at least Mjrge lines betwecn each answer.
- 11. . Separate answer sheets from pad and place finished answer sneets face down on your denk or table.
- 12. Use abbreviations only if -they are commonly used in f acili ty li tetutgr_e.
- 13. The-point value for each question is indicated in parentheses after the '
question and can be used as a guice f or the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not. j l
- 15. Partial credit may be given. Therefore, ANSWER ALL PAR 15 OF THE J l
DUESTION AND DO NOT LEAVE ANY ANSWER ELANK. '
- 16. If parts of the examination are not clear as to intent, ask questions of
.the egggLner onl y.
- 17. You must sign the statement on the cover cheet that indicates that tne wor k is your caci and you have not received or been given assistance in completing the exami nation. This must be done after the examination has been completed.
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- 18. When you complete your examination, you shall:
- a. Assemble your examination au follows:
(1) Exam questions on top.
( 2 ') Exam aids - fiqures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the e:: amination 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 examiner. If after leaving, you are found in thi's area while the examination is still in progress, your license may be denied or revoked.
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. EQUATION SHEET f = ma ,
y = s/t Cycle efficiency = (Network out)/(Energyin) 2 .
w = ag s = V ,t + 1/2 at E
E = ac t ,-E t KE = 1/2 av ,,gyf , y )fg 3 , g, g , g PE = agn Vf = V, + at w = e/t a = an2/t.1/2 = 0.693/t1/2 2 t jj g = @ m M W = v aP.
A= 30, [(g1j2) + (t,)] ,
AE = 931 m ,,y avA ' "E*
I=If q = aCpat 6 = UAaT I*I'* o g p, , gg I = 1,10 TYL = 1.3/u P = P,10 sur(t) HVL = -0.693/u p = P et/T ,
SUR = 26.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26e/t* + (s - p)T CR j (1 - K ,ffj) = Gg O - keff2)
T = ( **/s ) + [(s - e V is] M = 1/(1 - K,ff) = CR j/CR, T = s/(s - s) M = (1 - K,ff,)/(1 - K,ffj)
T = (s - e)/(Ie) ,
SDM = (1 - K,ff)/K,ff -
e = (K,ff-1)/K,ff = 4Keffeff s' = 10 seconds I = 0.1 seconds *I e = [(t*/(T K,ff)] + [s,ff /(1 + IT))-
Id Id i i =2 ,2 gd 2
P = (34V)/(3 x 1010) Idjj 22 2 E = eN R/hr = (0.5 CE)/d (meters) l R/hr = 6 CE/d2 (feet) ,
i Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010dps 3 kg = 2.21 lba 1 ga[. = 3.78 liters 1 ft" = 7.48 gal. 1 hp = 2.54 x 1 8tu/hr Density = 62.4 lbg/ft3 1 m = 3.41 x 1 8tu/hr Density = 1 gm/cW lin = 2.54 cm Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.
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-DUESTION 5.01 (2.50)
- a. What is.the subcooling margin (SCM) of the plant if the f ollowing j conditions exist? (1.0)
Th- = 5BO F= Ppzr=2185 psig Tave= 550 F .
Pug = B50 psig ;
~
Tc = 520 F
. b. 'If plant l power ic raised from 50 to-100%, how wi'11-SCM cnange
'(increase, ' decrease, . stay the same) ? Why? (0.75)
~
c .' Which one of'the following would give a smaller SCM7 Assume identical RCS pressures. Briefly explain why. (0.75)
- 1. . SCM . duri ng . a controlled natural circulati on cooldown ~ imniedi ately -
following a reactor trip from loss of flow. .;
- 2. . SCM from continued operation at 5% power..
- 3. SCM produced when all RCP's are operited at normal no-l oad temperature after extended, shutdown.
QUESTION 5.02 (1.00)
What is the reason that Tech Specs allow up to two hours of operation with 1 a-Quadrant Power Tilt Ratio in excess of 1.02 but less than~1.097 j (Assume Rx power at 80%)
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1 QUESTION 5.03 (1.00)
The reactor is operating at 100% power with all rods out, near DOL with equilibrium Xenon conditions when power is to be reduced to 50%. The operator' observed that-AFD is -7, and within its band, and deciden to
. lower power by borating, leaving rods in the ARO position. Actual Tavg.
f ollows programmed Tavg. Describe the change that will occur in AFD and why it occurs, prior to changes in Xenon having a noticeable effect.
OtJESTION 5.04 (3.00)
- a. Define FO(2) ac per Technical Sp ec i i i c at i on,s. (1.0) I
- b. Liut the four conditions that ensure that FQ(Z) is matntained within limits. (2.0)
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THEF MOD YN Af11 CS PAGE 3 QUESTION 5.05 (1.00)-
(Fill' in the' Blank) cDuring a. Xenon-free reactor startup, critical data was inadvertently taken two ducades below the required Intermediate Range (IR) . level. Assuming RCS temperatures and' boron concentrations were'the same, the critical rod position taken at the proper IR level _________ the critical rod position taken two decades below the proper IR level,
- a. Is Less Than ,
- b. Is The Same As c.. Is Greater Than
- d. Cannot be Compared To
-QUESTION 5.06 (2.00) ,
s Expl ain why. a dropped rod, with.no Rx trip, is worth approximately 200 pcm and a stuck rod is worth 1000 pcm even though the same rod could be considered in both cases. (Explanation of the effects of a stuck rod and
, dropped. rod required.)
DUESTION 5.07 (1.00)
Which one of the f ollowing statements concerning the power def ect is correct?
- a. The power defect is the difference between the. measured power coefficient and the predicted power coefficient.
- b. .The power defect increases the rod worth requirements necessary .
to maintain the desired shutdown margin following a reactor trip.
- c. Lecause of the higher boron concentration, the power defect is more. negative at beginning of core life. -
- d. The power defect necettitates the use of a ramped Tavg program ,
to, maintain an adequate Reactor Coolant System subcooling margin. !
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QUESTION 5.08 (1.00)
TRUE-'or FALSE 7
- a. As.Keff approaches unity, a smaller change in neutron level will result for identical. changes in Keff.
- b. As Koff approaches unity, a longer period of time is required to reach the equilibrium neutron level'for identical changes in Keff.
Q.
QUESTION 5.09 (2.00)
IndicateLon your answer sheet whether the following statements are TRUE or- l FALSE:
- a. Pump shut off head is a term which could be used to ! describe a. l
' centrifugal pump operating with a shut discharge valve.
b.. -Latent heat'of vaporization is another term f or la' tent heat of fusion. ;
- c. ~ A law applying to centrif ugal purop s in closed loop systems states that power required is directly proportional to the square of the pump speed.
- d. For-two IDENTICAL centrifugal pumps running at DIFFERENT speeds in IDENTICAL. systems, the faster running pump requires a greater NPSH.
QUESTION 5.10 (1.00)
Choose the ccrrect phrase to correctly complete the sentence.
As the core agen from BOL to EOL, the ratio of PU-239' atoms to U-235 atoms ;
increases. This changing ratio causes the ...
- a. reactor period to decrease.
b.- -void coefficient to become less negative.
7:. moderator tempereture coefficient to become less negative.
- d. delayed neutron fraction to increase.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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- QUEr y10N , . 5.1' 1 - ( 1. 00) -
Which :o'na Lof the f ollowing statements concerning Xenon-135 production and renioVal' is correct?.
a.- At: full power, equilibrium conditions, about half : of the Xenon is LproducedibyfIodine decay.and.the.other half.is produced as direct k ii ssi on prodLitt.
- b. ; F611owing ' a reactor trip from equilibrium conditions,. Xenon peaks because delayed, neutron precursors. continue to decsy to Xenon-while neutron absorption' (burnout) has ceased.
Xenon. production and removal increases linearly as p'ower l evel
~
ac. !
1 11ncreases;Ji.e., the:value of 100% equilibrium Xenon is twice that.
of 50%, equilibrium Xenon.,
'd. At low power. levels, Xenon-decay-is'the major removal method. "At. i
'high power level s, . burnout is the major removal method.
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QUESTION U.12 (2.00) s 4
-Answer the-following questions concerning coefficients of reactivit y, '
- a. How and.why doesithe value of MTC change as temperature is increated in an over' moderated core? (0.75) i 'l
- b. Which initial t'emperature, 58 F or 557-F, will give the largest'-
ctiange in the magnitude of MTC as boron concentr ation is lowered-from-1000 to 500 ppm? BRIEFLY EXPLAIN'WHY. (0.75)
- c. . WHY does power defect become more negative as the cure ages? (0.5)
OUESTION 5.13 (2.50)
- a. , Drittl e f r acture of your reactor's pressure verssel can occur atJutresses.well below its yield stress. TWO> conditions must.be present besides high yield stress. What.are these TWO conditions? !
(1.0) , ;
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'b. .How do heatup/cooldown rate limits on toe reactor coolant i
system reduce-the probability of brittle fracture? (0.5)
- c. . Why does the concern about brittle f racture of the reactor pressure vessel increase as the plant ages? Include in your answer the specific mat eri al PROPERTY that is affected. (1.O'>
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'DUCSTION 5.14; (2.00). ,
- a. What_happens.to the power required by a motor driven centrifugal pump as'the dischargo valve is'trottled:in the shut direction? (1.0)
- b. :What are:TWD offects of. operating a motor driven centrifugal pump at- "
or beyond a' runout condition? (1.0) l DUESTION 5.15' (1.00)-
Howl do each of; the f oll owing ~ parameter s change (increase, decrease, or no chan'ge) 'if. one main steam isolation valve closes' with the' plant at 50%
1 load. . Assume all controls are in automatic and that no trip occurs.
a.. L.Affected l'oop steam generator-level (INITIAL change only)
'b.- . Unaffected' loop steam generator level (INITIAL' change only)
- c. Utiaf f ected ' loop steam generator,phesuure
- d. Unaf f ected loop cold leg temperature
. QUESTION 5.16' (1.00)
Indicate whether' the f ollowing will cause shutdown margin to incresce,
' decrease, or. remain the same.
l.- Control. rods withdrawn 10 steps'with no; change in turbine power or. baron. !
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- 2. Turb'ine power increased 10 percent with no change in boron or rod. position.
3.. Normal' power increaue f rom 50 to 60 perc ent with no boron change, rods-in automatic. i 2
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- 4. Doron concentration is increased 10 ppm with rods in automatic, J turbine power is constant. )
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(+**** END OF CATEGORY 05 *****)
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62__ELGNI_SYSIEbe_DE@l@N,_ggNIBg63_6Np_IN@lEUDEUIBIlgN FAGE 7 DUESTION 6.01 (2.00) .
Indicate whether the f ollowing situations will ARM ONLY, ARM AND ACTUATE, or HAVE NO EFFECT on the steam dump system. (Assume condenser interlock C-9 i s met. )
- a. Turbine trip, Tavg = 655 degrees F, steam dumps in Tavg mode.
- b. 50% power, 7.5%/ min. ramp decrease in turbine load for 3 minutes, Tavg , Tref by 6 degrees F, steam dumps in Tavg mode of operation. ,
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- c. Hot zero power, Tavg = 559 degrees F, steam dumps in STM PRESE mode ;
with 1092 psig set into the steam pressure controller.
- d. 50% power, 10% step load increase, Tavg < Tref by 5 degrees F, steam dumps in Tavg mode of operation.
QUESTION 6.02 (1.00)
What i s the F-14, Hi-Hi S/G level, setpoint for Units 1 and 2?
l CUESTION 6.03 (1.50)
- a. What is the Technical Specification level that shall be maintained in the Condensate Storage Tank? (0,5)
- b. What is the basis for the level asked for in part a. above? (1.0) s OUESTION 6.04 ('1. 00 )
Which one of the f ollowing f ucl handling tools is pneumatically opcrated?
- a. New fuel assembly handling tool.
- b. Thimble plug handling tool .
- c. RCCA change fixtur e.
- d. Durnable Poison rod accembly handling tool.
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QUESTION 6.05 ( 2. 00)' .
State whether the OP delta T setpoint will. INCREASE, DECREASE, or REMAIN THE SAME for each.of the-following' conditions.- (Consider each case separately.)
a.- Pressurizer.. spray' valve sticks open for 15 seconds. ( Assunie no. Rx trip.)-
- b. During power ascension.from 50% to 100% occurs.
c.. The reactor trips.from 75% power.
- d. Delta-flux at 100% power is reading +12.
s QUESTION- 6.06 ( .50).
TRUE or: FALSE 7 On a Containment Spray (CS) actuation signal , the CS eductor suction valve must go fully open'before:an auto start signal for the CS pump is enabled.
' QUEST 10N '6.07 (2.00) 1.
Permissive P-4 is actuated when a reactor' trip breaker and its associated bypass breaker-are both opened. List FOUR f unctions that P-4 perf orms.
QUESTION- 6.08 ( .50)
What pressurizer channel is used to monitor refueling cavity level during ref ueling' (Mode 6's operations?
OUESTION 6.09 (1.50) i List all signals used by PORV 456 and 4554 when placed in " Arneed Col d j Overpressure" position.
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OUESTION 6.10 (1.00) .
- a. What is the lowest minimum flow achievable by the PD pump if it is in auto on the PDP speed controller? (.50)
- b. What is the lowest minimum flow achievable by one charging pump if its controller is in auto? (.50)
QUESTION 6.11 (2.00)
,a. If the power swi tch f or the Fire Hazards Control Panel is "ON",
and the normal / fire switch is in the fire position, list EIGHT indications that will be available at this panel. (1.6)
- b. What, if any, indications will.be available at the Fire Hazards Control panel if the power switch is in the "OFF" position? (0.4)
QUESTION 6.12 (2.00)
The SRD, was making hic plant tour ano noticed on the gas radwaste panel the waste gas discharge control valve OGWO14 (W) (white light) valve light indicator was illuminated. What condition (s) would cause the (W) (white light) to be illuminsted on the gas discharge control valve?
OUEETION 6.13 (2.00)
List 2 indications from the Liquid Radwaste Panel or the Control Room an SRO can use to determine when a liquid radioactive water release is in progress and if its cetpoint was exceeded. (Enclude RM-11).
DUESTION 6.14 (2.00)
Describe how a diesel stator electrical fault is detected by the di esel when the diesel is running, but the output breaker is not shut.
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st_ ELONI_S.YSIEMS_DEgigh _gDNIB96_BND._IUSIBU!$ENIBilgd PAGE 10 i QUESTION 6.15 (2.00)
Lict all of the radic. tion monitoring instrumentation on the RM 23 which are contained in lechnical Specifications. (List only one train of redundent inctrumentation.)
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I QUESTION 6.16 (2.00)
The Reactor Operator noted a +2deg.F tempcrature error on the Tave/Tr ef 1
- chart recorder and a Odeg.F temperature error on the Tref /Tave meter on the BOP panel. What are the inputs to each of these instruments?
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(***** END OF CATEGORY 06 *****)
~ 'Z,.__EBQCEDUBEQ_ _NQBd@La_@DUQBd@61_EUESQEggy_8NQ PAGE 11 B6Dl96991C66_G991896
. QUESTION 7,01 (1.00)
In BEP-O RCP Tr ip Cri t eri a, Unit i recuires RCS pressure of < 1370 psig for normal containment, and < 1670 psig for adverse containment condi tions, while Unit 2 has onl y the < 1370 psig requirement. Why doesn't Unit 2 have the adverse containment RCP trip. criteria requirement?
QUESTION 7.02 (2.00)
- a. In accordance with a CAUTION in BOP DG-11, " Diesel Generator Startup," there are TWO actions which must be performed if a DG must be run and a rocker arm failure is suspected. What are these two action's? (1.5)
(Fill in the Blanks)
- b. If any cylinder exhaust temperature decreases by _______ degrees F ,
or more over a _________ minute.stime period compared to the other I cylinder exheust temperatures, then a rocker arm f ailure should be susp e c t eri. (0.5)
I DUESTION 7.03 (1.00) l You are the SRO and have entered BEP-1, and the SCRE gives you the following information concerning the status trees:
- 1. Core Cooling - YELLOW (go to BFR C.3)
- 2. Integrity - ORANGE (go to BFR P.1) {
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- 3. Heat Sink - YELLOW (go to DFR H.5) 4.- Containment - ORANGE (go to BFR Z.2)
D. All others are GREEN i
r Which of the above listed procedures woul d you tr ansition to in accordance with the rulou on status tree implementation?
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'OUESTION= 7.04- (2.00)~
PEP-3,. " Steam Generator Tube Rupture," Step-4b. directs'you to control f eedi flow ?to maintain narrow range level in-the ruptured S/G between 4%^
and 66% (34%-50% for adverse ~ containment). What is the basis f or controlling f ued ' flow to maintain. ruptured S/G level greater than'4%
but'less than 66%7- i l
QUESTION' 7.05 -(1.00)
~
'In the steam generator. tube rupture-(BEP-3): procedure,.a' step specifies to q stop RCS cooldown when the average of. ten highest' core exit TC's _is less !
than required. temperature. For what specific accident does'this )
-i
' temperature restriction guard against?
' QUESTION 7.06 (2.00) ,
TRUE or FALSE 7-The f ollowing are examples of when intentional entr y into an. action staten,ent 'may" be permi ssi ble.
- a. ,In order to avoid an unnecessary, continuing degradation of equipment ,
- b. In order to perform'a surveillance
- c. In order to perf orm routine, but not necessary, preventative i maintenance !
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-d. In order to perf orm necessary preventative maintenance
.i OLIESTION 7.07 (1.00) i
.TRUE-or FALSE 7 The. Fuel' Handling Enhaust Filter plenumc are not required to be operable
'unt11' irradiated fuel is in the storage pool.
QUESTION 7.00 (1.00)
'There is new fuel, but no irradi ated 4 ucl in the storage pool. Esth f uel building . isolation radioactivity hipn and criticality channel s are
' i noper abl e. What' action must be met 2n order to meet Tec hni c a) 1 Specification f or : both' channels inoperable? )l
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1 QUESTION 7.09- (1.00)
TRUE or FALSE'7, In the RCS overpressure protection system Tech Specs, the action statement does'not, apply if'two PORV's are operable or if'two RH suction-relief valves are operable.
~OUESTION- 7.10~ ( l '. 00 )"
. ; When L i s; Reactor Vessel Level' Indication System (RVLIS) considered
" inoperable" per a surveillance?.
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QUESTION 7.11- (1.50)
- Y6u are performing BOA ROD-4, " Dropped RoN Recove.y," procedure and the' Station Nuclear Engineer tells you to maintain reactor power less than :the
. trani mum f uel pr econditioned limit. What i s the maximum f uel
. preconditioned limit' defined as? (Assume this is during.the i ni t i a11. power ascension af ter. a r ef ueling f uel handling outage.)
QUESTION 7.12 (2.00)
- a. List FIVE conditions that would require you to emergency borate in accordance with BOA PRI-24 " Emergency Boration," procedure.
DUESTION 7.13' (2.00)
List'FIVE annunciators and/or symptoms that would indicate a CC system I mal f uncti on as described in BOA PRI-6, " Component Cooling Malfunction,"
procedure.
OLIESTION 7.14 (1.50)
The operator ac ti on summar y f or BEP-O series procedures requires you to trip.the.RCP's,if.CC water to the RCP is lost, or' containment Phase B is actuated or'all of three conditions exist. What are these THREE
- condi ti on s?,- (Include setpoints for normal and adverse containment condi ti ons , . as' appli cabl e. )
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DUESTION 7.15 (2.00)
~
.In. Step 2.of BEP-0, "Rx" Trip:or Saiety Injection," the procedure has you verify turbine trip. 'If the. turbine did not trip as expected, what are the FOUR possible. actions that .could be taken, in accordance with l.1 BEP-0,.to trip /stop the turbine?-
QUESTION 7.16 (1.00)
During execution of BCA-0.0,'" Loss.of.All AC Power,":the. procedure cautions you not to decrease S/G pressure to less than 160 psig. What is thefreason.for this' limit?
t OUESTION' 7.17 . ( 1. 00 ) -
TRUE or FALSE 7 .
.If you have an-ORANGE on.BST-2 ( c or e,. c ool i n g ) during execution of'BCA-Oio,
" Loss-of All AC Power,".you should .first check the other status trees for RED' conditions, and.if no RED condi ti ons exi st ,- IMMEDI ATELY implement the -
CRANGE path procedure.
GUESTION' 7 . 1'8 (1.00)
In BFR-H.2, " Response to Steam Generator Overpressure," the procedur e refers ~to "affected" S/G's. What constitutes an affected S/G in
- accordance with'this procedure?
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h__GDU191SIBGIIVE_EBp,CEDUBEL _CpyplIJpgL _6Np_L.ldlIGIJpyg PAGE 15 QUESTION 8.01 (1.00) 1 RUE or FALSE 7
. Leakage through a FORV i s c. nsidered identified leakage as per lechnical Specifications.
)- '
l QUES 110N 8.02 (1.00) l An operator notes that the It<C For eman is using hi s personal timex watch to conduct the reading and voltage surveillances on the diesel. Explain why thi s timepiece is/is not acceptable por technical cpecifications.
QUESTION 8.03 (1.00)
TRUE or FALSE 7 Inoperable equipment, once repairs are completed, can be returned to service to perform the operability test while still in a technical j I
specification action statement.
- OUESTION 8.C4 (1.00) j In Technical Epecifications, there are Unit 1 tables which contain equipment located on the Unit 2 side of'the station. Is this equipment t'equired for Unit 1 operation? ( Ye s / No'7 ) ]
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OUESTION 8.05 (2.00)
Technical Specifications require two safety loops of component cooling water systems be operable i n modes 1, 2, 3, 4. Mark with an X on the ;
attached draw 2ng, (Figure B-1), all the c on.p on en t s , pipes, val ves. , which woulo constitute one safety loop. (Select either Train A or Train B)
DUESTION B.06 (2.00)
List the two conditions in accordance W2th IJAP 100-13, " Guidelines for Performance cf Independent Verifications of Proper Equipment AlignmenL,"
in which the f unctional checks and status moni toring .would be acceptable for performing an independent verification.
(**++* CATEGORY OF CONTINUED UN NEX1 PAGE ++*++)
7 e,__8Db1NIDIBBIIYE_EB9GE99BES4_G9ND1119NE4_6ND_6101IGII9NS PAGE~ 16 QUESTION 8.07 '(1.50) .
i What three actions must be taken when the SCRE/ Control' Room Supervisor decides that a Unit NSO sha11' leave the "At the Controls Area," when his reactor 'i s stable and under control, to help on the other unit.
j QUESTION- 8.08- (1.00). 3
'TRUE.or FALSE 7 *
' Independent verification shall always be' performed " apart in time" at all 4 times from the initial activity or status review. .
1 1
QUESTION' B. 09 - (1.50) f l
' List 3 conditions identified 'in BAP 300-1, " Conduct of Operat. ions," which '
I
.must be' met to use'the " pull.to lock" position for testing on'an active ECCS/ESF system. ,
t OUESTION G.10- (2.00)
Match the icilowing:
- 1. Master Out of Service Card (a) White with red letters and a red cross-hatched border i
- 2. -Qut of' Service Card (b) White.with blue lotters and a blue border
.3. Personnel Protection Card (c) Yellow with black l letters )
- 4. Caution Card (d) Used only on control panels so swi tch }
labeling lights won't )
be covered L. Miniatur Out of Service Card (e) White with red 3:7tters and a red border i l
i 4
I i
I i
i
.(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
'l
. _ _ . . ..a
1 ac__GDulN1@l86Ilyg_ESQQEQUSES t _QQNDlIlQUS2,@NQ_LidlI@IlQNS PAGE 17 I
OUESTION O.11 (1.00) -
What mechanical preventative measure, per BAP 330-2, " Temporary Alterations," shoulo be taken to prevent cross-contamination when !
connecting service air systems or the demineralized water systems to other syst ems? (Briefly exp.ain how this will prevent cross-contamination.) l l
DUESTION 8.12 (1.50) 1 Answer the f oll owi ng concerni ng BAP 330-2, " Temporary Alterations."
l
- a. ' When does the Shift Engineer routinely perform the review of the J temporary alteration logs f or new additions? (.5)
- b. When dbes the Operating Shif t Supervisor routinely perf orm this same review? (.5)
- c. When does the Operating Shift Supervisor routinely initiate the field verification of the actual temporary alteration location? (.5) r DUESTION B.13 (2.00)
Any individual that is permitted to enter a high r adi ation area and is exempt from issuance of an RWP must satisfy one of three technical l specification requirements concern 2ng. radiation monitoring. What are these three requirements?
DUESTION B.14 (1.00)
Two hour s into your shift, B-man, Mr. Goodwrench, broke his leg and had to be transported to the hospital, which 1 eft you with on1y four qualified Fi r e Bri gade members anti te. How long, in accordence wi th 'lechni ca3 Specifications, is it permiscible to operate with this shift crew 4 c omposi t i on? l OUESTION 8.15 (1.00)
(Fill i n the Blanks)
In accordance with Technical Specification 3.6.1.4, Primary containment internal pressure shall be maintained between._____ and _____ psig, l
(***** CATEGORY 08 CONTINUED ON NEX1 PAGE ****+)
r
, i
. ?
8t__0DdIUISIBBIlyg_PBgggpUBEQ3_GQUp1IJ9BS2_8Np_(Jd118IJgNS PAGE 18 1
. 1 OUESTION 8.16 (2.00) .
Technical Specification 3.4.6.2 allowc only 1 gpm total I reactor-to-secondary leakage through all steam generators not i sol ated f rom the RCS and 500 gallons per day through any one steam generator.
What are the bases for these limits? ,
i 1
i DUESTION B.17 (1.50) l 1
t If the Shift Engineer is incapacitated during an accident condition, with
. Rx at 20% power, who (title or position), if required, would act as ,
Station Group Director? List in order of preference. (Assume emergency !
has not been declared.)
QUESTION G.18 (1.00)
What two responsibilities of the Statipn Director may not be delegated?
OUESTION 8.19 ( .t . 00 )
TRUE or FALSE 7 If a " General Emer gency" is reclassified to a " Site Emergency," af ter the initial notifications were made for the General Emergency, Procedure ;
BZP 310-1, " Initial Notifications," must be repeated.
i J
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l l
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(++*** END OF CA1EGORY 08 *****)
(**+***+++++++ END OF E X A111 N AT ] DN + + * * * * * * + * * + + + + )
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. ANSWERS -- BYRON 1 -87/05/04-LENNARTZ, J. I 1
l ANSWER 5.01 (2.50)
- a. From the C-E Steam Tables:
Tsat for 2200 psia = 649.5 F (0.5) j SCM = Tsat-Th = 649.5-580 = 69.5 F (0.5)
- b. Decrease. (0.25) l Th increaser, as unit delta T increases with power. (0,5) l
- c. 1. (0.25) - Core delta T during natural circulation cooldown will approach full load delta T. Thot is greater than in the other 2 cases. (0.5) l i
REFERENCE ]
l l
PWR Fundamental s Tw> t Heat .Transf er , Chapter 7 i
! Steam Tables 193OOSK115 ...(KA'S) f- ANSWER 5.02 (1.00)
- Allows f or time to identify and correct a dropped or misaligned rod (1.0) l REFERENCE Byron Tech Spec Bases 3/4.2.4, pg. B 3/4 2-5 ,
! -193OO9K101 ...(KA'S)
ANSWER 5.03 (1.00) 1 1
Due to the greate' decrease in the temperature of the coolant e>:i ti ng the ]
core relative to the decrease of the inlet coolant (0.5). more positive i reacti vity will be added in the upper cor e regi oni+ , resulting in a n. ore _
positive (less negative) AFD (0.5).
REFERENCE Wettir.ghouse Nuclear Tr ai ni n ci Op er at i ons , Chcpter 8 j 1930091::102 . . . ( t:: A ' S ) 1 1
l I
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l
Di__ISE0BY_0E_UWGLE68_E9 WEB _E68NI_QEEBSIlgNz_E6Q1QE3_@ND PAGE 20 ;
ISEBURDYNadlGE l ANSWERS -- BYRON 1 -87/05/04-LENNART2, J.
l l
1 ANSWER 5.04 (3.00)
- a. It is the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, )
allowing for manufacturing tolerances on f uel pellets and rods.
(Equivalent wording accepted for full credit.) (1.0)
- b. 1. Control rods in a single group move together with no 1 individual rod position differing ~by more than + 12 steps, !
indicated from the group demand position. (0.5) ;
- 2. Control rod groups are sequenced and overlapped. (0.5) ;
- 3. Control rod insertion limits are maintained. (0.5) 1
- 4. Axial power distribution is maintained within limits. (0.5) i l
REFERENCE liyron Tech Specs, pg. B 3/4 2-1 thru 2-4 j 193OO9K103 ...(KA'S)
ANSWER 5.05 (1.00) b.
REFERENCE i Westinghouse Fundamental s of Nuclear Reactor Physics, Chapters 5 and 7.
192OOSK110 ...(KA'S) i 1
ANSWER 5.06 (2.00)
\
t When a rod as stuck out with all other rods inserted, that rod " sees" a much higher flux than the rest of the core. Since rod worth is a j 1
function of the rel ative flux difference between the rod and the core averbge flux, the rod is worth about 1000 pcm which is much more than nor mal . (1.0)
If a rod is dropped, that rod depresses the flux in the ar ea i t is in relative to the rest of the core. Since rod worth is a function of the re)ative flux dif f er ente bes. ween the rod and the core aver age ilux , thc rod is only worth about 200 pcm. (1.0) {
d REFERENCE IP2 Reactor Theory, pg. 7-59, 60. ,
192OO5K105 ...(KA'S) /
)4 w
= - - - . ..... .___.__. ___........_... _. .....
E.__1UEQBY_9E_NQCLEGB_EQWES_E6@N1_QEgB911gNz_ELylD@3_9NQ FAGE 21 IUEB099YUOulCg ANSWERS -- BYRON 1 -87/05/04-LENNARTZ,-J.
l ANSWER 5.07 (1.00) l b.
REFERENCE 1 Westinghouse Reactor Core Control, Chapter 3 l 192OO4K113 ...(KA'S)
ANSWER 5.08 (1.00)
- a. False i
- b. True (0.5 pts each) 1 REFERENCE I l
I DCC Westinghouse Reactor Physics, pg. '
-I - 4 , 13-14 1 192OO2K108 ...(KA'S)
- i 1
l ANSWER 5.09 (2.00) ll
- a. True
- b. False i
- c. Falso 1
- d. True j (0.5 pts each) j 1
REFERENCE l Westinghouse Thermal-Hydraulic Principles, Chapter 10 191004K112 191004K115 ...(KA'S)
ANSWER 5.10 (1.00>
a.
REFERENCE Westinghouse Reactor Core Control , Chapter 2 192OO3K107 ...(KA'S) 6 l
, _m_,__-. -._ _ ________________ ___
Ut__IBEQBY_9E_UWGLEGB_E99E6_E66MI_ GEE 6811gNt _E6ylD@i_GNQ PAGE 22 l
.1UE6dgDYN901Q@
ANSWERS -- BYRON 1 -87/05/04-LENNARTZ,.J.
ANSWER 5.11 (1.00) d.
REFERENCE Westingbotse Reactor Physics, pg. 1-5.63-76.
1 9 2 O O 6 K 1 0:., 192OO6K104 ...(KA'S)
ANSWER 5.12 (2.00)
- a. It becomes less positi ve. (0.25) Increasing temperatur e f orces more bbrated water out of the core. This reduces thermal absorption and adds positive reactivity to the core. (0.5)
- b. 557 F (0.25). The change in density at higher temperature is much greater than at l ower temperature. So more boron atoms enter or leave the core at higher temperatures which causes a larger reactivity change for a given temperature change (0,5).
c Lowering barcn concentration over life makes MTC more negative which causes its contribution to power defect to increase. Hence, power defect increases. (0.5)
REFERENCE
- a. Westinghouse Reactor Core Control, pg. 3-14
- b. 3-19, 3-20 .
- c. 3-2D 192OO4K108 194004K106 ...(KA'S)
ANSWER 5.13 (2.50) _
- a. 1. Pr esence of a flaw (or crack of sufficient cine). (0.5)
- 2. Low temperature. (or Temp below NDT) (0.5)
- b. Reduces the thermal stress. (Reduced DT across the RV wall reduces total / thermal / tensile stress.) (0.5)
- c. Neutron exposur e (integrated) (0.5) makes the material more brittle (raises NDT) (Reduces ductility) (0.5)
REFERENCE WEC-Thermal Hydraulic Principles and Applic, to the PWR II, pg. 13; bE-62 193010K101 193010K104 193010K105 ...(KA'S)
, , i I
Di__1HEQ8Y_QE_NyQ6E@B_EQWQB_E66NI_QEEBGllOUz_E6ylpQt_6NQ PAGE 23 IUE6dQQYN@dlGQ ANSWERS -- BYRON 1 ~B7/05/04-LENNARTZ, J. !
1 ANSWER 5.14 (2.00)
- a. Power decreases. ( 1. O )
- b. 1. Overload (overheat) of motor (trip of the motor breaker)
- 2. Cavitation
- 3. Pump. efficiency decreases
- 4. Overheating of pump bearings (Any 2 @ 0.5 each)
REFERENCE Theory and FF 191004K107 191004K112 ...(KA'S)
ANSWER 5.15 (1.00)
- a. Decrease i
- b. Increase
- c. Decrease
- d. Decrease (0.25 pts each)
REFERENCE Westinghouse Theory and FF 193OO5K103 ...(KA'S)
.i. ANSWER 5.16 (1.00) i
- 1. Remain the same
- 2. Remain the same
- 3. Remain the same
- 4. Incr easp (0.25 pts each) 1 i
REFERENCE f Westinghouse Theory and FF j i
192OO2K113 ...(KA'S)
6.__EL6MI_SYS.IEUS_QEQ1QN 4 1 _QQNIQQLi_6ND_lNQlBUdENI611QN PAGE 24 i
, ANSWERS'-- BYRON 1 -87/05/04-LENNARTZ, J.
ANSWER 6.01 (2.00)
- a. Arm only
- b. Arm and Actuate
- c. Arm and Actuate
- d. No effect REFERENCE System Description, Chapter 24, " Steam Dumps," pg. 7-9 041010K417 041020K411 041020K414 ...(KA'S)
. ANSWER 6.02 (1.00)
Unit 1 - 81.4% (G1)
Uni t 2 - 78.1%' (78) .
REFERENCE Byron Units 1 and 2 Differences, Rev. 4/1/86.
013OOOK101 ...(KA'S)
ANSWER 6.03 (1.50)
- a. At least 40%. (0.5)
- b. A minimum water level of 40% ensures that sufficient water (200,000 i gal.) is available to maintain the RCS at hot standby conditions (.25) for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> (.25) with steam discharge to the atmosphere (.25) concurrent with total l oss-of-of f si te power (.25).
REFERENCE Technical Specification 3.7.1.3 and Bases.
056000 GOO 5 ...(KA'S)
ANSWER 6.04 (1.00)
- c. ,
REFLRENCE Syctem. Description, Chapter 52, pp. 24-27 OOOO36EK2O ...(KA'S) i l
i l
6&__ELONI_SYSIEd@_DE@l@Nz_QgNIBQ63_6ND_lNSIEUNENI611QU PAGE 25
. ANSWERS --' BYRON 1 -87/05/04-LENNARTZ, J.
ANSWER 6.05 (2.00)
- a. Remain the Same '
- b. Decrease c.. Remain the Same
- d. Remain the Same (0.5 pts each - 2.0 total)
REFERENCE System Description, Chapter 12, pg. 34-35 012OOOK607 ...(KA'S) l ANSWER 6.06 ( .50) l i
True. (0.5)
REFERENCE ,
I System Description, Chapter 59, pg. 22 013OOOK101 ...(KA'S)
ANSWER 6.07 (2.00)
- 1. Actuates a turbine trip. !
- 2. Shuts main feedwater regulating velves if Tavg. is less than or equal to 564 degrees F..
- 3. Locks in circuit to prevent re-opening the MFW valves which were closed by either an SI actuation or a high-high S/G 1evel.
- 4. Provides a signal to the Saf ety injection block and reset logic circuit.
(0.5 pts each) i REFERENCE 1 System Description, Chapter 61, pp. 32-33 j 013OOOK115 013OOOK401 013OOOK412 ...(KA'S) I i
i ANSWER 6.08- ( .50)
Cold calibrated pressuriner channel (1LT-462) i l
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.,. . ANUWERS -- '1:YRON .1 -8 7 / Ob/ 04--L E NN ART Z , J.
' Re.f ERSCE .
Chapter 1 4 - 3'S 0020C ur::402 .. (K2'G) i l,
AN BJE R
. 6.O? (3.30)
(evdt. I t w 'hai l ect tcwperature Asct. I o.J colJ )en t proper atur e kide tange precLure .
W.b pts-cacb>
REFERENCE Chapt er .1ei-46, F ) pur e 14-ISC 010000K40S ...(KA'S)
ANCHER 6.10 ().00)
- e. 52 qpn ,.
- b. 5 2 cipin FEFERENCE Chapter '; be- G 3 Ch.4p t er 1 tia.- ' 4 004000K604 ... W4'b>
AWWER 6 11 ( '.: . 00 )
( f'.i s y 23 at 0. 2 pt e each)
- a. 1. klS W. R hot J' ea t einpe r a t urc
- 2. RC:b l46 R cu) d 1 :+c1 t E mper at ur e
- 3. PZh prceu (Ch-4SS) 4 P. Ti ,cve] ( Ch--4 59 )
Ti . S / t;: 14 p r.-e s ( Ch- t.1 1 )
- 6. B/G 14 lJ. R Jevm 7 3 'C iD pt 6 z:, u ( Ch--! 25 i G. 4/L 1 }. W. ii: 3 e s<f- i
- 9. 1. nut c e Rarcp ( T r,si n L. )
~
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(C.2 p + tr Ld. s t l e <
l.. 1. Uou r c.e h ar. q .~ ( Tr ni n P
- 2. l'u m+ - R ,.u ig u ( 3 i . i.e r nt. d a a ;. e Renou/ iTr ni n FO I
r i
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Ai4HWLh5 " IWRlal 1 -87/OU/04-LEWhART', /JJ .
F4:F'E P B ;C E .
Ct.epter 62, pg. 62-25 OE60003021 ...(UA'5) p,ri'!W:::R 6.12 (2.C0)
- 1. G.n, c i w c h arcm i n procir ess
- 2. ilj oh r adi i.3t i c i m ynal not rer etant REFEREllCE L4a nt te gat tii scharcic ca.itt ol val ve M.hernat i c 6 b - U - 4 0 . % - Gl" O fi 071000K609 ...(KA-S) 1 i
ANSWER 6.13 (2.00) .
l
- 1. l .otsc oi slow on Rcl eat e Tin n- Fl c w Rac or der or R el eas.e 1 e'il. i- l a .i liit et ra t:oi , .
L. Li qtti ti Rc.d w u,t e Uc..n t < n; Pane] 'l r e ub i o alar n wl 'l ; annunciato a4 tne r el c.a he sulpoiut hr* be on ter coudod .
".. Ga il ne ct 3 ve J i qua c; r ei c ac c' valse (OWi 353; and bypans valve ( i il l), c:'i t )
Anoicate 2. hut. .i 1
( fin y 2 at 1. 0 eat t. )
t i
REFEF,Et;CE l f mdweiste PaiUI , F <l1- 2 3 . DAR v000U9090E . . 4A'S?
OtG lF F L.13 ( 2 . '.s .~O
'ha ftali u i l cauce ,1 hi 9;. 4 ur r oi.i 41 v:. the t in c it t:ie neutral or .a i it.
t r e n.; r t ,r n er , ( n c u t r e- 1 ut uunci e vE: .r l ar ) . il Oi ut .i c h wiil acttste ch ?
neut t n1 y ot.nd #eJa. (i.0)
RLI LHD3 c,/ 2c E ti 0;u.. Ju 1 L, / M L- :t . -%
.s % ~U 6/20 i. - 3 .-4 6 ;> A 0560 'ON6M - ..(Kn'5) i i
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f.N SWE h 6. I 'd (2 CO?
- 1. Au:< . 10(19 Vent Sttck 16 d b Hrir qe Gac Morii tor -
2 '. Cont;a a nn ont Hi ah Raiige (TV<u 20,21 )
- 3. "1ni o S t c< atr 1 i ric- ( A.<0 2 2,23) 4c Contro) Room Out si de A.t r I n t at:0 . t PRi n J- 1,32,33,34)
D. Cont ai nnsunt Fuel 'r;ar.cn x nq :. n c. i d en t (AHO 11.32) 6, F ue! 61dy. F L 01 Hand.Ang ((iku ;55, b6)
( % 33 cech,'2 o +or.a.t1 ed NEF EfiENCli lecnnjceJ Epsccificetiour 0000626cX5 ...(KA-5)
ANdt )E R 6 . .:. 6 (2.00)
'le.m 1re4 cleart r ecc t ch'r (2 TR 432>-'aut.1 hi 'Ias e Arid T ur bi r ee I%)ul w Pr ot. cure (F1 305) ( 0. S c a. t..b > .
f r c * / T a v ia ti,i t er (2 11 412/O-- .;uct loe 7 av te and T .tr bi ne ' rq ul s. c I'r cw s tr e:
., i (I'l DOS) (C.D cacn!
REF EREINCE 6/20 E-2-4 51 t$ 3.i 6/20 !!~.'-4 031 MB l '/
001DOOA101 O 3 1 0 0 C I-il!. 2 . . . O Ji 'S)
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9 k
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(',NSWER 2.01 O.00)
Un > t. 2 -'RCL Wi de .anc1s p.rrture trancniittern h:4 ve lieen pl aced c,uts.i de o f.
- on t ai ntnea t (0.n) and there+ure ar e not a+4ected by adverso.conta
- nnon;..
c onc;i t ron s, 10.3).
REFEHENCE l 1;y r on Lin.i i u 1 cnci '2 Di s f er enr os., F&v. 4/3/06.
1v40016C10 . . . ( kA ' S)-
, a ANSWER 7.02 ( ;' . t' 0 )
1 l
- a. ' 1 '. livn i i cir crarikcan ci1 1avel ( . 3 7 '.,), untiI t he -f nel rack iE
- d i s ccion ec t ed ( . :U b) . .
-2. I mtnedi atel y con t ac t' unt: br ni c el unai ret e n a nce ( . 3M 'i to ' di s' ontm t.t i
duel ratk on ih+ n.,vr.;ps:1 cviiace- (.373).
u ., 10,0 de gr ean F ; 20 rna r ut i+ (0,23 pl4 oach)
RLiFEREhCd l'DP DE-11 3 pg. 4 064OS: 1,4 0 ( . '/ O c >tK%W.OO2 ... W.A'L)
(.N3UER "/ . O J. (i.00)
BFR P. J. iDrange .,n 1 riter.i t y )
4LI F,.} ?.fiNCE lty r un r1atur 'I s L t s:
.1 V k :.' 3 ) b 3 i .3 .., ( LS ' 3 )
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(.iW h .f 7.64 ( 2. .:G )
14 c ::,n a o1.1 a og ieed 41 c.b to the wt w etl E/C, at no hind res tire oot - "i ia i ;
t o, i.) C o v er f i l l in.'.) o n .:.. 2 es t mLd 6
- i es e n d ma .i. i i t a i n si e water 1ne in this ruptur ed S / la 4: b wcc t h e.: t op of the J-ttoce ( U . S .i in order tii p r m ic.i i i t h er io41 s t. r .. L i + 1 c. n t 2 c n (
. '.5> ':.o . rcn evit r u .s t :.tr eti 5./G depro murizati.zu (0.0).
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