IR 05000352/1989013
| ML20248E754 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/26/1989 |
| From: | Conte R, Easlick T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20248E745 | List: |
| References | |
| 50-352-89-13OL, 50-353-89-22OL, NUDOCS 8910050408 | |
| Download: ML20248E754 (147) | |
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e U.S. NUCLEAR REGULATORY COMMISSION
REGION I
' Report N'os.: 50-352/89-13 (OL)
50-353/89-22 (OL)
L Docket Nos.: 50-352 i
50-353 License Nos.~: NPF-39 l
. Licensee: Philadelphia Electric Company P. O. Box A Sanatoga, Pennsylvania 19464 Facility Name: Limerick Generating Station, Units 1 & 2 Examination At: .Sanatoga, Pennsylvania
. Examination' Conducted: June 12 - 16, 1989, and June 19, 1989 Examiners: A. Howe, Senior Operations Engineer D. Jarrell, (PNL)
G. Buckley, (PNL, Examiner Certification)
M. Riches, (PNL, Examiner Certification)
B. Wetzel, Operations Engineer, RIII (Observing)'
Chief Inspector: 9 .2 G ? 7 Theodore A. Eas11ck a perations Enginee Date Approved by: d '
b Richard J. Conte, ChiefF BWR Section S/2 c Date f
Operations Branch, Division of Reactor Safety Examination Summary: Written examinations and operating tests were admini-stered to three (3) senior reactor operator (SRO) candidates and six (6)
reactor operator (RO). candidates. The SR0s passed these examinations. The R0s passed the' written test but only five (5) passed the operating test. The one j RO candidate was not well prepared for the licensing examination, as demon- j strated by failing the walktarough and simulator portions of the Operating
. Test. The candidate performed marginally'during the licensee's audit examina-tions. Except as noted above, the candidates were well prepared for the licensing examination j l
8910050408 890927 _
PDR ADOCK 0500 )
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For the purpose of multi-unit.(Limerick 1 and 2) licensing of personnel who i were previously licensed on Limerick 1, operating tests (plant walk-through)
were administered to two (2) reactor operators. The written examination and simulator part of the operating test were waived in accordance with provisions of 10 CFR 55 and the guidance in NUREG-1021. The R0s passed these examination ' A requalification re-examination operating test (simulator part) was admini-stered to one (1). reactor operator. The RO passed this operating tes .
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DETAILS
, Introduction and Overview The' NRC examiners conducted this replacement examination.for three (3)
Senior Reactor Operators (SRO) - one (1) SRO instant, two (2) SRO upgrades and six (6) Reactor Operators (RO). The examinations were administered in accordance with NUREG 1021,-Rev. 5, dated January 1, 1989, Examiner Standards (ES). The results are summarized below:
l R0 l SR0 l
1 PASS / FAIL I PASS / FAIL l-l Written i 6/0 1 3/0 l l Operating l 5/1 1 3/0 l l- Overall l__ 5/1 1 3/0 1 The NRC examiners conducted a Unit 1 and 2 differences examination for two (2) Reactor Operators (RO). The results are summarized below:
l R0 l SR0 -l l PASS / FAIL l PASS / FAIL- l l Operating- l 2/0 l N/A l l Overall l 2/0 1 N/A l The NRC examiners conducted a requalification re-examination for one (1)
Reactor Operator (RO). The results are summarized below:
l R0 l SRO l l PASS / FAIL l PASS / FAIL l l Operating l 1/0 l N/A 1 i Overall i 1/0 l N/A l 2.0 Persons Contacted U.S. Nuclear Regulatory Commission
- A. Howe, Senior Operations Engineer
- T. Easlick, Operations Engineer B. Wetzel, Operations Engineer (RIII)
G. Buckley, Examiner (PNL)
M. Riches, Examiner (PNL)
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l 2.2 Philadelphia Electric Company
- R. Nuntz, OPS Training Supervisor
- L. Hopkins, PSST. Superintendent OPS
- E.-Firth, Superintendent - Training
- D. Weiksner, LOT Lead _ Instructor (G.P.)
H * Martin, LOT Instructor (G.P.)
- B. Sokso, Licensing l J. Monaghan, Operation (*) Denotes persons attending exit meeting 3.0 Examination Related Findings / Conclusions 3.1 The following is a- summary of general strengths and deficiencies noted on the operating tests. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require Strengths Good use of Technical Specifications by SR0 Good overall control board operations by R0s, Knowledge of entry conditions for Emergency Procedures, . Good use of Emergency Trip-200 Series Procedure Deficiencies Implementation of step LQ-7, of Trip procedure T-117, for terminating and preventing all injection into the RP Informal communications between crew member Inconsistent use of Alarm Response Procedure .2 The following is a summary of general strengths and deficiencies noted from the grading of the SRO and RO written examination. This-information is being provided to aid the licensee in upgrading licensee requalification training programs. No licensee response is require Strengths Knowledge of recirculation pump restart limitation Knowledge of Thermal Limit Immediate operator actions for " Inadvertent Opening of a Relief Valve," OT-11 Knowledge of Feedwater Control System response on a transmitter failure.
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Deficiencies Knowledge of hydraulically isolating an HCU and still maintain l cooling to the ro t
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Knowledge of basis for " Alternate. Shutdown Cooling," T-11 L Definition of Delayed Neutron Generation Tim .3 Training Program Comments
'A review of-the simulator portion of the operating test. indicated .
l training was weak in the area of " Level / Power Control," procedure .)
T-117, specifically,-that step in the procedure which requires the -'
operators to lower level by terminating and preventing all injection -
into the RPV. Two (2) of the three (3) examination crews had difficulty implementing that step, in that all systems were not
. prevented from injecting as required. This apparent weakness was discussed with the' facility and actions would be initiated to analyze and correct the proble Except as noted below, the applicants were well prepared for the licer : mg examinations. The facility provided reference material was' A equate.and in accordance with the NRC's "90-Day letter."-
The licensee was granted a one-week extension on the due date fo submitting the Personal Qualification Statements (NRC Form 398), for two individuals, after the training department expressed a concern about their satisfactory completion of the training program. This extension was granted to' provide time for the license'e to further evaluate the applicants. The licensee decided that only.one of the two would be recommended to the NRC for a license examination. NRC review of the results of the individual's operating test indicated a wide range of weaknesses. The candidate failed the Control Room Systems and Integrated Plant Ooerations portions of the Operating Test. The candidate did not possess an adeq'uate knowledge in the area of plant system design and operation. The examination brings into question the adequacy of the certification process established by the licensee. This concern was discussed with the licensee when they.were notified of the final examination results. The licensee agreed with the NRC's evaluation and stated in retrospect that the candidate should not have been recommended for a licensing examina-tio .0 Management Meeting On June.1,1989, a pre-examination review was conducted at the NRC Region
.I Office. The licensee was informed that, although a post-examination review was not conducted, comments on the written examination would be accepted. These comments were received at the exit meetin a_ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ - _ - _ - _ _ . _ _ - - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
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The control ' room staff was very cooperative in maintaining an environment
,' conducive to an operating test administratio Facility access went smoothly with good support from Health Physics:and l Security.
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The- examination strengths and deficiencies in. section 3.1 and the training comments in section 3.3 were~ discussed along~with possible
- l corrective actions. The final results of the examination were not discussed at its exit meeting. Every' effort would be made to provide the applicants results in approximately 30 working day It was requested that the licensee provide an evaluation report on the individual administered the requalification re-take examination per ES-601, section E. This report was submitted June 28, 1989, to the regional office and did meet the requirements of ES-60 Attachments:
1. Senior Reactor Operator Written Examination and answer key -
2. Reactor Operator Written Examination and answer key 3. Facility Comments on Written Examination after facility review
.4. NRC response to facility comments Simulator. Fidelity Report.
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. DRAFT COPY U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION 1 FACILITY: Limerick 1 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 89/06/12
< INSTRUCTIONS TO CANDIDATE:
/ Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing orade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 24.00 24.00 4. REACTOR PRINCIPLES (7%)
THERMODYNAMICS (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
32.50 32.50 EMERGENCY AND ABNORMAL PLANT
\ EVOLUTIONS (33%)
43.50 ' 43.50 6. PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (13%)
10 % TOTALS FINAL GRADE All work done on this examination is my ow ~
I have neither given nor received ai Candidate's Signature DRAFT COPY
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' REACTOR PRINCIPLES (7%)-THERMODYNAMICS Page -2 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
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QUESTION 4.01- (2.00)
for EACH of the events below, STATE the order in which the three reactivit Doppler) will'y coefficients affect reactor (Moderator reactivity Temperature Coefficient, (FIRST, SECOND, Void, and. THIRD).
ALSO indicate whether.each coefficient will be adding + or .
reactivity during the event. Use all three coefficients for each event, A turbine stop valve' fails shut at power (1.0) One recirc pump trips at power (1.0)
ANSWER 4.01 (2.00)
. Void (+), Doppler ( , MTC - Void (-), Doppler (+ , MTC +
[+0.25] for. each + or - and [+0.25] for the order REFERENCE lL merick: LOT-145 . Limerick: LOT-146 .. Limerick: LOT-148 K110 292004K105 292004K101' 293006K113 291004K104
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,h REACTOR PRINCIPLES (7%). THERMODYNAMICS ~ Page;. 3'
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I L L ^ QUESTION ' L4.02 (1.00)5 SELECT which ONE -(1); of the following statements correctly defines the Delayed. Neutron _ Generation Time.
m 'The length of time'between ..... (1.0)
(a.). ~ neutron' absorption and subsequent delayed neutron thermalizatio . (b~. ); the production of a delayed neutron and subsequent neutron absorptio (c.) the production of delayed neutrons in successive' lifetime '(d.) delayed neutron thermalization and subsequent neutron absorptio '
- ANSWER 4.02 (1.00)
(c.) [+1.0]
REFERENCE \ Lime' rick': LOT-1420, L.0.'s 5 and K102 ..(KA's)
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_ _ _ REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 4 (7%) AND COMPONENTS (10%) (FUNDAMENTALSEXAM1 QUESTION 4.03 (1.00)
The power generated by the reactor at the beginning of core life comes from U-235 thermal fission and U-238 fast fission. Later in core life, larger and larger fractions of power generation are produced by fission of WHICH ONE (1) of the following isotopes? (1.0)
(a.) Am-241 (b.) Cm-244 (c.) Pu-238 (d.) Pu-239 ANSWER 4.03 (1.00)
(d.) [+1.0]
REFERENCE Limerick: LOT-1420, L.0.'s 5.b. and ~
292003K103 ..(KA's)
QUESTION 4.04 (1.00)
STATE the reason that the xenon concentration will INCREASE following a significant DECREASE in reactor powe (1.0)
ANSWER 4.04 (1.00) !
The decrease in the burnout term [+0.5] along with the continued production of xenon from iodine at the higher power dominates [+0.5]
causing the xenon concentration to increas (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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' REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 5 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
REEERENCE Limerick: LOT-0950 L.0. 2 K102 ..(KA's)
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(7%)-AND COMPONENTS (10%).(FUNDAMENTALS EXAM)_
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i-QUESTION- 4.05 (2.00)
MATCH'th'e following terms lon the left with their correct definition-on the right. (Only one correct answer each) (2.0)_
TERMS ANSWERS DEFINITIONS a.- Keff: 1. .Keff =. Critical The fraction of the thermal-neutrons that had been born
~. c Supercritical by the decay of delayed neutron precursor, d.- Excess Reactivity- The fraction of fission Effective Delayed- neutrons born from the-Neutron Fraction decay of a delayed neutron precurso #autdown Margin
, Multiplication factor-dealing with an infinitely large reactor 5. . Neutron population is decreasing from
\ generation to generatio . A measure of how far the I reactor is below the critical conditio . Multiplication factor dealing with a realistic reacto This includes leakag . The amount of additional i
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reactivity built into a reactor to allow it to reach full power for its operating cycle lif . Keff > 1 10. A measure of the departure from criticalit (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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., REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 7 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
ANSWER 4.05 (2.00) I , 0 03
[+0.33] each REFERENCE Limerick: LOT-142 . Limerick: LOT-143 K104 292002K110 292002K107 292002K111 ..(KA's)
QUESTION 4.06 (1.00)
SELECT ONE (1) of the following statements which correctly describes the effect of an INCREASE in the amount of non-condensible gases in a steam turbine condense (1.0)
(a.) Condenser pressure decrease (b.) Circulating water outlet temperature decrease (c.) Turbine Generator megawatts decreas (d.) Condensate subcooling increase ANSWER 4.06 (1.00)
(c.) [+1.0]
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--. - REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 8 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
REFERENCE i Limerick: General Electric Thermodynamics Text, Section . Limerick: LOT-050 K107 ..(KA's)
QUESTION 4.07 (1.00)
SELECT ONE (1) of the following statements that describes a correct heat balance relationshi (1.0)
(a.) If the feedwater temperature used in the heat balance calculation was HIGHER than the actual feedwater temperature then actual reactor power is HIGHER than calculated reactor powe (b.) If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED then actual reactor power is HIGHER than calculated reactor powe (c.) If the steam flow used in the heat balance calculation was LOWER than the actual steam flow then actual reactor power is LOWER than calculated reactor powe (d.) If the RWCU return temperature used in the heat balance calculation was HIGHER than actual RWCU return temperature than actual reactor power is LOWER than calculated reactor power.
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l ANSWER 4.07 (1.00)
l (a.) [+1.0]
REFERENCE Limerick: LOT-1300, 2.6/3.1 2.3/2.9 l
293007K113 293007K111 ..(KA's)
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-QUESTION' 4.08..- (1.50)'
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MATCH the a CONDITION;(ppropriate FAILURE L1-L3).to each Thermal MECHANISM Limit ~(F1-F3) AND LIMITING (a-c) given belo FAILURE LIMITING MECHANIS CONDITION Linear Heat Generation Rate:(LHGR)- '(0.5)
. ' Minimum-Critical Power Ratio (MCPR) (0.5) Average Planar-Linear Heat Generation Rate (APLHGR) (0.5)
FAILURE MECHANISML LIMITING CONDITION
- F Clad cracking from L1. Clad. plastic strain less becoming vapor, than 1%.
" blanketed".E F Clad cracking caused by _L Maximum clad temperature high stress from pellet of 2200 degrees expansio LF Clad melting' caused by L3. Coolant transition boilin decay & stored heat-following a LOC ANSWER 4.08 (1.50)
~ F '+0.25' '
L +0.25 F ~+ 0.25 L l+0.25
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' F '+0.251 L ' +0. 25.'
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i 14. REACTOR PRINCIPLES:(7%) THERMODYNAMICS- Page 10 t 1.7%) AND COMPONENTS (10%)-(FUNDAMENTALS EXAM)-
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fl.' Limerick:, LOT-1380, L.0. . Limerick: LOT-1390, L.0.- .
293009K107.- -293009K108' 293009K111 293009K112- ..(KA's);
-QUESTION: :4.09 :(1.00).
- During a cooldownLof the reactor. vessel from outside.the control
, . room, reactor pressure decreased from 885 psig.to 595 psig in
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.one HALF hou ~
' DETERMINE the reactor cooldown rate in deg/hr. (SHOW all work). (1.0)
ANSWE .09 (100) . .
First, convert psig, to. psia, by adding 14.7 psi.[+0.25]
- Then, referring to the' steam tables; .
900 psia. = 532 deg.F [+0.25 l'
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610 psia.=488deg.F[+0.25.;
, 532 deg.F -- 488 deg.F =44 deg.F / half
- hour, or 88 deg./hr. (+/- 2 deg./hr.) -[+0.25]
REFERENC . Limerick: LOT-1100, L.0. 3.
l, Limerick: l LOT-1230 ,
293003K123 ..(KA's)
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, REACTOR PRINCIPLES (7%) THERMODYNAMICS Page 11 (7%) AND COMPONENTS (10%) (FUNDAMENTALS EXAM)
QUESTION 4.10 (1.00)
SELECT the correct statement concerning the operation of a jet pum (1.0)
1 (a.) As flow exits the jet pump diffuser, fluid velocity increase (b.) The static pressure at the jet pump nozzle is greater than the downcomer annulus pressur (c.) The low pressure at the nozzle discharge draws the surrounding fluid in the jet pump throat are (d.) The constant area of the mixing section maintains the pressure constan ANSWER 4.10 (1.00)
(c.) [+1.0]
REFERENCE
\ Limerick: LOT-1290 L.0. 1 K105 ..(KA's)
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-QUESTION- '4.11 -(2.00)
' SELECT ONE (1) of the pump curve changes-(column B) which describes the effects. for EACH of the pumping system modifications (column A).
ASSUME one single centrifugal pump is already:in the system. (ASSUME ideal conditions.) Not all answers may be used and some may be used
,more than onc .(2.0)
COLUMN A ANSWER COLUMN B Pumping- . Pump System Modifications- ' Curve Changes
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'(a.) ' Throttle the~ discharge 1. Decreased flow rate valve on.the existin with decreased hea pump 2. Increased head and (b.)' Add one identical pump increased flow rat in paralle .-Double the flowrate (c.) Add one identical pump with the same hea in serie .. 4.' Increased head with
.(d.) _ Double-the speed of the decreased flow rat existing pum \ -5. Double the head with the same flow rat ' ANSWER- 4.11' (2.00)
l; (a.) [+0.5]
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(c.) X [+0.5] .5
i_ .- (d . ) [+0.5]
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- y -(7%)-AND COMPONENTS (10%)-(FUNDAMENTALS EXAM)'
REFERENC b' : l '. TLimerick: - ' LOT-1290' L.0. 13-& 1 z l q
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- ~ Water' hammer i's Edefined as the. liquid shock imposed on a piping '
system resulting-from a rapid change in flow.. The factors affecting q-themagnitude'of-impulse'forcearetime,(a.) ,
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ANSWER 4.12 (1.00)'
ci 0.5] ( bat Ceeur n D4fC ^ """^ *h " " "NS '
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\S % Tne eres %csto4SC To ?A w REFERENCE'
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< Limerick: LOT-1291 q
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j-QUESTION' 4.13 (2.50)
- A condensate filter / demineralized is removed from service
[' on high outlet conductivity of 0.1 micrombos/cm and high delta-P-of psi (0.5) !
- STATE why the following parameters are used as indications that filter demineralized should be removed from servic (2.0)
- 1. High conductivity
~ High delta-P- l I
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" ' REACTOR-PRINCIPLES (7%) THERMODYNAMICS: . Page 14
.g .(7%) AND COMPONENTS (10%).(FUNDAMENTALS EXAM)
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- ANSWER' 4.13 (2.50) g La . . 25 [+0.5]
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~ . (Conductivity is affected by. chloride, pH and othe impurity concentrations.) Therefore high conductivity is a' good. indication of degraded water qualit %tt Att,m " laucAwoo or top hea AcGc bch2Tio#'.)[+1.0] .
2. High' delta-P is an indication of particulate clogging (and_ reduced filtering efficiency.) :[+1.0]
REFERENCE 1.- Limerick: LOT-0520 L.0. 6. c & Limerick: Tech Specs' Bases Po. B 3/4-4-3
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QUESTION' 4.14 .'(1.00)i
';_ Given the following conditions: Recirculation pumps ~are runnin
< constant \ minimum speed, low thermal reactor. power (approx. 20%)g at .
-STATE how AND why core flow varies (INCREASE, DECREASE, or REMAIN THE SAME) when reactor power is increned by control rod withdrawa (1.0)
ANSWER 4.14 (1.00)
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-Core flow would increase [+0.5] due to an increase in natural circulation [+0.5] .
REFERENCE ' Limerick: LOT-0040, L.0.1 '291006K112 ..(KA's)
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QUESTION 4.15 (3.00)
543.1.A " Start-up of the Recirculation System" limits the number of starts.that may be attempted on a recirculation pumps, With Motor Generator AND recirculation pump at ambient temperature, (1) pump start (s) may be attempte Another pump start may NOT be attempted until a waiting period of (2) has elapse (1.0) With Motor Generator. AND recirculation pump warm from at least fifteen (15) minutes operation, (1) pump start (s) may be attempted. Another pum attempted until a waiting period of (2)p start mayhas N01elapse be (1.0) The RESTART limitations are to protect which ONE (1) of the following? (1.0)
(1.) The recirc. pump bearings due to inadequate lube oil being supplied on pump starts by the shaft driven oil pump (2.) The motor generator windings from overheating due to increased current flow on pump start \
(3.) ' The power supply to the recirc. pump motors due to possible internal electrical faults causing the pump trip (4.) The fluid drive coupling scoop tubes from exceeding their design cyclic stress limit ANSWER 4.15 (3.00) . two [+0.5]
2. one hour [+0.5] . one [+0.5]
'2. one hour [+0.5] (2.) [+1.0]
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4. < REACTOR PRINCIPLES (7%)' THERMODYNAMICS ~ Page 16-p, *
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. REFERENCE Limerick: LOT-0030, Pg.~28.
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- Limerick: 543.1.A., 7. K106 ..(KA's)
QUESTION 4.16 (2.00)
a.- DEFINE Net Positive Suction Head (NPSH). -(1.0)
1 EXPLAIN WHAT happens and WHY to the'available Recirculation Pump NPSH-if HPCI initiates at 70% powe (1.0)
- ANSWER 4.16 (2.00) NPSH is difference between total pressure at the eye of the pump and saturation pressure for the liqui .
(Equation Acceptable: NPSH = Psuc - Psat)
opsa r up - w.; - wq+ g[-+1.0]
- b. c The NPSH would increase [+0.5] because the temperature of the feedwater'would decrease,.thus increasing the subcooling at the eye of the recirculation pump. [+0.5]
REFERENCE 1. - Limerick: EIH: Heat Transfer and Fluid Flow, Chapter 6, L.0' # 10.9, 10.1 . ' Limerick: LOT-1210
' Limerick: LOT-123 !
291004K106 ..(KA's) I
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QUESTION :5.01 (l'00)-
.
' N Ascording' to SE-1 " Remote Shutdown" SELECT ONE:(1) of-the- following -
.RCIC system interlocks whichl remain' active when control is -
,' ,
transferred'to::the~ remote. shutdown panel.- (1.0)
h ~
"
-- ( a . ) . RCIC. turbine trip on_overspee '
- (b.) ' Steam supply valve closure on high reactor water level.
,4
- (c.).- Transfer _'of. suction.from CST.to suppression poo (d;)
- . Start on low reactor water level (-38").
U ANSWERL ;5.01~f(1.00) _
(a.):; ~[+1.0].
.1 REFERENCE'
-- 1.- -Limericki LOT-0735 L.0. : : L\imerick:- SE- '295016K201- ...(KA's)
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> EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS- Page 18-D (33%)
-QUESTION ~ 5.02 '(3.00)
,
=Concerning OT-114 (REV. 4) " Inadvertent Opening ofLa Relief. Valve":
FILL in the correct response for each blank in the following
' statements ~on Immediate Operator Action Place loop (s) of Suppression Pool Cooling in t servic (0.5)'
t Rapid shutdown per GP-4 initiated if the valve-cannot be shut within minute (0.5)
' :If the.SRV cannot be shut within (1) minutes or! suppression pool. temperature reaches (2)
deg F,. place the reactor mode switch in (3) .
(1.5). .High suppression pool temperature is an entry condition for T- . (0.5);
f LANSWER 5.02' .(3.00)
\ (both) [+0.5].
.
'b. 'I.5 minu'tes-[+0.5]
. c '. . minutes' ~+0 . 5'
. .
110 deg.F l+0.5l 1 shutdown [+0.5]
d. .T-102 (Containment):[+0.5]
REFERENCE Limerick: 0T-11 A102 295013G010 295013G011 ..(KA's)
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5. ' EMERGENCY AND ABNORMAL ~ PLANT EVOLUTIONS Page 19 (33%)
f' QUESTION 5.03 (1.00)
SELECT ONE (1) of the following * statements which is an immediate action in accordance with OT-117 "RPS Failure". (1.0)
(a.) Perform rapid plant shutdown per GP- (b.) Initiate rod insertion using the Reactor Manual Control Syste (c.) Place the mode switch in shutdow (d.) Trip the reactor recirculation pump . ANSWER 5.03 (1.00)
(c.) [+1.0]
REFERENCE Limerick: 0T-11 G010 ..(KA's)
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, , EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 20 (33%)
QUESTION 5.04 (1.00)
SELECT ONE (1) the following statements which is the basis for the Technical Specification minimum suppression pool level (22 ft). (1.0)
(a.) To ensure that the downcomers are submerged and that no blowdown flow can bypass the suppression poo (b.) To ensure that adequate head will be available to the suction of the ECCS pump (c.) To ensure that a sufficient volume is available to absorb the heat released during a blowdown and not exceed the design pressure of containmen (d.) To ensure that the SRV tailpipes will remain submerged so that r,o steam flow can bypass the suppression poo ANSWER 5.04 (1.00)
(c.) [+1.0]
\
REFERENCE Limerick: Technical Specifications B3/4-6-4 Limerick: TRIP Procedure T-102 295030G004 295030G011 ..(KA's)
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QUESTION 5.05 :' -l(1.00) '
. SELECT which ONE (1)'oflthe following failures in the Electro-
Hydraulic; Control:(EHC)systemwouldcauseahighpressure
- transient' on the+ reactor. ASSUME no other failures are present-in the syste (1. 0) '.
(a')- . Pressure': regulator ' A'. fails hig ~ (b.); Pressure regulator ' A' fails low.-
.(c.) . Pressure setpoint' fails hig (d.); . Pressure ~setpoint' fails lo JANSWERL '5.05 (1.00)-
(c.) [+1.0]-
REFERENCE: Limerick: . LOT-0590 295025K208' 295025A102 ..(KA's)
+
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=(33$)
o QUESTION 5.'06 -(3.00)
i; , Concerning the immediate actions on a loss of A.C. power: Confirming indications are received that a loss of 0FF-SITE power has-occurre . STATE the TWO (2) DIFFERENT plant components that require ~
starting verificatio (1.0) STATE the TWO (2) system lineups that must be performed in order to supply cooling to the drywell cooler (1.0) Subsequent to the loss of off-site power a loss of ALL occurs. STATE the TWO (2) plant responses that must be verifie (1.0)
ANSWER 5.06 (3.00) . 'a .- Diesel Generators '+0. 5' ESW pumps l+0.5:
\
2. at Line up ESW to RECW [+0.5]
b. Line up RECW to Drywell Coolers [+0. 5] . Reactor Scram ~+ 0. 5' MSIV Isolation '+ REFERENCE Limerick: LOT-1566, L.0.'s 2 & . Limerick: E-10/20, pg 1 & . Limerick: E-1, pg G010 295003G011 ..(KA's)
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I QUESTION .5.07 (1.00)
" ' 'if the. Reactor Water Level reaches 100 inches. STATE the. reason for-this action.: .~ (1.0)
ANSWE .07: (1.00).
To prevent unnecessary flooding of the main steam line [+1.0]
REFERENCE' Limerick: Technical Specifications,.3.11.2.1 a & . Limerick: EP-10 . Limerick:. LOT-1520.-
295008A103 295008G006- ..(KA's)
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QUESTION 5.08 (1.00)
Concerning the Feedwater Control S the reactor. level setpoint to (a) ystem, the automatic.setdown'of inches'is' initiated-(b)- seconds after the level decreases to (c)
inches. This avoids unacceptably (d) water levels following SCRAMS from high powe (1.0)
ANSWER' . 5.08 (1.00)
.l*7 R l lhSN(inches) (seconds) .5'(inches) high
.[+0.25] each f
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? - Limerick: LOT-550 IV.: '
295006K304 ..(KA's)
>.
-t
. QUESTION 5.09 (1.00)'
The. TRIP procedure for SCRAM T-100, Step S-7, directs operators.to SELECT
"ONEtrip (the turbine when generator load reaches 50 Mwatts.1) of the-following statements w ifor this ste (1.0)
,e
"' f(a.) ' Assures there is-adequate. capacity of the bypass valves to handle this load transfe (b.) ' Guards against overspeed of the turbine by preventing the generator: from tripping on _ reverse powe :(c.): Assures that.the End of Cycle Recirculation Pump Trip does not occu (d.) Prevents MSIV' closure upon NSSSS actuation on low steam line pressur \ .
. ANSWER 5.09 (1.00)
(b.) [+1.0]
REFERENCE- Limerick: T-100 LOT-1560 L.0.- K305 295005K304 ..(KA's)
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QUESTION 5.10 (2.50)
~
Pertaining to the Trip Procedures MATCH ONE (1) level in the right hand column to each of the function statements in the left hand colum (2.5)
FUNCTION STATEMENT ANSWER WATER LEVEL Upper limit for Reactor + 60 in water level per T-100 + 54 in Lower limit for Reactor water level per T-100 + 39 in Top of Active Fuel + 12.5 in Level required for T-111 in '
Level Restoration in 4 Entry condition for T-101 L RPV Control in in in
\
ANSWER 5.10 (2.50) ATR l ' s d 8 e 5
[+0.5] each
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REFERENCE:
1.- Limerick: T-100, T-101, T-111, LOT 1560 L. A101 .
295031A102- 295031A103 -295031A104 . . (KA's) .'
-QUESTION: 5.11 (1.00)'
Trip procedure T-101 "RPV' Control" is in In accordanc 'with the' T-200 series procedures, TWO (2) progress.other. systems to be used to inject Boron into the.RPV IF Standby Liquid Control CANNOT
'be used are (1) AND.(2). .
(1.0):
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A'NSWER 5.11 '(1.00) Control Rod Drive System (CRD)
2. .- Reactor Water Clean Up (RWCU)
j
[+0.50] each REFERENCE \;
' Limericki Trip Procedures: T-101, T-110,-T-111, T-112 Limerick: LOT -1562-295037A110' ..(KA's)
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' QUESTION _5.12- (2.'50) '
Given 'the initiating conditions in COLUMN A, Select the ' appropriate emergency classification from COLUMN B. EP-101 is provided as-Attachment 2. (NOTE: ~Each classification can be used more than once or not.at all. If the initiating conditidn-does NOT meet-the entry level requirements, state NO CLASSIFICATION REQUIRED.) (2.5)
COLUMN A~ ANSWER COLUMN (Condition): -(Classification) Confirmed earthquake of Unusual Event
.0.08g while at powe . Alert Unidentified leak' rate of 33 gpm from the reactor : Site Emergency Main Steam Line break General Emergency
. outside primary containment with the failure F022A &
F028A MSIV's to close Suppression pool temperature exceeds the upper limit of the heat capacity curve AND RPV level below to Active Fuel (TAF) p of e.- Offsite doses are calculated to be 3 rem whole body ANSWER 5.12' (2.50) . [+0.5] . '[+0.5] . [+0.5] . [+0.5] .4. .[+0.5]
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. REFERENC f. .
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' Limerick: ' LOT-1520, L.O. 1 (SRO). >
QUESTION 5.13 (1.50)
,' In(3) accordance ways that with theprocedure SE-1 reactor can be" Remote scrammedShutdown",
if evacuation LIST the of the THREE main ~ control . room' is required BEFORE immediate operator action
'can.beltaken.: (1.5)
ANSWER 5.13 '(1.50) Trip the Main Steam Line Rad Monitors (Aux.H Equipment Room) [+0.5] Open'the RPS breakers-at the 'Y'. panel (Aux. Equipment Room) [+0.5]
\'
i Op'en.the RPS breakers at the RPS Breaker Panel-(Inverter Room 1)- [+0.5]
As.s o Atc.tet AN R esee x e.s 10 Aute:meer uj/ SE- Sies 4. 5. n , 4. s 7, AJb 9.5. A .
REFERENCE Limerick: SE-1 Rev. 9 pg. A101 ...(KA's)
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-QUESTION 5.14 (2.00)
Concerning T-115 " Alternate Shutdown Cooling": STATE the reason for maintaining Reactor pressure a minimum of 50 psi AB0VE Suppression Pool pressur (1.0) STATE the reason for maintaining Reactor pressure BELOW 154 psi (1.0)
ANSWER 5.14 (2.00) To ensure that the safety relief valves can be kept open.. [+1.0] To ensure that the' low head ECCS systems can provide make-up to the RP [+1.0]
(W t Lt. At. S o Accatrr : " To Edsoe.c Sornc.icpT Ecouac Fi.ow REFERENCE M 00 Cr A T A C C c'A E - ")- Limerick: LOT 1560 Limerick: T-115 Limerick: T-99 295021A104 ..(KA's)
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QUESTION 5.15' (1.00)
OT-112. " Recirculation Pump- Trip" states that if both recirculation pumps have tripped, " Ensure RWCU is in service using two pumps."
SELECT ONE (1) of the following statements which describes the i basis. for this. step..- ASSUME- recirculation pump trips occur at full
-
power operatio (1.0)
(a.) ' Increased amount of particulate settle'out under these flow conditions. RWCU in-service will remove most of.the particulate.and reduce the likelihood of instrument plugging on recirculation pump restar . (b.) Enhances natural circulation through the core by increasing the delta-T between the core and the bottom head region . (c.. ) Maximizes the amount of. cold water removed from the bottom head ~ area and reduces the potential for thermal stratification problem ~ (d.) Prevents. idle recirculation loops from experiencing excessive cooldown from seal purge flow in-leakage which can cause excessive thermal stress transients on the loop piping on recirculation pump restar \
ANSWER 5.15 (1.00)
(c.) [+1.0]
REFERENC . Limerick: LOT-1540 L.0. . Limerick: OT -112 295001G007 ..(KA's)
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R : EMERGENCY AND ABNORMAL ~ PLANT EVOLUTIONS; Page 31-l'
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, QUESTION < 5.16: (3.00)'
For each ofLthe_ conditions / situations in column A, SELECT ALL TRIP. procedure (s) from column B that should be entered. (Consider
'the inital-conditions to be that'the plant is at 90% power operations.). Iflnone are applicable, state NONE. NOTE: Not all
'of the procedures in column B need be used, and some may be used
.more than onc (3.0)
COLUMN'A ANSWER COLUMN B (Conditions) '(TRIPProcedures)- Scram from improper T-100 SCRAM ranging by operator _
in intermediate range T-101 RPV control b.- Drywell Temp =140 deg F T-102 containment control Drywell Press = 1.7 psig ,
. T-103 secondary HPCI area radiation containment control leveis-are 10 times the alarm leve . T-104 radioactive release control
' Reattor Enclosure HVAC Isolated on~High Ra . T-111 level restoration T-112 Emergency blowdown T-117 tevel/ power control T-116 RPV flooding l
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1kNSWER' 5.16 :(3.00)-
,
a.: 1:
<
,b.: 3-
_ .(112,3'
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![+0.5] each REFERENCE
- 1. , Limerick: LOT-1560 Trip procedures -T-100,101,102,103,104.-
295036G011 :295032G011- 295010G011- '295038G011 ..(KA's)
-QUESTION: . 5.17 ,(2.00)
.\
Concernirig condenser vacuum, MATCH ONE (1) SETPOINT (Right Column)-
.to the: automatic action (Left Column).~ (2.0)
Automatic Actio'ns ANSWER- .'Setpoints-a.- MSIV closure 1. 25" Hg Va I
. Feed Pump Turbine Trip- 2. 24.7" Hg Va , Main. Turbine Trip 3. 22.2" Hg Va Bypass Valve closure 4. 18.5" Hg Va ]
5. 11" Hg Va . 8.54" Hg Vac.
[ 7. 7" Hg Va .a :
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LANSWE .17: (2.00).
. a.; '
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REFERENCE y . .
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1; Limerick:- LGS OT-116
- QUESTION 5.18
- (1.00)
SELECT .0NE- (1) ofithe following' statements which describes th effect of a loss of CRD pump 'A' due to low suction pressur (1.0)
(a.) A reactor recirc pump trip on loss.of ' seal purge flo (b.) Automatic ' opening. of strainer ~ bypass. valve and CRD pump 'B'
.
Pauto-start when suction pressure is above'25 " Hg Abs.'
(c.) The: suction flow path automatically transferring to the CS (d.) A continuing drop in system pressure unless manual action is take :.ANSWERL L5.18 (1.00)
,
?q\t . (d.): [+1. 0] '
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REFERENCE.- . Limerick: LOT-0070 L.0. . Limerick: LOT-1550 L.0. . Limerick: ON-107 295022A202 ..(KA's)
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-QUESTION 5.19 (2'.00)
.0ff-Normal procedure ON-113, " Loss of RECW", requires the tripping of the reactor recirculation pumps .if cooling is lost for ten (10)
minute SELECT ONE-(1) of the following statements which describes the basis for tripping the recirculation pumps ten (10) MINUTES after the loss of REC (1.0)
(1.) 'To provide time to insert control rods and stay within the. constraints of the operating map once the recirculation pumps are . trippe . (2.') To provide time for the operator to re-establish RECW operabilit (3.) Continued operation of recirculation pumps after ten minutes could damage'the pump seal (4'.) Continued operation of recirculation pumps after ten minutes could warp the pump shaf SELECT ONE (1) of the following statements which describes the basis for.the TEN (10) SECOND interval between the tripping of recirculation pump (1.0)
(1.) To prevent initiating reactor level 8 trips due to '
the level transient that would resul (2.) To prevent placing the reactor in the region of thermal-hydraulic instabilit (3.) To allow time for the first recirculation pump's discharge valve ~to close prior to exceeding its designed operating delta- (4.) To allow time for the first recirculation pump to coast down prior to tripping the second pump.
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JANSWER -5.19- (2.00)
a." ' (3. ) , [+1.0]
n' -
' (1.) .[+1.0].
REFERENCE'
~ ' 1.- ' Limerick: LOT-1550 L.0.-3.
l Limerick: Lot-46 . Limericki ' 0N-11 K303 ..(KA's)
.
QUESTION' 5.20-' (1.00)
^
Step 2.4.1 of.0N-119 " Loss of. Instrument Air" states that reactor power should be. reduced.to less than 45%. if instrument air, pressuie drops below 80 psig. SELECT ONE (1) of the following statements which dek,cribes the reason for the power reductio '(1.0)
(a.) Ensure adequate flow to the reactor should the condensate or feed pump minimum. flow valves drift ope (b.) Limit the reactivity transient caused by the closure of the MSIV's on a complete loss of ai (c.) . Prevent the violation of fuel thermal-limits caused by rods drifting into the cor (d.)
~
Limit the reactivity transient caused by the feedwater heater dump valves failing close .
5 ANSWER 5.20 (1.00)
- (a.) [+1.0]
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REFERENCE
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1.- Limerick: LOT-1550 L. . Limerick: ON-119 Base K203 ..(KA's)
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6'. : PLANT SYSTEMS (30%)' AND PLANT-WIDE GENERIC Page 38
RESPONSIBILITIES (13%)
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QUESTION: ~6.01- (l'.00) Reactor water; level has decreased to below -129 inche Drywell pressure is 1.2 psig.- The proper ECCS pump (s)
is operating. . ADS. valves will open in second (0.5) Reactor water level- has decreased to below -129. inches. .
Drywell pressure is 4.6 psig. The proper'ECCS pump (s)
o is operating. ADS valves will open in second (0.5)
.
-ANSWER 6.01 (1.00) '525 [+0.5] . [+0.5]
, REFERENC . Limerick: LOT-0330, pp. 9, 12, and 13, Figure 6, L.O. 2, 5, and \
218000K501 ..(KA's)
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~6. iPLANT SYSTEMS:(30%) AND PLANT-WIDE' GENERIC 'Page 39 RESPONSIBILITIES (13%)-
?> ' QUESTION 6.02 '(3.00)
For each of the reactor water levelsTin Column A SELECT'ALL of ,
,' the plant systems in Column B which would have DIRECT actuations/-
. confirmation setpoints associated with these water level .(3.0)
'
Column A- Column B W' (RPV Water Level) . (PlantSystem).
La; level 8 1.~ NSSSS level-3- RPS
' c .' : level 2 ECCS
. level 1 RRCS- Recire. System RCIC
'
- Main Turbine ANSWE * 6.02 (3.00) .3,6,7'.D9G , 3, 5 , 1 D o.'lO
.
T , . 3, 4, 6, (5) Ep.7C
'd . :1, 3 Do.'153
[30.25]-attr* [ i D.2od EAc.w ^B
'
. REFERENCE Limerick: LOT-0050; LOT-0180, L.0. 2.a; LOT-0300, L.0. 4;
. LOT-0215, t. 0. .216000K404 216000K405 216000K406 216000K407 ..(KA's)
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. . QUESTIO .03 (1.00).
SELECT ONE '(1) of Lthe' fo11owing statements- that describes the conditions will cause automatic initiation of the Standby Liquid
.: Control system..lHigh RPV Pressure (1093 psig) AND:'
(1.0).
l I+ (a.)l Low reactor water level '(-38 ' inches) AND 118 see timer-
.
timed ou (b.). No APRM downscale (4%) AND 118 see timer timed ou (c.) . Low reactor water, level (-129' inches) AND 118 see timer timed ou (d.) No APRM downscale (4%) AND Low reactor water: level
- (-38 inches);
.
ANSWER' 6.03_ (1.00)
(b.)L [+1.0]-
REFERENCE \ Limerick:. LOT-0310, L.0.1 A308 ..(KA's)
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QUESTION 6.04 (1.00)
Technical Specifications for Emergency Core Cooling System (ECCS)
allows "one (1) Automatic Depressurization System (ADS) valve to be inoperable for up to fourteen (14) days provided High Pressure Core Injection-(HPCI), Core Spray System (CSS) and Low Pressure
,
Coolant. Injection (LPCI) are operable."
f Concerning the basis for this Technical Specification, SELECT which ONE (1) of the following statements is true? (1.0)
(a.). Safety analysis only takes credit for four valves, so one
. valve out of service'for up to fourteen (14) days does not reduce system reliabilit (b.) -Risk assessment'for these valves indicates a negligible chance for a'second valve failure in fourteen (14) day (c.) Safety analysis only takes credit for HPCI, together with CSS and LPCI to provide adequate decay heat removal from 100%
power for up to fourteen (14) day (d.) Safety analysis shows heat capacity of the suppression pool is conservatively analyzed to allow continuous operation with one (1) fully opened Automatic Depressurization System (ADS) valv ANSWER 6.04 (1.00)
(a.) [+1.0]
REFERENCE Limerick: Technical Specifications 3/4, 5-2, Basis Sectio G006 ..(KA's)
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-QUESTION 6.05 (1.00)
SELECT the one (1) correct response. The opening of a SRV during full power operation would result in:
(1.0)
(a.) A decrease in indicated core flow due to a decrease in delta-P across the reactor cor (b.) An decrease steam flow signal being sent to the feedwater control syste (c.) An increase steam pressure signal being sent to the electro-hydraulic control syste (d.) An increase in indicated core flow due to a decrease in delta-P across the reactor cor ANSWER 6.05 (1.00)
(b.) [+1.0]
REFERENCEg Limerick: LOT-0120, L.0. 10.a; LOT-0550, L.0. A109 ..(KA's)
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QUESTION 6.06 -(1.00)
ASSUMING full power operation, three-element control, and no operator action, SELECT which ONE (1) of the followin would be the expected Feedwater Control System (FWCS)g responsestatements if the selected level-transmitter failed HIG (1.0)
RFP turbines would lock due to loss of level signal input
"
(a.)
and level would remain approximately the sam (b.) Steam and feedflow inputs would compensate for the error signal and level would. stabilize at a slightly lower leve (c.) Level input would automatically transfer to the other level transmitter and level would remain approximately the sam (d.) RFP turbines reduce speed in response to high level signal and level will continue to drop unless manual action is take ;
ANSWER 6.06 (1.00)
(d.) >[+1.0]
s
REFERENCE Limerick: LOT-0550, L.0. !
259002A203 ..(KA's) .
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QUESTION 6.07 (1.00)
A-surveillance was run to test the ECCS actuation signal'on the-
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, Division I. Diesel. Using-the following data and the attached
' Tech. Spec '
. Diesel RPM: 900 in-7' seconds
, Generator Voltage: 4052 in 9 seconds
,e ' Generator Frecuency: 60 Hz in 9 seconds
- Generator' Loac : 2800 kW in'135 seconds
- Diesel: Air Start Receivers::230 psig
- Generator was aligned with the emergency bus and run with the
'
- load that was recorded above for 60 minute a . .- DETERMINE.the operability status of the Division I Diesel Generator (0,5) . STATE the Tech.; Specs. used for this' determination. ASSUME no.other trip. inputs are being-teste (0.5)
. ANSWER 6.07 (1.00)
a.- Diespl shoul'd be declared IN0PERABLE. [+0.5]
-
b.: Tech. Spec. 4.8.1.1.2.a.A [+0.5]
k mw REFERENCE- Limerick: LOT-0670, L.0. G005 ..(KA's)
-QUESTION 6.08 (3.00)
STATE _SIX (6) control room indications which could verify that a reactor safety relief valve is open. ASSUME that the Reactor is at 100 % powe (3.0)
,
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tANSWER .6.08 ~(3.00)
- L1. ' generator. load reduction 2. ' bypass; valve. closure
'3. ~SRV/ head vent-valve leaking ~ alarm 4. safety relief valve open alarm
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5. relief valve position lights steam flow / feed flow mismatch increasing suppression. pool. temperature
- s>. -rAo ec. Taw dny sTx76) $ETEO.50] each; max. [+3.0]
REFERENCE
- Limerick: OT-114 - Limerick: Technical Specifications 4.8.1.1.2. A401 ..(KA's)
!; LQUESTION 6.09 (1.00)
A LOCA occurs concurrent with a loss of offsite power. SELECT which ONE (1) of,the following pieces of equipment must be MANUALL restarte (1.0)
(a.) dyrwell chiller compressors-(b.) drywell chilled water pumps (c.). CRD pumps (d.). 'ESW pumps
.
ANSWER- 6.09 (1.00).
'(c.) [+1.0]
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' REFERENCE 1.- . Limerick: LOT-0660, L.0; 6; LOT-0450 L.O. 9.
j ~ 262001K301 262001K602 ..(KA's)
~
. QUESTION '6.10 (2.00)
'.The following plant. conditions apply:
Reactor power 45%'-
Total. recirc ~1oop flow '35%.
' Two recirc loop operatio .
LReactor power is being INCREASED to 75% with control rod manipulation ONLY. Using the APRM flow-biased power formulas, DETERMINE if
.this power. increase would result.in a. ROD WITHDRAWAL BLOCK; FLOW-
~ BIASED NEUTRON FLUX SCRAM, Tor NO RESULT. ASSUME total recirc. flow remains constan (Showwork). (2.0)
ANSWER- _6.10 (2.00)
.\'
.R. Block = 0.58W +-50%- [+0.25]
= 0.58*(35) + 50 = 70.3% [+0.5]
Scram = 0.58W + 59% [+0.25]
= 0.58*(35) + 59% = 79.3% [+0.5]
Would result.in a Rod Withdrawal Block [+0.5]
. REFERENCE Limerick: LOT-0270, L.0. ~ . Limerick: Technical Specifications 3. A104 ..(KA's)
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-QUESTION 6.11' (3.00)
' For each of the.NSSSS group isolations in COLUMN A SELECT all'of-the plant conditions in COLUMN B that would initiate isolation signals.. If none are applicable, STATE NONE. Note - not all of the plant-conditions in COLUMN B need be used,.end some may be used more than once.'(ASSUME: the Mode Switch is in RUN) (3.0)
COLUMN AJ COLUMN B Group Isolations ANSWER Reactor Conditions MSIV's and Steam Drain- 1. Low reactor water level Lines (IA) (-42 inches) Main Steam and Reactor Main steam line high
~ Sample Lines (IB) rad (3.5X) RHR Heat Exchanger 3. Steam supply low Vacuum Breaker Lines pressure (700'psig)
(IIC). Reactor Water Cleanup 4. High drywell pressure Lines (III) (1.72) HPCI Turbine Exhaust 5. Steam line low pressure Vacuam Breaker Lines (90 psig)
~(IVB)' - AND-high drywell pressure RCIC Turbine Exhaust (1.72 psig)
Vacuum Breaker Lines (VB) 6. Low condenser vacuum-(11.5 psia)
'
7. Standby liquid control initiation 8. Steam line high flow (140%)
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ANSWER 6.11 (3.00) ,3,6,8j(O 1, 2 D o,sa D 'O , 4 t'o 53 , 7 r. + o 53 c+o.sn NONE C+o Cl
{+0d5] cacii correct-response-.mm REFERENCE Limerick: LOT-0180, L.0. K101 223002K102 223002K104 223002K107 ..(KA's)
QUESTION 6.12- (1.00)
Given the following scenario, SELECT the ONE (1) trip that caused the RWCU isolatio (1.0)
-
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-During startup, reactor water level is bein0 controlled with the reactor water cleanup blowdown lin Reactor water level is slowly INCREASING and the operator FULLY opens the blowdown valve to lower RPV leve Some time later annunciator F-2 on panel 186802, "RWCU System Isolated" is received and reactor water level is again INCREASIN ASSUME reactor water level is 28 inches; reactor temperature is 320 deg. F; reactor pressure is 75 psi (a.) high filter /demin delta-P (b.) high downstream pressure (c.) hign reactor vessel level
]
(d.) high filter /demin inlet temperature l
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) ANSWER' 6.12 (1.00)
'
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[+1.0]
e REFERENCE'
- Limerick: LOT-0110, L.O. 7.
!
204000A304 204000KO3 ..(KA's)'
QUESTION 6.13' (1.00)
'During a performance of the 7-day Operability Surveillance on
, Division'1 125VDC Battery System, it was found that Battery 1A1
'did not meets its category "A" requirements for float voltag Using the Technical Specifications, DETERMINE which ONE (1) of the
- following statements accurately reflects the status of the DC Electrical syste (1.0)
(a'. ) - Division I is' INOPERABLE and must be' restored to operability wi' thin 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s-or be in hot shutdown within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> . (b . )' _ Battery 1A1 is OPERABLE if all of- its' category ."B" limits are verified within limits in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and category "A" and "B" limits are met within 6 day .(c.)~ Battery 1A1 is INOPERABLE and must be restored to operability within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or verify the operability of the other Division I batteries within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (d.) Battery 1Al is OPERABLE if all of the remaining category
-
"A" limits are verified within limits and the out-of-limit parameter is restored to operable limits within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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ANSWE ;6i13l -(1.00)
(b.)- [+1.0]. Eh j s b-so Atte.PTAELE P REFERENCE
.o Limerick: LOT-0690, L.O. .
2.- Limerick: Technical Specifications, Table 4.8.2.1-1.
n n 263000G005 ..(KA's)-
QUESTION 6.14' (l'.00)
With the reactor operating at 80% power, the operator notices a sudden increase in BOTH reactor power AND water level. SELECT which ONE (1)_of the following events.would explain this transien .
(1.0)
(a.) RFP turbine speed increas (b.) 'A main condenser bypass valve openin . T (c.) An'SRV openin (d.) "3A feedwater heater isolates.
,
- e RNSWER 6.14 (1.00)
(a.') [+1'. 0]
REFERENCE
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. Limerick: LOT-0540, L.0. 1 K312 ..(KA's)
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. QUESTION 6.15 (1.00)
SELECT which ONE (1)'of the following' statements is TRUE concerning
,
the Control Rod Drive Hydraulic System response during a. SCRAM 7
-
(1.0)
! .
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(a.), The. Scram pilot valve energizes to vent the air off the i Scram inlet and outlet valves.-
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(b.) The Scram Discharge Volume (SDV) vent andEdrain air pilot !
valves energize to vent the air off the Scram discharge ;
volume ~ vent and drain valve (c.) If one of the Scram Discharge Volume (SDV) vent and drain air pilot valves fails to reposition, the Scram Discharge Volume will remain vented ~and drained.-
.
- (d.) '
If any scram' pilot valve fails, the action of the backup H scram valves will.cause the-rod associated with the failed l scram pilot valve to scram.~
f ANSWER 6.15 .(1.00)
(d.)- [+1.0] l
!
REFERENCE .
i Limerick: LOT-0070, L.0. 5 and K107 ...(KA's)
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QUESTION 6.16 (2.00)
Answer'the following concerning the Recirculation System: An Anticipated Transient Without a Scram recirculation pump trip (ATWS-RPT) initiates on or on . (1.0) An End of Cycle recirculation pump trip (E0C-RPT) initiates on a when power is greater than as sensed by first stage pressur (1.0)
. . ,
ANSWER '6.16 (2.00) Reactor Water Low Level (-38 inches) ~+ 0.51 Reactor High Pressure (1093 psig) '+ Turbine Trip ;+0.5; 30% + REFERENCE
\ Limerick: LOT-003 K413 202001K414 ..(KA's)
QUESTION 6.17 *(1.50)
Concerning the Control Room Ventilation System, STATE the TWO (2)
conditions which would cause the Control Room Emergency Fresh Air Supply System to initiate. INCLUDE where each of these conditions a'e r sense (1.5)
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o ANSWER 6.17 l(1.50)
.a;- l '. High' Radiation [+0.5] in the Air Intake Duct' [+0.25] High Chlorine' [+0.5] in the Air Intake Duct [+0.25]
REFERENCE
' 1. . Limerick: LOT-0450' L.O. 3 & '
-290003K101:- ~290003K102 290003K401 ..(KA's)
.-
QUESTION 6.18_ (1.00).
The MSIV Leakage Control System exhaust is routed through-the (1) system and the'(2) which minimize'the amount of radioactivity released to the environmen (1.0)
ANSWER: \ 6.18
,
(1.00) Standby Gas Treatment Sysiem -[+0.5]
. Reactor' Enclosure Recirculation System [+0.5]
( Ano Acccer ReAtxot. Eatt.osest Eqost gtpr tout Agr wee r eta Aost sqatu),
REFERENCE-
.
L Limerick: LOT-0125 L.0.'s 1 & .- Limerick: -5-40.1A . Limerick: SE-10 .
i-239003K102 ..(KA's)
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QUESTION -6.19 (1.00)
SELECT which ONE (1) of the following systems, in conjunction with the Centrol Rod Velocity Limiter reduces the consequences of a rod drop acciden (1.0)
(a.) Rod Drive Control System (b.) Rod Worth Minimizer System (c.) Reactor Manual Control System
%
(d.) Rod Block Monitor System ANSWER 6.19 (1.00)
(b.) [+1.0]
REFERENCE
, Limerick: LOT-0100 L.0. \
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201006K105 201006K501 201006G004 ..(KA's)
QUESTION 6.20 (1.50)
Concerning the Condensate System during full power operation:
STATE the TWO (2) automatic actions that would occur on a condensate pump trip. INCLUDE setpoint (1.5)
ANSWER 6.20 (1.50) Recirc pump runback [+0.5] to 75% core flow (60% speed)
[+0.25] RFP turbine runback [+0.5] to 78% speed [+0.25]
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- REFERENCE'
, ' Limerick: LOT-0520, L.0. 1 K304 ..(KA's)
QUESTION 6.21 (2.00) STATE the: safety limit and limiting safety' system setting
- that protects against-overpressurization of the reactor
>
vesse1~. (1.0)-
.If an~ overpressurization transient occurred that exceeded
~ the safety limit, WHAT agencies / individuals must be notified: within one (1) hour of event (0.25)-
2. 'within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of event (0.75)
ANSWER 6.21 L(2.00)
. psig :+0.5] psig :+0.5]
' . NRC Oper'tion a Center [+0.25
- 2. 'Vice President, LGS :+0.25]]
Plant Manager Nuclear Review Board [+0.25][+0.25]
REFERENCE Limerick: - LOT-0010, L.0.1, 2, and 3 (SRO) .
~ . Limerick:- Technical Specifications 2.1.3, -Table 2.2.1-1, K507 290002G005 290002G003 290002G002 ..(KA's)
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QUESTION 6.22 (1.00)
FILL in the chart of exposure limits as stated in HP-102,
' Administrative Dose Limits, Guidelines and Notification Requirements."- (1.0)
ASSUME the following: A signed and completed NkC-4 Form is on fil . No exposure extensions have'been authorize . All exposure will be received at PEco facilitie QUARTER YEAR (MREM) (MREM)
Whole Body .
Extremity Skin 3.
ANSWER 6.22 (1.00) WRem '+0.251 ' mrem l+0.25l mrem '0.25
+ mrem '+0.25 REFERENCE Limerick: LOT-170 . Limerick: HP-10 K103 ..(KA's)
QUESTION 6.23 (1.50)
Prior to exceeding a dose control level of mrem / quarter and up to 1500 mrem / quarter, a dose extension request must be approved by the and the s (1.5)
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ANSWER' 6.23 (1.50) : [+0.5]
2. . Department Senior Engineer [+0.5] Senior Health-Physicist [+0.5]
REFERENCE
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- Limerick: LOT-1705 Limerick:. HP-102 294001K103' ..(KA's)
QUESTION 6.24 '(1.50) -
In accordance'with Administrative Procedure A-7, Section 5.5, STATE THREE (3) steps which.the Control Supervisor is authorized to take to limit and control, access to the control' room area.' (1.5)
\
ANSWER l6.24 -(1.50)
- . Require all business not pertaining to-the control panels
,
or operating stations be conducted outside of the. " Blue"
"
lin . Require pass-through traffic to follow the " Yellow" lined rout ' . Order any individual from the control area whose presence
' poses.a threat or potential threat to plant safet .. Order the immediate departure (or removal) from the control
, room any unnecessary people when a plant emergency exist ; Any three (3) [+0.5] each; max. [+1.5]
?
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' . Limerick: LOT-1570,.V. . Limerick: Administrative Procedure, A-7, Section 5.5.- >
- 294001K105 ..(KA's)
-QUESTION _6.25' (1.00)
L SELECT which ONE (1) of the following statements is the basis for
~the change :in' chloride limit from; power operations (0.2 ppm) to cold' shutdown (0.5 ppm). (1.0)
'
(a.) 'The rate of formation of insoluble metallic corrosion
. products is proportional to'. coolant temperatur .(b.) -Galvanic corrosion rate increases proportionally with the coolant temperature.
l .. (c.) Stress corrosion' cracking of stainless' steel increases
.with increasing temperatur l (d.). Oxidation of. carbon steel increases with temperatur \
ANSWER- .6.25 (1.00)-
(c.) [+1.0].
L l REFERENCE l~ Limerick: Technical Specification A114 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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QUESTION 6.26 (1.00) USING the shell prewarming time graph (attachment #1) and given the conditions below, DETERMINE the time at which prewarming the shell will be complet (0.5)
- Shell pressure increased to 60 psig at 1:30 First stage shell temperature (initial) - 155 deg During shell warming, STATE the maximum limit as it pertains to heat-up rat (0.5)
ANSWER 6.26 (1.00) :45 a.m. (+/- 5 minutes) [+0.5] , deg F/Hr [+0.5]
REFERENCE Limerick: LOT-0560 and 059 \ Li'merick: GP-2, 3. . timerick: 501. A108 ..(KA's)
QUESTION 6.27 (1.50)
During shift turnover, operators reporting for duty shall receive a verbal report from the previous shift. STATE the six (6) general topics required to be covered during the verbal turnove (1.5)
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.' l ANSWER '6.27' (1.50) General operating condition of the plan [+0.25] ~ Specific operations performed and difficulties encountered
' during the previous shift. [+0.25]
l 3. . Scheduled plant operation [+0.25]
~ Equipment outages and maintenance work in progress.- [+0.25] Status:of safety related equipment and condition [+0.25] Temporary circuit authorization [+0.25]
REFERENCE
' Limerick: Administrative procedure A-7, Section 5.1 A105 ..(KA's)
QUESTION 6.28 (1.50)
\
According'to Administrative Procedure A-7, Shift Operations: In the control room, the minimum number of licensed operators required to be in the reactor control area of each unit is
.
(0.25) The number of licensed operators assigned exclusively to the reactor in the control room for startup, scheduled shutdown and recovery from a reactor trip is . (0.25) The reactor operator at the controls may momentarily be absent from the designated area with permission of the Control Supervisor during an emergency affecting safety of operations for two reasons. STATE the TWO (2) reason (1.0)
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ANSWER 6.28? (1.50)-
- a.: 1;'[+0.25]!
t
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b.: 2 [+0.25]
c.- '
Acknowledge' annunciators . [+0.5]
Initiate = corrective actions [+0.5] .
-l ~d
-.
' REFERENCE-
< . Limerick:' Administrative Procedure A-7, 5.1.3.1, 5.4.1,'
'
5.4.3, and'5.10. ;2L Limerick: . Technical. Specifications, Table 6.2.2-1.-
f 294001A103, ..(KA's)
QUESTION - 6.29 (1.50).
In'accordance with.the Emergency: Exposure Guidelines, STATE the allowable' doses'(in REM) for the'following situation (1.5)
,
WHOLE
'
~
BODY
, Life Saving and'
Reduction of Injur .
OperationLof Equipment'
to Mitigate ~an< Emergency Protection of Health and Safety of the Public .
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' PLANT SYSTEMS (30%) AND PLANT-WIDE GENERIC Page 62 RESPONSIBILITIES (13%)
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ANSWER 6.29 .(1.50)
, [+0.5]
2. . 25 [+0.5] [+0.5]
REFERENCE Limerick: EP-314 294001A116 ..(KA's)
QUESTION 6.30 (2.00)
In accordance with Administrative Procedure, A-7," Shift Operations", STATE what FOUR (4) items must be verified to assure that a procedure is vali (2.0)
s ANSWER 6.30 (2.00) Must be stamped in red " Control Copy." [+0.5]
ha T M W Ac-c . Stetion-6 superintendent's (or alternate's) signature (and date). [+0.5] QA Superintendent's (or alternate's) signature (and date).
[+0.5] Date of use must be later than the effective dat [+0.5]
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4 PLANT SYSTEMSL(30%)5 AND PLANT-WIDE GENERIC Page 63 RESPONSIBILITIES-(13%)
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... LREFERENCE x
' Limerick: : LOT.1570, L.0. ~
, l . Limerick:.? Technical' Specification 6. r
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' Limericki.'AdministrativeProcedureA-7,5.2.1'.
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.294001A101- ..(KA's)
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QUESTION; VALUE REFERENCE L '
-4;014 2.00' .900025 .4.02 1.00 9000254
"
c .4.03 1.00 9000255-
.4.0 41 -1.001 9000256 4.05 '2.00 900025 .0 ?1.00 9000258 4.07; 1.00 : 9000259
, 4.08- 1.50 9000260.
p 4.09' 1.00 9000261:
4.10 1.00, 9000262
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- 4.11
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2.00 :9000263 !
4.12L 1.00- 9000264 i
.4.13 2.50 9000265:- '!
4.14 L1.00 9000266 R 4.15c 3.00 ^9000267 4.16 2.00 9000268 i
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24.00 '
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5.01 1.00 9000269 5.02 3.00- 9000270-F 5.0 .00 9000271 l
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5.04 1.00 9000272
, 5.05- 1.00 9000273 ~I 5.06: 3.00 9000274- !
5.07 1.00- 9000275- l 5.08 1.00' 9000276 -
5.09 l'00a
. 9000277' 'i 5.10 \ ' 2.50 .9000278:
.5.11 1.00- 9000279 5.12- 2.50- 9000280 i 5.15
'1.00
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.5.16 : J 3.00
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9000284 .!
5.17- 2.00L 9000285 J 5.18 1.00 .9000286- !
5.19 2.00 9000287 1 5.20 1.00 9000288
....... ;
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32.50 i 6.01: 1.00 9000298 6.02 3.00 !
9000299 l 6.03 1.00 9000300 6.04 1.00 l 9000301
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6.05- 1.00 9000302 !
6.06 1.00' '9000303 6.07- 1.00 9000304
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. 6.08 - 3.00- 9000305 l 6.09 1.00 9000306 ]i 6.10- 2.00 9000307-6.11 3.00 9000308 i
'6.12 1.00 9000309 6.13' 1.00 9000310 )
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,00ESTION'
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VALUEL REFERENCEL 6.14- - 1.00 ' 9000311 15.15 1.00- 9000312
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6.16- 2.00 9000313-
'6.17 -1.50 -9000314 6.18 1.00i 9000315 6.1 .00 900031 '6.20' 1.50; 9000317 4 ;6.21' 2.00 9000318 6.22 11.00 .9000289,
- .2 :1.50 9000290-6.24- 1.50 9000291-
'6.25 r1.0 .26 -1.00 9000293
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'6.27 1.50 9000294-6.28 1.50 9000295
'6.29- .1.50: .9000296 6.30 2.00 900029 .
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DRAFT COPY U.-5. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION' 1 l, FACILITY: Limerick 1 REACTOR TYPE:' BWR-GE4 l
DATE ADMINISTERED: 89/06/12 INSTRUCTIONS TO CANDIDATE:
'Use' separate paper for:the, answers. Write answers on one side onl Staple question sheet on top of the answer sheets. Points for each question are indicated in' parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers ~will be picked up six (6) hours after-the examination starts.-
% OF CATEGORY. % OF CANDIDATE'S CATEGORY-VALUE TOTAL SCORE VALUE ' CATEGORY 25.00~ 25.00 REACTOR PRINCIPLES'(7%)
THERMODYNAMICS (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
27.00 27.00 EMERGENCY AND ABNORMAL PLANT
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EVOLUTIONS (27%).
48.00 48.00 PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC RESPONSIBILITIES (10%)
10 % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor received' ai Candidate's Signature DRAFT COPY l
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< REACTOR PllINCIPLES (7%) THERMODYNAMICS Page 2 (7%) AND !TMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.01 (2.00)
For EACH of the events below, STATE the order in which the three reactivit Doppler) y coefficients (Moderator Temperature Coefficient, Void,will affect reactor rea ALSO indicate whether each coefficient will be adding positive (+)
or negative (-) reactivity during the event. Use all three coefficients for each even Turbine stop valve fails shut at power (1.0) One recirc pump trips at power (1.0)
ANSWER 1.01 (2.00) Void (+), Doppler (-), MTC (-) Void (-), Doppler (+), MTC (+)
[+0.25] for each + or - and [+0.25] each (a., b.) for the correct order
' REFERENCE Limerick: LOT-145 . Limerick: LOT-146 . Limerick: LOT-148 K110 292004K105 292004K101 ..(KA's)
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QUESTION 1.02- (l'.00) -
SELECT ONE (1)-of the i following statements which correctly completes the definition of'the. Delayed Neutron Generation tim (1.0)
, It is the length of time between...
- (a.')
-
neutron absorption and subsequent' delayed neutron
.thermalizatio L(b.)- the: production of a delayed neutron and subsequent neutron absorptio ,(c.) the production of delayed neutrons in successive lifetime ' (d.)i ' delayed neutron thermalization and subsequent neutron-absorptio : ANSWE .02 (l'.00) '
(c.) [ 1.0]-
REFERENCE Limerick: LOT-1420, L.0.'s 5 and K102 ..(KA's)
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' QUESTION l'.03 '(1.00)
~
.T.he power' generated by.the reactor at the beginning of core life
-
, comes from U-235-thermal fission and U-238 fast' fission. SELECT ONE (1)' isotope which, later-in core life, produces. larger and
'
larger fractions of power by fissio (1.0)-
(a.) Am-241'
(b;)' Cm-244
'
(c.)L Pu-238
. (d . )l Pu-239 ANSWER -1.03 ' (1.00)
(d.) [+1.0]-
.
REFERENCE-
~ Limerick: LOT-1420, L.0.'s 5.b. and K102- ..(KA's)-
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QUESTION; 11.04.-(2.00)
For EACH of-the TERMS-(a.--f.) on the left MATCH the correct-c DEFINITION on the right. -(Only one correct answer each) (2.0)_
TERMS ANSWER DEFINITIONS ~ Keff- ' Keff = 1
.b.- Critical The fraction of the thermal neutrons that had been born by the decay of delayed
~ Supercritical' The fraction of~ fission neutrons born from the decay of a delayed neutron
..
precurso Excess Reactivity 4. Multiplication factor' dealing with an infinitely large reactor Effective Delayed Nhutron Fraction 5. Neutron population is decreasing from generation
_ _ to generatio ' Shutdown Margin 6. A measure of how far the u reactor is below the critical condition.
l 7. Multiplication factor dealing I
with a realistic reactor. This L includes leakage.
L The amount of additional reactivity built into a reactor to allow it to reach full power for its operating cycle life, o
9. Keff > 1 10. A measure of the departure from criticalit (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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'1; ~ REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page :61 l' <
(7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
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ANSWER 1.04 (2.00)
" '
- :7l j
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' e.- 2'
g OV".10( ,
L l
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-[+0.33] each REFERENCE-
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!
p Limerick: LOT-142 l Limerick: LOT-143 I 292003K104 292002K110~ 292002K107- 292002K111 ..(KA's)-
b '1.05
_ QUESTION (1 ~.00)
SELECT ONE (1) of the following statements which describes the 'l purpose;of the End of Cycle Recirculation Pump Trip (E0C RPT). (1.0) l
\ i (a.)- Adds negative reactivity quickly by void formation to function as a backup to the control ~ rod ,
L (b.) ' Adds negative reactivity _.in addition to the control-rods upon a turbine trip to compensate for the lower rod worth !
at EO 'L(c . ) ' Adds negative reactivity during a turbine trip to l compensate for the longer time required for the control
'
)
rods to reach the high flux region of the cor !
.(d.) Adds negative reactivity to compensate for smaller l beta, effective caused by buildup of plutonium and burnup of ;
uraniu w ;
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ANSWER 1.05 (1.00) .
(c.) [+1.0]
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3 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
lI REFERENCE'
, Limerick: LOT-0040, pp. 12,-20,_21, and 2 I
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292008K126 ..(KA's)
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QUESTION' 1.06 (1.00) l
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'For'EACH of the TWO (2) Thermal Limits given below MATCH the appropriate Failure Mechanism (F1-F3) AND the Limiting Condition (L1-L3). (1.0)L .
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!' THERMAL LIMIT, FAILURE LIMITING MECHANISM CONDITION
. .
.;
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, . Linear Heat Generation Rate .(LHGR)
L', . b .- Average Planar _ Linear Heat !
Generation Ratel(LHGR) j i
' FAILURE'MECHANI'M S LIMITING CONDITION
<
- (F1.)
'
t, lad melting caused _
(L1.) Coolant transition boiling by decay heat & stored -i heat following a LOC ;
t3 ;
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(F2.) Clad cracking caused (L2.) Clad )lastic' strain less from becomming vapo ' than 'T blankete '(F3.) Clad cracking. caused (L3.) Maximum clad temperature !
by high stress from of 2200 degrees F pellet expansio !
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' ANSWER '1.06- (1.00)
~;a (F3.)' +0.25?
i i
(L '+0.25 i (F l+0.25 l- _(L L+0.25l . ,
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E, . REFERENCE- -J
, Limerick: LOT-1380, L.0. .
1293009K119, ..(KA's)
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, QUESTION 1.07 (1.00)
SELECT ONE (1) of the following statements which describes a correct R heat' balance' relationshi (1.0)
.(a.) If'the feedwater temperature used in the. heat balance calculation was HIGHER than the actual feedwater temperature
.
then actual reactor power is HIGHER than calculated reactor
. powe (b.) If the reactor recirculation. pump heat input used in the heat balance calculation was OMITTED then actual reactor powe'r is HIGHER than~ calculated reactor powe '
, (c.) .If the steam flow used in the heat balance calculation was LOWER than'the actual steam flow then actual' reactor power is LOWER than calculated reactor power.-
(d.) If-.the RWCU' return temperature used in the heat balance calculation was HIGHER.than actual RWCU return temperature than actual reactor power is LOWER than calculated reactor powe ANSWER 1.07 (1.00)
-(a.) [+1.0]
. REFERENCE Limerick: LOT-1300, 2.6/3.1'2.3/ K113 293007K111 ..(KA's)
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~(7%) AND COMPONEN15 (11%) (FUNDAMENTALS EXAM)
p g l QUESTION 1.08 ~(1.00)
..$ ELECT ONE (1) of-the following statements which correctly describes the effect of an INCREASE -in the amount of non-condensible gases
- in a steam turbine condense (1.0)
'
'(a.) condenser- pressure decreases (b.) circulating water outlet temperature decreases (c.) . turbine generator megawatt output decreases (d.) condensate subcooling increases ANSWER -1.08 (1.00)
(c.) [+1.0]
F REFERENCE o
- ' Limerick
- -General Electric Thermodynamics Text, Section 6.
!- . Limerick:' LOT-050 K107 ..(KA's)
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- QUESTION: 1.09~ (1.50)
i During a cooldown of the reactor vessel, reactor pressure decreased from'885 psig to 595 psig.in one half hour. DETERMINE the reactor n cooldown rate. ~ (SHOW all of your work). (1.5)
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LANSWER- 1.09 (1.50)
First, convert'psig, to psia, by adding 14.7 ps [+0.50]
Then, referring to the steam tables; 900 psia. = 532 de ;+0.25; 610 psia. = 488 de +0.25 Then calculate using the correct formula; 532 deg.F - 488 deg.F =44 deg.F / half hour [+0.50]
(or 88 deg.F / hour)
(will allow a variance of (+) or (-) 2 degrees)
REFERENCE Limerick: LOT-1100, L.0. '293003K123 ..(KA's)
QUESTION 1.10 (1.00)
. STATE,TWO (2) reasons WHY feedwater heating improves power plant efficienc (1.0)
i ANSWER 1.10 (1.00) The. energy recovered in feed heating would otherwise be lost to the main condenser. [+0.5] Less heat is required from the reactor to reach the desired conditions [+0.5].
REFERENCE Limerick: LOT 1230, L.0. K105 ..(KA's)
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QUESTIONJ 1.11 '(1.00)
Given'the' conditions that the reactor is in start up. .at.the point
- of. heat up but prior to boilin SELECT ONE (1) of the~ following statements which correctly describes Lone: type of heat. transfer from the fue1~ to the reactor coolan (1.0).
~
L(a.) Heat transfer through the fuel pellet is via convectio <
~
'(b.) Heat transfer'across the fuel clad is via convectio '(c.) Heat transfer' through' the laminar water layer is via conductio (d.) Heat-transfer from the laminar film into the bulk of the-coolant is via: conductio ANSWE .11' (1.00)
. (c. ). [+1.0]l l REFERENCE (
, Limerick: LOT-1350, 293007K101- ...(KA's)
QUESTION -1.12 (1.50)
Water hammer is defined at the liquid shock imposed on a piping
. system resulting from a rapid change in (a.) .
.The factors affecting the magnitude of impulse force are
- time,(b.) , and (c.) .
(1.5)
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ANSWER 1.12 .(1.50)
.( flow- ;+0.5; ( mass + ( velocity [+'0 .5] or ucele m h C+a 153 REFERENCE Limerick: LOT-1291, L.0. I and K114 ..(KA's)-
QUESTION .
1.13 (1.50)
' SELECT ONE' (1) of the_ pump curve changes (right column) for each ~of the.three pump modifications (left column).being considered.for a constant speed single centrifugal pump system. (ASSUME ideal conditions). .Not all answers may be used and,some may be used more than onc (1.5)
ANSWER PUMP CURVE CHANGES PUM{ MODIFICATION add one. identical decreased flow rate pump in parallel with decreased head capacity b.- add one: identical
~ increased flow rate pump in series with increased head capacity c.- double the speed double the flow rate
'of the original with the same head pump capacity decreased flow rate with increased head capacity same flow rate with double the head capacity (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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ANSWER- .1.13: -(1.50)
e
<
, .. ;+0.5[
T b'. . ,+ .
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S . . +0 . 5, .
' REFERENC Limerick: SLOT-1290; L.0. 13,..1 .
291004K104- ~293006K113 ..(KA's)'
QUESTION 1.14 . '(2.50) ,
' A condensate. filter / demineralized is' removed from service on high outlet conductivity of 0.1 micromhos/cm and high delta-P of- psi (0.5)
. ~
_ STATE WHY the following parameters are used as. indications that filter demineralized should be removed from service.
< 1 g
High conductivit (1.0).
. High' delta pressure (l'0)
.
ANSWER- .1.14 -(2.50).
a.- 25 [+0.50]
b'. 1.;(Conductivity is affected by chloride, pH and'other g-((a c, Im-impurity concentrations.) High conductivity is a good - or- g g c(ep(eg indication of degraded wate.r qualit [+1.0]
2.-High'. delta-P.is an indication of particulate clogging
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(and reduced filtering efficiency.) [+1.0]
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REFERENCE Limerick: LOT-0520 L.0. 6. c & . Limerick: Tech Specs Bases Po. B 3/4-4-3 291007K108 291007K109 ..(KA's)
QUESTION 1.15 (1.00)
G1VEN the conditions of: recirculation pumps are running at constant minimum speed; low power conditions (20% thermal power).
STATE HOW and WHY core flow varies (INCREASE, DECREASE, REMAINS THE SAME) as reactor power is increased by control rod withdraw (1.0)
ANSWER 1.15 (1.00)
Core flow would increase 1:+0.5] due to an increase in natural circulation [+0.5].
REFERENCE
\ Limerick: LOT-0040, L.0. 1 K112 293008K129 293008K137 ..(KA's)
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REACTOR PRINCIPLES (7%) THERMODYNAMICS- Page 15 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM)
QUESTION 1.16 (3.00)
543.1.A " Start-up of Recirculation System" limits the number of i starts that may be attempted on a recirculation pum a'. With the motor gene'rator and recirculation pump at ambient temperature, (1)
,
pump start (s) may be attempte Anotherof-(2 period pump) start may hasNOT be attempted until a waiting elapse i (1.0) ; With the motor generator and recirculation pump warm from at least fifteen minutes of operation, (1) pumpstart(s)
L may be attempted. Another pum until a waiting period of (2) p start may has NOTelapse be attempted (1.0) SELECT ONE (1) of the following statements which describes ,
what the RESTART limitations protec (1.0)
'
i (1.) The recirc.-pump bearings due to inadequate lube oil I being supplied _on pump starts by the shaft driven l oil pump (2.) The motor generator windings from overheating due to
' ,
increased current flow on pump start '
(3.) The power supply to the recirc pump motors due to possible internal electrical faults causing the pump trip ,
l l (4.) The fluid drive coupling scoop tubes from exceeding !
their design cyclic stress limits.
l ANSWER 1.16 (3.00) (1) two [+0.5]
(2) one hour [+0.5] (1) one [+0.5]
L (2) one hour L+0.5] (2.) [+1.0]
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..t m yu I . . . REACTOR PRINCIPLES-(7%) THERMODYNAMICS Page 1.6' i
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4 (7%) AND COMPONENTS (11%) (FUNDAMENTALS EXAM) l
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REFERENCE'
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- 1. - -Limerick: LOT-0030, Pg. 2 !
2.- Limerick: 543.1.A., !
l p 291005K106 ..(KA's) ,j
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i QUESTION 1.17 (2.00) l DEFINE Net Positive Suction Head (NPSH). (1.0). -l . STATE the effect (INCREASE, DECREASE,.or REMAIN THE SAME)
on available Recirculation Pump NPSH.if HPCI initiates at !
<
70% power AND EXPLAIN WHY this effect occur (1 ~.0) j
i
!
' I
' ANSWER 1.17 -(2.00) j
' 'NPSH is the difference between' total. pressure at the eye of !
- - - the pump. and. saturation pressure. for the liquid. (Will .
!
accept the equation equivilent; NP!H = Psec - PsoC7 [+1.0] i = S-He - He tHz T l LThe NPSH would increase [+0.5] because the temperature of a the feedwater would decrease, thus increasing the subcooling j at'the eye of the. recirculation pump [+0.5].
'
,
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REFERENCE
' . - Limerick: EIH: Heat Transfer and Fluid Flow, Chapter 6, l
'L.O..# 10.9, 10.1 .j
291004K106 ..(KA's) I t
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!22 ' EMERGENCY AND ABNORMAL' PLANT EVOLUTIONS Page 17; l: (27%).
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l l'f ~ QUESTION . 2.01 (1.50) j i
STATE THREE-(3)' symptoms .of a jet' pump-failure 'according to ON-10 (1.5)-
l-; 1
'
' ANSWER 2.01- (1.50) .!
i
' 1.- Unexplained drop in reactor powe I 2 .' . Unexplained rise in core-flow indicatio i
!
1 Unexplained rise in recirculation flow' to the loop containing- {
the defective jet pum ! Unexplained drop in indicated delta pressure on the jet-pump sharing the riser with the defective jet pum Any three (3) [+0.5] each; max. [+1.5]
REFERENCE ,
-!
- L{merick:LON-10 l
1 295001K206 295001K207 202001K301 202001K502 ..'(KA's) 'I i
!
. QUESTION :2.02; (2.50) l i
During refueling operations, the control room operator observes '
~ that the Source Range Monitor count rate is' increasing. STATE l FIVE (5)'immediate actions according to ON-120 " Fuel Handling 1
.
Problems". (2.5)
!
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i l , 52. -EMERGENCY AND' ABNORMAL PLANT EVOLUTIONS' Page 18 (27%)
L LANSWER< -2.02 (2.50)
' Raise the. fuel assembly from the core so it clears the
, . upper grid.
l Evacuate the fuel fico ' Inform supervision.- Ensure all'insertible control rods'are inserte .- ' Notify Health Physic . LNotify. Reactor Engineerin ~ Do not resume fuel _ handling operations without permission
'from Superintendent'of Operations.
.
Any five (5) [+0.5] each; [+2.5]~ ma ! REFERENCE' Limerick:. LOT-155 . Limerick: -ON-12 . Limerick: ON-120 Base G010 295023G010 ..(KA's)
\;
QUESTION 2.03 (1.00)
-Trip Procedure T-101, "RPV Control," is in progres In accordance with the T-200_ series procedures, TWO other systems to be used to Linject boron into the RPV IF the Standby Liquid Control CANNOT be used are (1) and (2) . (1.0)_
ANSWER 2.03 (1.00)
(1)- Control Rod Drive System (CRD) [+0. 5]
(2)- Reactor Water Clean-Up (RWCU) [+0.5]
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1 1 EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 19 !
(27%) l l
ll REFERENCE Limerick: TRIP - T-101,'T-110, T-111, T-11 . Limerick: LOT-156 i
. A110' ..(KA's)
QUESTION 2.04 (2.00)
Pertaining to the TRIP procedures MATCH ONE (1) level in the right I column-to each of the function statements in the left column. EACH level can be used more than onc .(2.0)
i i
FUNCTION STATEMENT ANSWER LEVEL upper limit for reactor +60 i water level per T-10 . +54 i lower limit for reactor water level per T-100 +39 i Top of active fuel +12.5 i ' Entry condition for T-101 i RPy control i [ i . -161 i . -167 i ANSWER 2.04 (2.00)
, If ' [+0.5] each (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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REFERENCEL
'1.: Limerick:J T-100, T-101,-and T-11 . Limerick:~ LOT-1560, L.0. A102 295031A103 295031A104 ...(KA's)
295031A101L l
~
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. QUESTION' '2.05- .(1.00)
-
"
The~ TRIP. Procedure for Scram, T-100, Step S-7,: directs;operatorsJ u ,
to1tripithe' turbine when generator load reaches 50 MWatts. SELECT:
"
ONEL(1) ofsthe,following statements which describes the reason
.for;this ste (1.0)
L(a.)- Assures there is adequate capacity of the. bypass v'alves to handle this load transfer.
- . . (b .~) Guards against'overspeed of the turbine _b H
- generator .from. tripping on reverse power.y preventing th (c.)' Assures lthat~ the End-of Cycle Recirculation Pump trip does
>
not occu (d.) . Prevents MSIV closure upon NSSSS actuation on low' steam 1(nepressur '
ANSWER .2.05 (1.00)
(b.) -[+1.0]
REFERENCE Limerick: T-10 . Limerick: LOT-1560, L.0. K305 295005K304 ..(KA's)
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' iEMERGENCY AND ABNORMAL PLANT' EVOLUTIONS Page 21:
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QUESTION ~2.06: (1.00)-'
Concerning.the Feedwater Control S thereactor. level.setpointto-(a)ystem,:theautomaticsetdownof-(inches)-isinitiated
.
E, :
.
!!
H
.(b)' seconds after the level decreases to (c)- - inche .This avoids: unacceptably (d)' ' water. levels following-
'
- SCRAMS.from high' power. . (1.0)
n f
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2.06
-
ANSWER.- .-(1.00)..
a 17 nches)
- '- b 10 (seconds) '
-
'
c 12.5:(inches) '
.
'(d high~.
-[+0.25]-each! ,
l' ~ REFER'ENCE -
pn ,.
L ~ ,
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. Limerick:c LOT-550, IV. K304 ...(KA's)
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l-QUESTION. 2.07 -(2.00)
t
. FILL in the blanks from.the' list provided.-
To hydraulically disarm the directional control valves for a control rodhydrauliccontrollunitclose.the:(a) and (b)
= isolation valves. To maintain cooling to the control drive
- mechanism the -(c) ,(d)' and cooling isolation-valves should remain.ope (2.0)
LIST insert scram discharge
~
. .c drive ' withdraw
'
1 exhaust 6. . charging
' ANSWER -2.07 (2'.00)
(a) -3.\ drive)
(b) exhaust)
(c) insert)
(d) 4. (withdraw).
(a..and b. may be reversed and c. and d. may be reversed)
[+0.5] each REFERENCE Limerick: ON-104, Rev. 5, p. '
295006K203 ..(KA's)
-QUESTION 2.08 (1.00)
. STATE the TWO concurrent conditions which cause the main turbine to automatically run-back generator load to 22.5% (no-cooling-load rating). (1.0)
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'2.08
'
- ANSNER' (1~.00) -
, ..High'statorcurrent(greaterthan7469' amps) [+0.5]
t Loss of stator cooling (low coolant-inlet pressure or ..
hightemperature.onthestator.bulkcoolant. outlet) [+0.5]
REFERENCE-1.' - Limeribk: LOT-060 . . Limerick: ON-11 . 295005K20 ..(KA's).
' QUESTION :2.09 ' (1. 50)' .
- STATE the THREE (3) entry conditions for T-104 " Radioactivity Release Control." For'EACH entry condition INCLUDE the setpoint or actio (1.5)-
ANSWER- . 2.09' (1.50)
.a.- North Stack Monitor '[+0.25], (exceeds) 1.0 x 10E-2 (uCi/cc) [+0.25] .
b.' South Stack Monitor [+0.25), (exceeds) 1.2 x 10E-2 (uCi/cc) . [+0.25] . Reactor Enclosure Steam Flooding Damper [+0.25]
actuation [+0.25]
REFERENCE Limerick: T-104 295038G011 ..(KA's)
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ti QUESTION: - 2.10 - (3.00) l -l Concerning OT-114 (Rev. 4), " Inadvertent Opening of: a Relief Valve":
~ FILL in the correct response for the following statements from the IMMEDIATE ACTIONS section of OT-11 .P1 ace . loop (s) of suppression pool cooling ~in :
servic (0.5)
'
l
, Rapid shutdown per GP-4 is initiated if the relief
! '
valve CANNOT be shut within minute (0.5)' ll 1 If the SRV cannot be shut within (1) minutes or _;
suppression pool. temperature reaches (2 deg 6 F, place the. reactor mode switch in (3))
positio (1.5) j
~ ~ High Suppression Pool ~ temperature.is an entry condition
. for TRIP procedure- .
(0.5)'
y ANSWER 2.10 (3.00) ! two.(both) [+0.5)
I .5 minutes [+0.5] (1) 2 minutes- '0.51
+
(2) 110 deg F '+0.5' ;
'
(3)-shutdown- !+0.5
, T-102 [+0.5] ,
REFERENCE -
i ' Limerick: 0T-114 l
295013A102 295013G010 295013G011 ..(KA's) l
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- EMERGENCY AND ABNORMAL PLANT E'iOLUTIONS Page 25 (27%)
QUESTION 2.11 (2.00)
Off-Normal procedure, of the reactor' ON-113, recirculation " Loss pumps of RECW,"
if cooling requires is lost for tenthe trip (ping 10)
minute SELECT ONE (1) statement which describes the basis for tripping the recirculation pumps ten (10) minutes after the loss of-REC (1.0)
(1.) To provide time to insert control rods and stay within the constraints of the operating map once the recirculation pumps are trippe (2.) To provide time for the operator to re-establish RECW operabilit (3.) Continued operation of recirculation pumps after ten (10) minutes could damage the pump seal (4.) Continued operation of recirculation pumps after ten (10) minutes could warp the pump shaf SELECT ONE (1) of the following statements which describes thebasisfor.theTEN(10)secondwaitbetweenthe tnpping of the first and second recirculation pump (1.0)
(1.) To prevent initiating reactor level 8 trips due to the level transient that would resul (2.) To prevent placing the reactor in the region of thermal-hydraulic instabilit (3.) To allow time for the first recirculation pump's discharge valve to close prior to exceeding its designed operating delta- (4.) To allow time for the first recirculation pump to coastdown prior to tripping the second pum i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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.__ _ _ , EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 26 (27%)--
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ANSWER 2.11 (2.00) (3.) [+1.0] (1.) [+1.0]
REFERENCE 1.- Limerick: LOT-1550, L.0. . LOT-460 ON-113 295018K303 ..(KA's)
. QUESTION 2.12 (1.00)
SELECT ONE (1) of the following statements which describes the effect of a loss of CRD pump "A" due to low suction pressur (1.0)
(a.) A reactor recirculation pump trip on loss of seal purge flo (b.) Aqtomatic opening of strainer bypass valve and CRD pump "B" auto start when suction pressure is above 25 inches Hg Ab (c.) The suction flow path automatically transferring to the CS (d.) A continuing drop in system pressure until manual action is take ANSWER 2.12 (1.00)
(d.) [+1.0]
l REFERENCE Limerick: LOT-0070, L.O. ' Limerick: LOT-1550, L.0. . Limerick: ON-107 295022A202 ..(KA's)
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l QUESTION 2.13 (2.00)
,- Concerning condenser vacuum, MATCH ONE (1) SETPOINT (right column)
to the AUTOMATIC ACTION (left column). (2.0)
AUTOMATIC ACTION ANSWER SETPOINTS MSIV closure in. Hg Vac, Feed pump turbine trip .7 in. Hg Va Main turbine trip .2 in. Hg Va Bypass valve closure .5 in. Hg Va . 11 in. Hg Va . 8.54 in. Hg Va . 7 in. Hg Va ANSWER g 2.13 (2.00) [+0.5] each; max. [+2.0]
REFERENCE Limerick: LGS OT-116 Limerick: LGS LOT-1540 f
295002K302 295002K303 295002K304 295002K305 ..(KA's) {
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i QUESTION- 2'.14 '(1;00)-
ID
" LIf control is transferred to the' remote shutdown-panel, SELECT ONE (1)-
of the' following RCIC System-interlocks which will remain activ (1.0)
(a.)- ._RCIC turbine trip on overspee (b.). Steam supply valve closure:on high reactor water level.
,
'(c;) - Transfer of suction'from the CST to suppression pool.
o., .
'(d.). Start on low reactor water level (-38 inches).
. ' ANSWER 2.14 (1.00)
" ( a'. ) ' [+1.0]~
R'EFERENCE
.1, Limerick: LOT-0735, L.0. . qmerick: SE- dK201 ..(KA's)
QUESTION 2.15 (1.50)
In accordance with SE-1 " Remote Shutdown" LIST the THREE (3) ways the reactor can be SCRAMMED if evacuation of the main control room is required'BEFORE immediate operator action can be take (1.5)'
/
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- ANSWER' 2.15- (1.50)-
~
' 1. . Trip-the Main Steam Line Radiation Monitors . . ,c SC~I step 4,$./ e g y L 4 y (Aux. Equipment Room) . [+0.5] wereg Open the RPS Breakers at "Y" Panel e (Aux.'EquipmentRoom) [+0.5] .5F-i surd .s+ep /g4.EA eprkeb7 Open the RPS Breakers at the RPS Breaker Panel (InverterRoom1) [+0.5] .path O p $ d s fa g (, g , 3. ef u M e a
,
REFERENCE Limerick: SE-1, Rev. 9, p. K201 ..(KA's)
QUESTION 2.16 (1.00)
OT-112, " Recirculation Pump Trip," states that if both recirculation pumps have tripped, " ensure RWCU is in_ service using two pumps."
SELECT ONE (1) of the following statements which describes the basis foh this step. ASSUME recirculation pump trips occur at
' full power operatio (1.0)
(a.) Increased amount of particulate settle out under these flow conditions. RWCU in-service will remove most of the particulate and reduce the likelihood of instrument plugging on recirculation pump restar (b.) ~ Enhances natural circulation through the core by increasing the delta-T between the core and the bottom head region (c.) Maximizes the amount of cold water removed from the bottom head area and reduces potential for a thermal stratification proble (d.). Prevents idle recirculation loops from experiencing excessive cooldown from seal purge flow in-leakage which can cause excessive thermal stress transients on the loop piping on recirculation pump restar (*****' CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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- EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 30 (27%)
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ANSWER 2.16 (1.00)
- (c.). [+1.0]
REFERENCE Limerick: LOT-1540, L.0. . Limerick: 0T-112 295001G007 ...(KA's).
QUESTION' 2.17 (2.00)
'In accordance with Trip Procedure T-100 " SCRAM," Caution number 20, STATE TWO (2) conditions in which ECCS pumps may be secure (2.0)-
ANSWER 2.17 (2.00) confirmed misoperation [+1.0]
- adequate core cooling is assured [+1.0]
REFERENCE Limerick: T-100 295021A104 ..(KA's)
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, PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 31 RESPONSIBILITIES (10%)
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n QUESTION .3.01 (2.50) !
j STATE FIVE (5) control room indications that could verify that
.
'
an SRV valve is open OTHER THAN increasing suppression pool temperatur (2.5)
ANSWER 3.01 (2.50)
l generator load reduction
. bypass valve closure SRV/ head vent valve leaking alarm ' safety relief valve open alarm . relief valve position lights gteamflow/feedflowmissmatch
.t 2 O"2 Anyfive(%[)L+0.5bac%& h; max. [+2.50]
REFERENCE Limerick: OT-114
\
239001K309 239001K603 ..(KA's)
QUESTION 3.02 (3.00)
Concerning the Reactor Recirculation System: STATE the TWO (2) input signals which automatically initiate the Anticipated Transient Without Scram (ATWS) Recirculation Pump Trip (RPT). (1.0) The THREE (3) input signals which automatically initiate the End Of Cycle RPT are (1) OR
'(2)' WHEN at (3) %
power, turbine first stage pressur (1.5) The ATWS RPT or the E0C RPT opens the two (2) RPT breakers located between the (1) and the (2) . (0.5)
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'
h ANSWER' 3.02 (3.00) IL ' Reactor Water Low Level (-38 inches for 10 seconds).
[+0.5]
2. High Reactor Pressure (1093 psig) [+0.5]
, . Turbine Trip (Stop Valve Closure):.[+0.5]
L
' Lopd Reduction (Control Valve Fast Closure) [+0.5]
. q% Power) [+0.25] at turbine .first stage pressure [+0.25]. (1) MG Set [+0.25]
(2) Recirculation Pump Motor [+0.25]
REFERENCE Limerick: LOT-0040, pp. 12, 20, 21, and 2 K413 202001K414 ..(KA's)
QUESTION .3.03 (2.00)
' STATE F0yR (4) input signals which automatically start the Standby Gas Treatment (SBGT). For EACH of the (4) input signals, INCLUDE the setpoin (2.0)
ANSWER 3.03 (2.00) Reactor Low Level; -38 inches High Drywell Pressure; 1.68 psig Reactor Enclosure High Radiation; 1.35 mrem /H . Reactor Enclosure to Outside Atmosphere Low Delta Pressure;
-0.1 inch H20 Reactor Enclosure to SBGT Connecting Valves Failed; Open Any four (4) [+0.5] each; ([+0.25 for the input signal;
[+0.25] for the setpoint)
6, bk beA Sgl k laMleh j 2.0 mkb 7. . 92d lbs. 4e 0>dsk Smosp l ea 6 bdkSess c'-j N O'I ' ;
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g : REFERENCE E i~ .imerick: LOT-180,'III.W;2,-LO 2-a, b, .. Limerick:~ LOZ-0200,.,III.I.1, LO +
1261000K401 272000K306 ..(KA's)
,
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QUESTION 3.04' (3.00)
Concerning;the^ Recirculation Flow Control System,' MATCH all
applicable conditions-in COLUMN 2 to the respective actions in COLUM (NOTE: not all conditions in Column 2 will be used, and some' actions may have more than one applicable condition)_ (3.0)
COLUMN 1 ~ COLUMN 2 ACTIONS' ANSWER CONDITIONS
- .20% Minimum Speed Total feedwater flow is less than'20% for 15 sec ' ~75% Flow Limiter Reactor water level is less than 27.5 inches
\.. -AND-- '28% Speed Limiter individual feedpump flow :
is less:than 20%
d. . MG Scoop. Tube Lock M/A station is-set at 20%
or less fluid drive oil high temperature - over 210 deg. F Total feedwater flow i greater than 90%
-AND-Less than three condensate pump breakers are closed Lube oil ~ low pressure (10~
psig)
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ANSWER '3.04 (3.00). , 5- I ,'6
' [+0.5] each; max. [+3.0]
,
- REFERENCE;
~
1.- Limerick: LOT-0040, LO 3, 4, and 5, pp. 9,18, and 1 K603 202002K605 202001K101' 202001K119 ..(KA's)
QUESTION' 3.05 (1.00). Reactor water level has decreased to below -129 inches.-
Drywell pressure is 1.2 psig. .The proper.ECCS pump (s)
-is operating. ' ADS' valves will open in second (0.5)
. Rgactor water level has decreased to below -129 inche Drywell pressure is 4.6 psig. The proper ECCS pump (s)
is operating. ADS valves will open in second (0.5)
ANSWER 3.05 (1.00) '525 '+0.5; b .- 105- '+ REFERENCE l Limerick: LOT-0330, pp. 9, 12, and 13, figure 6, L.O. 2, l 5,-and K501 218000A402 ..(KA's) ..
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. PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 35 RESPONSIBILITIES (10%)
QUESTION 3.06 (2.75)
'
For each of the NSSSS group isolations in COLUMN A, SELECT all of the plant conditions in COLUMN B that would initiate isolation signals. Note - not all of the plant conditions in COLUMN B need be used, and some may be used more than once. (ASSUME: the Mode Switch is in RUN) (2.75)
COLUMN A COLUMN B Group Isolations ANSWER Reactor Conditions MSIVs and Steam Drain Low reactor water level Lines (IA) (-42 inches) Main Steam and Reactor Main steam line high Sample Lines (IB) rad (3.5X) RHR Heat Exchanger Steam supply low Vacuum Breaker Lines pressure (700 psig)
(IIC) Reactor Water Cleanup High drywell pressure Lines (III) (1.72) HPCI\. Turbine Exhaust Steam line low pressure Vacuum Breaker Lines (90psig)
(IVB) - ANC-high drywell pressure (1.72 psig) Low condenser vacuum (11.5 psia) Standb liquid control initia ion > Steam line high flow (140%)
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y r,. ANSWER 3.06; (2.75)
a. . '2,3,6,8 ,7 2 c. - '1,T4 , 7 e.- 5- NONE
[+0.25] each correct response
.
REFERENCE-
. Limerick:' LOT-0180, L.O. K101 223002K102 223002K104 223002K107 ..(KA's)'
(QUESTION _ :3.07 (3.00)-
'For EACH of the Reactor Water Levels in COLUMN A, SELECT ALL of the. plant systems'in COLUMN B which use that level for actuation-or'confi pation' input signals. NOTE: there may be more than on answer f6.r each Leve (3.0)~
COLUMN'A . ANSWER (s)- . COLUMN B Reactor Water Plant Systems
. Levels level 8 NSSSS level 3 RPS level 2 ECCS level 1 RRCS Recire. System RCIC Main Turbine (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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I ' ANSWER 3.07 (3.00)
? a. . ;3, 6; 7y4# ,: 3, ' 5; /
" ,3,4,(5)',6
' , 3 e maxim &$ f 0*
REFERENCE Limerick: . LOT.-0050;; LOT-0180, L.) . 2.a; . L)T-0300, . L.O. 4;.
LOT-0315,.L.0. K404 216000K405 216000K406 .216000K407- ..(KA's)
~QdESTION 3.08 (1.00)
SELECT ONE-(1) of the following statements ~ that describes the conditions which will cause~ automatic initiation of. Standby Liquid
~. Control - (SLC) . High RPV pressure (1093 psig) AND: (1.0)
(a.) :lhw reactor water level (-38-inches) AND 118 see timer
~ timed ou ~(b.). no APRM downscale.(4%) AND 118 sec timer timed ou !(c.) low reactor water level (-129 inches) AND 118 see timer timedcou
(d.) no APRM downscale 4%) AND low reactor water level (-38 inches).-
LANSWER 3.08 (1.00)
.L(b. ) [+1.0]
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REFERENCE Limerick:
'
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- 1. :. LOT-0310, L.O. 1 ~211000A308- ..(KA's)
'
QUESTION: 3.09- (1.50)
L ,
.Concerning-Standby Liquid Control System:
- STATE the effect of a loss of. Instrument Air on the control . .
p 1
-' room SLC tank level indicatio (0.5)~ LSTATE'TWO (2) additional SLC control room indications a other than the SLC tank level indicator which can be used to p verify. that the SLC tank level has decrease (1,0)-
ANSWER: 3.0 (1.50). (decreases to) zero [+0. 5]
i .: SLC tank Hi/Lo level alann '+0 . 5'
2).~RunningSLCpumpstrip : +0 REFERENCE Limerick: LOT-310.
'
211000A101 ..(KA's)
{ --
[-
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. QUESTION 13.10 (2.50)
s
.The following plant conditions apply:
Reactor. power 45%;
- Total recirc -loopLflow 35%
Two recirc loop operation Reactor power is;being INCREASED to 75% with control rod manipulation 1 ONLY.. Using the APRM flow-biased power formulas; DETERMINE whether
.this power' increase.would: result'inta.R00 WITHDRAWAL BLOCK," FLOW
- BIASED NEUTRON FLUX- SCRAM, or N0' RESULT. SHOW work.' ASSUME that total; recirculation flow remains constan .
(2.5)-
i ANSWE .10 ~(2.50)
R. Block- .= 0.58W + 50%. -[+0.25]~
= 0.58(35) +.50% = 70.3% [+0.75]
- Scram; = 0.58W + 59%- [+0.25]
=0.58(35)+59%=79.3%' [+0.75]
Wouldrekult.in'aRodWithdrawalBlock [+0.5]
REFERENCE Limerick: LOT-0270, L.0. 2.a.-
2. . Limerick: Technical Specifications 3. K505 215005A104 202001K123 ..(KA's)
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E . QUESTION 3.11 (1.00)
ASSUNING full' power operating, three-element control,.and no
- operator action, SELECT ONE (1) of the following.which would b ~the expected Feedwater Control System (FWCS) response if the-selected level transmitter failed HIG (1.0)
- n (a.)' RFP turbines would' lock due to loss of level signal; input and level would remain approximately the sam '(b.) Steam and feedflow inputs would compensate'for the error .
signal and level would stabilize at a slightly lower leve '
(c.) Level input'would automatically transfer to the other level:
transmitter and level. would remain approximately the sam (d.) RFP turbines' reduce speed in response to high'1evel signal and level wi11' continue to drop unless manual action is'take '
ANSWER '3.11 (1.00)
=(d.) .
[+\ ]
~ REFERENCE 1. . Limerick: LOT-0550, L.0. A203 215002A202 ..(KA's)
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u l: '3.12. .(1.50).
'
-QUESTION
-
Concerning the Reactor Protection System (RPS), SELECT the ONE.(1):
.
most; limiting condition for each of the conditions given. MODE
"
' SWITCH.IS IN RUN. NOTE: Actions may be used more than once or not at all.' .(1.5)
ANSWER . CONDITION ACTIONS APRM "B" flow unit-fails downscal . Rod Block RPS Bus "B" shifts from Normal to - Half Scram
- Alternate power: supply
- Full Scram c. . SCRAM Discharge Volume water. level
.at 15 gallons ANSWER 3.12 '('1. 50)' ' :b.: \
[+0.5] each REFERENCE 1.- Limerick: -LOT-0300, L.0. 4, 5,.and A202 212000A204 212000A214 212000A217 ..(KA's)
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QUESTION 3.13 (1.00)
With the reactor operating at 80% power, the operator. notices a sudden increase in BOTH reactor power AND water leve SELECT ONE (1) of the following events which would explain this transien (1.0)
(a.) An SRV. open (b.) A main condenser bypass valve open (c.) RFP turbine speed increase (d.) "3A" feedwater heater isolate ANSWER 3.13 (1.00)
(c.) [+1.0]
REFERENCE Ljmerick: LOT-0540, L.0. 1 .
259001K301 259001K312 ..(KA's)
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QUESTION 3.14 (1.00)
SELECT ONE (1) of the following which describes the control rod drive hydraulic system response during a SCRA (1.0)
(a.) The scram pilot valve energizes to vent the air off the scram inlet and outlet valve (b.) The scram discharge volume vent and drain air pilot valves energize to vent the air off the scram discharge volume vent and drain valve (c.). If-one of the scram discharge volume vent and drain air pilot valves fails to reposition, the scram discharge volume will remain vented and draine (d.) If any scram pilot valve fails, the action of the backup scram valves will cause the rod associated with the failed scram pilot valve to scra ANSWER 3.14 (1.00)
(d.) [41.0]
REFERENCE Limerick: LOT-0070, L.O. 5 and K107 ..(KA's)
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QUESTION 3.15 (1.50)
FILL IN THE BLANKS for the following statements pertaining to the DC Electrical Distribution Syste Division I and II battery has two batteries; each with cell (0.25) (1) VDC Division and (2) I and II are VDCcapable to loadsoffor supplying up to (3) hours following a loss of AC powe (0.75) Each safeguard battery has a charger which converts VAC to DC voltage to charge the batter (0.25) Battery compartments are continuously exhausted by fans which vent directly to the stac (0.25)
ANSWER 3.15 (l'.50) [+0.25] (i) 125
- +0.25]l (2) 250 +0.25 (3) 4 '0.25[
+ [+0.25]
or 4to North [+0.25]
REFERENCE Limerick: LOT-069 A101 263000K602 263000K102 263000K103 ..(KA's)
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' 3.- PLANT' SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 45 RESPONSIBILITIES (10%) R, 3.16 (1.00)
QUESTION
)
' The fast closure of the Turbine Control Valves is an input to the .j Reactc,r Protection System. : SELECT ONE (1) of the following a statements which describes the basis for this RPS tri (1.0) !
(a.) To prevent rapid depressurization due to a sudden TG load j increase leading to NSSSS closure of MSIV '
'1 (b.)- To' prevent a pressure spike and resulting SRVs opening !
due to the loss of heat sin (c.)- To backup the End of Cycle Recirculation. Pump Trip and i lessen the severity of the pressure transien i (d.) To anticipate the pressure / neutron-flux / heat-flux transien .
<i ANSWER 3.1 (1.00) !
' (d . ) . [+1.0] -l REFERENCE\. Limerick: LOT-0300, L.0. K307 ..(KA's)
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LQUESTION 3.17: l(1.00)
' Pertaining to the Traversing Incore Probe.(TIP) System, FILL-IN-BLANKS
,
usingithe list of tenns belo '
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'a.' The TIP System provides:(1)' indication of neutron- a!
-
flux distribution in'the core at (2) locations.- (0.5) j r
' The TIP. System is used for'(1) calibration.an !
for process computer (2)
'
distribution calculation (0.5)
TERMS LPRM flow-radi al_- -power-IRM axial-i
..
.
.
L JANSWER .3.17 (1.00) i j (1) axial . [+0.25]
(2) radial [+0.25]' l b.' (1) 4.LPRM ' [+0.25]
o- (2) power [+0.25]
REFERENCE
, Limerick: LOT-290
.215001G004 ..(KA's)-
= QUESTION 3.18 (1.50)
l STATE.THREE (3) rooms which use chilled water from the Control Enclosure Chilled Water System for HVAC coolin (1.5)
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ANSWER (3.18? (1.50)
L Control' Room-
= Auxiliary Equipment Room Emergency Switchgear Room
<T Standby Gas Treatment Rooms Battery Rooms
>
,
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/ any: three'- (3) : [+0.5] . each; max . of [+1.5]
I REFERENCE.
E 1.- Limerick:; LOT-450, p. 32., LO :290003G007' 288000K503 ..(KA's)
QUESTION 3~.19 (2.00)
Given the following conditions;
~
. Reactor Thermal Power - 25.7%
Core Flow - 43%
7Recircula\ionPumps-Bothonatminimumspeed Steam Dome Pressure - 765 psig STATE the-Safety Limit _which has been violated AND.briefly F DESCRIBE your answer. . INCLUDE in your description the parameters
.which constitute exceeding'the Safety Limit' AND their value (2.0)
ANSWER 3.19 (2.00)
,
Thermal Power shall not exceed 25% of Rated Thermal Power e- with the reactor vessel steam dotue pressure less than 785 psig or.the core flow less than 10% of rated flow. [+1.0]
Thermal' Power exceeds 25% of Rated Thermal Power [+0.5] and the reactor vessel ~ steam dome is less-than 785 psig [+0.5]. i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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1 Limerick: Technical Specificaitons 2.1.1.- l
, - 2. - Limerick:-- LOT-138 l e
l 290002K507.~ ...(KA's) !
'. .
l-QUESTION 3.20' (3.00) l The_ High Pressure Coolant Injection (HPCI) process lines isolate automatically due to conditions being monitored by the Nuclear 1
'"
,. Steam Supply Shutoff System (NSSSS). For EACH of the THREE (3)
listed conditions which are monitored for NSSSS Group IV isolation, STATE the setpoint AND the reason for the isolatio (3.0)l l
l CONDITION SETPOINT REASON FOR ISOLATION' -l a HPCI Steam Lin (1) (2) i'
delta-Pressure High HPCI' Steam Supply (1) -(2)
Low Pressure c.. HPCI Exhaust Diaphragm- (1)_ (2)
High \Pressure-
'
. ANSWER 3.20: (3.00)
i .. 300% [+0.50] 1 Indicates a steam line ruptur [+0.50] . psig- [+0.50] '
Ensures high enough steam pressure to run turbin l
[+0.50] . 10 psig [+0.50] ; Indicates a rupture disk failur [+0.50]
, REFERENCE-
_ 1. .- .
LIMERICK: LOT 180 II K610 272000K304 ..(KA's)
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! QUESTION 3. 2'1I (2'.00)
STATE FOUR,'(4) items.that must be checked;to assure.that a procedure is.v'alid in accordance with Administrative-Procedure A-7 (Shift Operations)- .
(2.0)
' ANSWE .21 (2.00)
- 1. - Must be stamped in red " Controlled Copy." [+0.5] Station Superintendent"(or or Pimt alternate's) M4 signature (and
- date). -[+0.5] A-Superintendent (or alternate's) signature (and date).
~[+0.5].
4.-
~
Date of use must be'later than the effective dat [+0.5]
REFERENCE
- Limerick: . Lesson.P1an.T-LOT-1570- .2.: Liherick: . Technical Specification 6. . Limeri~ck: -Administrative Procedure A-7, Rev. 8, p. I?,
5. A101- ..(KA's)
QUESTION 3.22 (1.00)
Given the condition that you are performing a non-routine activity on a component using a valid procedure and you reach a step in which you believe you should NOT. follow the procedure as written, STATE TW0.(2)' actions you should take in accordance with 4
~ Administrative Procedure A-7 (Shift Operations). (1.0)
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' PLANT SYSTEMS (38%) AND PLANT-WIDE GENERIC Page 50 RESPONSIBILITIES (10%)
ANSWER- 3.22 (1.00) Place the component into a stable and safe condition. [+0.5]
(accept stable or safe) Notify shift supervisio [+0.5]
REFERENCE Limerick: Administrative Procedure A-7, Rev. 8, p. 13, 5. . Limerick: Lesson Plan LOT-1570, Rev. 002, p. 15, (a.,b.).
29G01A102 ..(KA's)
QUESTION 3.23 (1.50)
According to Administrative Procedure A-41 " Procedure for Control of !'lant Equipment": The person who is responsible for coordinating the individual verification of Blocking Permits application AND removal is the .
(0.5) Permission to release equipment or systems for maintenance or surveillance testing shall be granted by the OR the .
(1.0)
l ANSWER 3.23 (1.50) Chief Operator Shift Superintendent [+0.5]; Control Supervisor [+0.5]
REFERENCE Limerick: A-41 294001K102 ..(KA's)
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( QUESTION 3.24 l(1.00)
shall-receive a-
" !During verbal report shiftfrom turnover, operators'
the previous shift. STATE reporting FOURfor duty (4)'of the general topics required to be covered during the verbal turnove .(1.0). . .
ANSWER 3.24 (1.00)
- General operating condition of the plan . Specific operations performed and difficulties e'ncountered during.the previous shif . Scheduled plant operations.:
l . Equipment outages and maintenance work in progres . . Status of safety _related equipment and condition .. Temporary. circuit authorization Any four (4) for [+0.25] each; max. [+1.0]
REFERENCE-1. - . Limerick: Administrative Procedure A-7, Re ,'" Shift Operations, Section 5.10.1','p. 2 A105 ...(KA's)
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m. .
-QUESTION 3.25 (1.25) USING the Shell Prewarming Timegraph (Attachment 1) and given the conditions below DETERMINE the time at which prewarming the shell will be complet .
Shell pressure increased to 60 psig at 1:30 First stage shell . initial temperature was - 155 deg' at 1:30 (0.75). During shell. warming, STATE the limit for the Maximum heat-up rat (0.5)
ANSWER 3.25 (1.25) :45 a.m. (+/- 5 minutes) [+0.75]
- deg'F/Hr [+0.50]
REFERENCE 1.- L}merick: LOT-0560, " Main Turbine and Auxiliaries." Limerick: LOT-0590, " Electro-Hydraulic Control Logic." Limerick: GP-2, 3. .. Limerick: S01. A108 245000K502 ..(KA's)
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QUESTION. 3.26 .(2.00)
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FILL in the chart to.the exposure limits of HP-102,
" Administrative Dose Limits, Guidelines and. Notification Requirements." ASSUME a completed and signed Fors
, .NRC-4 is on file and.no exposure extensions have been
- authorized;for Quarterly Exposure. If exposure limits L ,do not apply for a particular category, enter NA'for not'
applicable.' .
(2.0)
QUARTER- YEAR
- (MREM) (MREM)
'Whole Body
. .
' Extremity
- l Skin ~ ;
,
ANSWER 3.26 (2.00)- N .\.~2500 mrem '+0. 5' mrem '+0. 5'
- mrem- '+0. 5'
4. .4500 mre l+0.5l
!
~. REFERENCE ,
i
' Limerick: LOT-1705, Rev. O, VI . Limerick: HP-10 i 294001K103 ..(KA's)
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QUESTION 3.27. -(1.50) :
In accordance with Administrative Procedure A-7, STATE THREE (3) '
steps' which the control room operators are authorized and required to enforce in order to limit.and control access to their control
. area of authorit (1.5)
,
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ANSWER 3.2 (1.50)
- 1.- Forbid access to the front or rear of all control panel [+0.5] Require verbal requests to cross the " red" line (surrounding '
their bench board panels). [+0.5] Order the immediate departure (or removal) of any unnecessary people from the control room when a plant emergency exist [+0.5] j REFERENCE Limerick: LOT 1570, . Limerick: Administrative Procedure A-7, K105 ..(KA's)
QUESTION 3.28 (1.00)
Technical Specifications set maximum limits on chlorides ir: the reactor cholant system to prevent damage to materials in contact with the coolan SELECT ONE (1) of the following which describes the basis for this limi (1.0)
(a.) Chlorides catalyze the oxidation of carbon stee (b.) Chlorides cause stress cracking of the stainless stee (c.) Chlorides increase galvanic corrosion at dissimilar metal junction (d.) Chlorides increase the formation of insoluble metallic corrosion product ANSWER 3.28 (1.00)
i (b.) [+1.0]
)
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- f.:p ; Limerick: -(Technica13 Specification y L294001A114' -204000A302- :
..(KA's).
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[? n- l TEST CROSS REFERENCE: Page 1
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OUESTION 'VALUE . REFERENCE
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'1.02: 1.00- 90003-1.03- .1.001 90005~
?1.04 2.00 90006.
- 1.00:
, 1.05f 90017.
L 1. 06 '. - 1.00 .90004
? 1.07- 1.00 .90007-E L1.08 L1.00 9000 '
f 1.09 - 1.50 90009 1.10 .1.00l 190010 1.11- 1.00 90016 L1.12- 1.50 .90001 11.~ 13 1.50 90011
-
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1 1.1 .' 1. 00 ; 90013 f'
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'1.16 3.00-2.00:
90014 1.1 ......
25.00 2.01 11.50 90018 2.02= 2.50 .90019-
- 2.03- ~1.00 90020 2.04' 2.00 90021-2.05- 1.00 190022 2.05' 1.00 90023
' 2.07.; 2.00 =90024 2.08 1.00 90025 2.09' 4 '1.50- 90026 2.10 3.00 90027
' 2.11'- 2.00 90028 L 2.12 1.00 90029 2.13 2.00 90030 2.14 1.00 90031 2.15 1.50 90032 2.16- 1.00 90033 2.17 2.00 90034
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27.00 3.01 2.50 90035 3.02 3.00 90044
.3.03 2.00 90045 3.04 3.00 90046 3.05 1.00 90047 3.06 2.75 90048 3.07 3.00 90049
'3.08 1.00 90050 3.0 .50 90051
.3.10 -2.50 90052 3~11
. 1.00 90053 3.12- 1.50 90054 3.13- 1.00 90055 3.14- '1.00- 90056 3.15' 1.50 90057
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1.00' 90058-r! 3.17' 1.00' 90059-cz
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- 3 '.18 ' '1150 .900601 0 :3.19 '2.001 .90061 5 .
i3.20 '3.00 90062 i; ? '3.21
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3.22: '1.00L 90037a I/ '3.23- '1.50f -90038 F 3.24 1.00: 90039-
.3.25 1.25 '90040 ,
3.26 2. 00 '. :90041-3.27- 1. 50 90042
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i ATTAC.H (6 6tJT NRC REACTOR OPERATOR EXAM (6/12/89) RESPONSES Category'1: Reactor Principles (7%), Thermodynamics (7%) and Components !
(11%) (Fundamentals Exam)
Question 1.1 Additional correct response should be " Direction".
Reference: _ LOT-1291, Page 5, VI. ...... . . . . . ... .............. . . . . . . . . . .
Question 1.12.b/c Additional correct response should be
" Acceleration" or "Ar.ea".
e Reference: LOT-1291, Pages 4 and 5 VI. A.3. a,b . " Acceleration" V. B. 1 " Area" (Similar question SRO 4.12)
.......................................
Question 1.13 As discussed at the Pre. Exam review, this question was changed in order to clarify the intent of the question. Assume ideal conditions was added in order for the examinees to "zero in" on the generic theory behind adding same size centrifugal pumps in series / paralle Ideally, the best you can get by adding a second pump in parallel is double the flow rate at the same head capacity (Assuming EQ system i (resistance to flow) Curve).
Additionally, the best you can get by adding one in series is the same flow rate with double the head capacit The answers on the original draft copy were correct as stated during the Pre-Exam Review and should be as follows: '
c-2 l J
Re ference : LOT-1290, Pages 8 and 9 a - G.5 a and b b - G.6 a and b
'
c - G.8 a (Similar question SRO 4.11 parts b, c, and d)
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- - _ _ - . _ _ _ . -. _- ._ _ _ _ _ _ . - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _
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NRC REACTOR OPERATOR EXAM RESPONSES (6/12/89) - Continued Question 1.14 ' Alternate answer would be " Indication of ION exchange _ depletion" or similar-wording, Alternate answer would be " mechanical blockage" or similar wordin (Similar question SRO 4.13)
. . . .... . . . .................... ...... . .
Question 1,15 As discussed at the Pre-Exam review, this question can be correctly answered different ways depending on assumption First, since the examinees weren't ;
provided an operating map (Power.to. Flow) this
- required them to assume conditions.
Secondly, the question did not state " Limit your answer to initial response".
For these reasons, correct responses to this question are: As written in the answer ke OR Decrease due to two. phase-flow resistance OR
'
\ Increases initially then decreases due to two phase flow resistanc Reference: LOT-0040, Page 25 i
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B.2.b.2) and T. LOT-0040 8 (Similar Question SRO 4.14) l
.. . . .... .......... . . . . .. . ...... ....... .
Question 1.17 a As discussed at the Pre-Exam Review, it was agreed to accept the equation equivalent for NPS A correct equation equivalent may also include the elevation (static) head term and/or the friction head loss ter Reference: LOT.1290, Page 5 D. (Similar question SRO 4.16a)
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NRC REACTOR OPERATOR E)UV4 RESPONSES (6/12/89) - Continued ;
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CateForv 2: . Emergency and Abnormal Plant Evolutions (27%)
Question 2.04.c- Correct response is "8" -
(.161 in.)
Reference: T.101 bases, Page 20
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RC/L.5 Discussion
(Similar SRO Question 5.10) l
.. . ............................. . .. . ... .
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Question 2.0 Correct response is 17 inches, n
Reference: LOT.0550, Page 11, (Similar SRO Question 5.08)
....................................--
Question 2.15 As discussed at the Pre-Exam Review, this question was changed to ask the three ways per SE- Request you consider candidates responding with, similar, correct, wording as to that which is in the answer ke For example, answer key states "Open the RPS breakers at "Y" panel". Similar wording - Open circuit 13 on "Y" pane _ Reference: SE-1, Page 2, Steps 4.5.1, 4.5.2, 4. .\'
(Similar SR0' Question 5.13)
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NRC REACTOR' OPERATOR EXAM RESPONSES -(6/12/89) . Continued Catevory-3: Plant Systems (38%) and Plant-Vide Generic Responsibilities (10%)
Question 3.01 Other correct responses would be the causes of the alarms for the open SRV and similar wording for the relief valve positon lights. These responses are:
.
Acoustic Monitors Tailpipe Temperatures Reference: Alarm Response Cards (ARC's)
ARC.MCR-110 Page 10 ARC-MCR-110 Page 11 (Similar question SRO 6.08)
. ... ..................................
-Question 3.0.2.b.3 Answer key requires:
"30% power at turbine first stage pressure" for full credit. Exam question provided:
% power, turbine first stage pressure. Correct
__ response therefore is 3 Reference: RO Exam Question 3.02.b and Answer Ke \
(Similar Question SRO 6.16 Part b.is correct as stated on SRO Exam / Key)
. .. . . . .................................
Question 3.03 Other correct responses are:
Refuel Area High Rad; 2.0 mrem /hr Refuel Area to Outside Atmosphere Low Delta Pressure; 0.1 inch water Reference: Technical Specifications pages.3/4 3 15, 3/4 3-16, 3/4 3-22 Trip Function Trip Function Table 3.3.2-1 Notation c Trip Setpoint Trip Setpoint (No Similar SRO Question)
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~ NRC REACTOR OPERATOR EXAM RESPONSES'(6/12/89I . Continued Question'3.0 Delete answer f (None) from answer' ke Question 3.06.f did not appear on RO exa It-did appear on SRO exam.. Question 6.1 Reference: RO Exam 3.06.f and answer ke .
(Similar SRO Question 6.11, which is correct as stated.)
.. . . . .. . . .. . .................. . . .
Question 3.0 Additional correct response for Level 3 (+1 inches) is Response 1 (NSSSS). It is an isolation signal for groups IIA and IIB of NSSS Reference: GP.8 Pages 128, 129, and 13 .(Similar SRO Question 6,07.)
.... .. . .. ... .....................
Question 3.1 Other correct response is 48 LGS uses 440 VAC and 480 VAC interchangeabl Reference: LGS FSAR Page 8.3-1 Description 8.3. (No Similar SRO Question)'
. .. . . .. .. .. .....................-
Question 3.18 Additional correct responses are:
-\
Remote Shutdown Panel Room Computer Room Reference: P61D M-78 Sheet 2 of 4 (No Similar SRO Question)
. . .... .. .... ................... -
Question 3.21 For the response " Station Superintendent" (or alternate) the title Plant Manager (or alternate)
is now used at LG Reference: Technical Specifications, Page 6 13, Procedures 6. (Similar SRO Question 6.30)
.............. .... ......... .. . . . ..
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NRC SENIOR REACTOR OPERATOR EXAM (6/12/89) RESPONSES (Continued)
' Question 3.2 For the response " Control' Supervisor" the title
,
applies to the SRO in charge of the Control Roo Other titles-used for this position at LGS are:
. " Room Supervisor"
" Shift Supervisor" Reference: Technical Specifications,-Page.6-1, Responsibility 6. (No Similar SRO Question)
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NRC SENIOR REACTOR OPERATOR EXAM (6/12/89) RESPONSES ,
.1 Category 4: Reactor Principles (7%), Thermodynamics (7%)'and Components i (10%) (Fundamentals Exam) l
Question 4.1 As discussed at the Pre-Exam review, this question was changed in order to clarify the intent of the question. Assume ideal conditions was added in order for the. examinees to "zero in" on the generic theory behind adding same size centrifugal pumps in serias/ paralle Ideally, the best you can get by adding a second pump.in parallel is double the flow rate at the same head capacity (Assuming E0 system (resistance to flow) Curve) .
Additionally, the best you can get by adding one in series is the same flow rate with double the head '
capacit The answers on the original draft copy were correct as stated during the Pre-Exam Review and should be as follows; b-3 d-2 Reference: LOT-1290, Pages 8 and 9 b - G.5 a and b
\ c - G.6 a and b d - G. (Similar question R0 1.13)
.................................
s Question 4.1 Additional correct response should be " Direction".
!
Reference: LOT-1291, Page 5, VI. ..................................
Question 4.12.b/c Additional correct responae should be ;
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" Acceleration" or " Area" Reference: LOT.1291, Pages 4 and 5 VI. A.3, a,b - " Acceleration" V. B. 1 " Area" (Similar question R0 1.12)
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NRC SENIOR REACTOR OPERATOR EXAM RESPONSES (6/12/89) - Continued Question ~4.13 Alternate answer would be " Indication of . ION exchange depletion" or.similar wordin b.2- Alternate answer.would be " mechanical blockage" or similar wordin (Similar question R0 1.14)
............. . . .. .............. . . . .
Question 4.~14 As discussed at the Pre-Exam. review, this question can be correctly answered different ways depending on assumptions. First, since the examinees weren't provided an operating. map (Power-to-Flow) this-required them.to assume condition Secondly, the question did not state " Limit your answer to initial response".
For these reasons, correct responses to this question are: As written in the answer ke OR Decrease due to two-phase-flow resistance l l
OR '1
\ Increases initially then decreases due to two phase flow resistanc Reference: LOT-0040, Page 25 B.2.b.2) and T-LOT-0040-8 (Similar Question R0 1.15)
. ...... ...........................
Question 4.16 As discussed at the Pre. Exam Review, it was agreed to accept the equation equivalent for NPS A correct equation equivalent may also include the elevation (static) head term and/or the friction head loss ter Reference: LOT-1290, Page 5 D. (Similar question RO 1.17a)
. _ _ . . - .
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t NRC SENIOR REACTOR OPERATOR EXAM RESPONSES (6/12/89) Continued
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[. Catenorv-5: Emergency and Abnormal Plant Evolutions (33%)
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L Question 5.0 Correct response is 17 inche Reference: LOT.0550, Page 11, (Similar RO Question 2.06)
. . . . ...............................
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Question 5.1 Correct response is "8" (-161 in.)
Reference: T-101 bases, Page 20 RC/L.5 Discussion (Similar RO Question 2.04)
... ..... . . . . . . . . . . . . . . . . .. . . . . . . . . . . .
Question 5.13 As discussed at the Pre-Exam Review, this question was changed ta_ask the three ways per SE- Request you consider candidates responding with similar, correct, wording as to that which is in the answer ke For example, answer key states "Open the RPS breakers at "Y" panel". Similar wording - Open circuit 13 on "Y" pane \
Reference: SE-1, Page 2, Steps 4.5.1, 4.5.2, 4. (Similar RO Question 2.15)
. ...................................
Question 5.14 Correct response is "To ensure sufficient cooling flow through the core."
Reference: T.115 Bases Step AK-ll Discussion (No Similar RO Question)
_ _ - _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _
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l NRC SENIOR REACTOR OPERATOR EXAM RESPONSES-(6/12/89)~ . Continued ji Question 5.1 As discussed at the Pre-Exam Review, entering T.100 on 1.68 psig drywell is not incorrect, but is not necessary since 1.68 psig requires entry to T.10 It was agreed that response 1 would be added to
_
part c, but would be bracketed (' ) . Therefore, the answer key to part c would read:
(1), 2, 3 Reference: SRO Exam Question 5.16.c and Answer Key (No Similar RO Question)
. . . ...................................
Question 5.20 Basis for this step in ON.119 " Loss of Instrument Air" also states that, a reactor scram, should it occur, would be a less severe transient when initiated at a lower reactor powe Answer 'B', Limit the reactivity transient caused by the closure of the MSIV's, should also be considered as a correct respons Reference: ON-119 Bases Page 3 Step 2. ;
(No Similar RO Question)
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NRC SENIOR REACTOR OPERATOR EXAM RESPONSES (6/12/89) - Continued Catenorv 6: Plant Systems (30%) and Plant. Wide Generic Responsibilities (13%)
Question 6.0 Additional correct response for Level 3-(+1 inches) is Response 1 (NSSSS). It.is an isolation signal for groups IIA'and IIB of NSSS Reference: GP-8 Pages 128, 129, and 13 (Similar RO Question 3.07)
.................................-
Question 6.0 With the information provided, the answer.for part
'b' should be Tech. Spec. 4.8.1.1.2.a.5 vice 4.8.1.1.2.a.4. . Tech. Spec. 4.8.1.1.2.a.5 refers to the load (i.e., 2800 kw).
Secondly, the question as worded, " Surveillance was run to test the "ECCS" actuation signal" may lead the candidate to Tech. Spec. Surveillance Requirement 4.8.1.1.2.e.5, which is an ECCS actuation test signa ;
.
This Surveillance Requirement requires the diesel to run unloaded and operate in standby for 5 minute l Using the data supplied:
\'
Generator Load: 2800 kw in 135 seconds would make that Surveillance Requirement fail, and declare that diesel INOP per Tech. Spec. 4.8.1.1.2. Reference: Technical Specifications Pages 3/4'8 3, 8 4, 8 5 4.8.1.1.2. .8.1.1.2. (No Similar RO Question)
.................................-
Question 6.08 Other correct responses would be the causes of the alarms for the open SRV and similar wording for the relief valve position lights. These responses are:
Acoustic Monitors Tailpipe Temperatures Reference: Alarm Response Cards (ARC's)
ARC.MCR.110 Page 10 ARC-MCR-110 Page 11 (Similar question RO 3.01)
............... - ...... . . . .... - - .
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_. _ __
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NRC SENIOR REACTOR OPERATOR EXAM RESPONSES-(6/12/89f - Continued Question 6.13 Without providing a specific float voltage, the candidate may elect to be conservative and also consider the float voltage.not within the allowable i value for category B which would lead him to choose answer (a) which is:
(a). " Division 1 is INOPERABLE and must be-restored
. to operability within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in 110T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."
Reference: Technical Specifications Pages 3/4 8-10, 8-11, 8-12, 8-13 Table 4.8.21-1 Notes (1) and (3)
Action: On page 3/4 8 10 for INOP Batter ~
(No Similar RO Question)
.......................................
Question 6.30 For the response " Station Superintendent" (or alternate) the title Plant Manager (or alternate)
is now used at LG Reference: Technical Specifications, Page 6-13, Procedures 6. (Similar SRO Question 3.21)
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Limerick Exam.- 6/12 - 6/16/89'
R0 Written Exam
' Facility Comments and Resolutions L
. Question 1.12.a.
L The. question asks- for a completion of-the definition for Water Hammer. The l
stem of the question is verbatim from the-facility training material, .I. A.1.,
Partial credit will be considered for each response which describes a Ate of water hammer AND which is grammatically correct; for example, the facility suggested " direction."
Question 1.12 b./ The question is seeking three factors which affect the magnitude of the impulse force. Time is given. The facility training material' formula for impulse -
force'is F=M x V/t. The correct response for full credit, therefore, would be " mass" and " velocity."
Partial credit [+0.25] for " acceleration" will be given only if the second response is " mass" completing the formula F=M Answer key is modifie Reject the facility recommendation to allow the additional response " area."
V.B.1 is a review of the calculation of mass of water in a pipe in a discussion of the momentum of wate Question 1.13
. Accepit facility comment, correct answers will be change to a. 3, b. 5, c. Answer key is modifie Question 1.14 Comment accepted. Answer to include alternate response " indication of ion exchange depletion." Answer key is modifie Question 1.14 Answers similar to " particulate clogging" only will be accepte Question 1.15 As agreed to in the pre-exam review, the candidates response will be graded as per the answer key. Other responses such as facility recommendation 2 and 3 wilI be considered for full credit if assumptions are stated and the response is correct, i.e. flow change is stated AND the reason for the change is stated and. correc Question 1.17 Comment accepted. Answer key is modified to change the equation equivalent for NPSH to NPSH= Hp-Hf-Hvp-H _ _
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Page 2 Limerick R0 Written Exam Facility Comments and Resolutions Question 2.04 Comment accepted. Answer key is changed to "8" (-161") for correct respons Question 2.06 Comment accepted. Answer key is changed to "17 (inches)" for correct' respons Question 2.15 Comment accepted. Responses in agreement with SE-1 steps 4.5.1, 4.5.2, and 4.5.3 will be accepted. Answer key modified to include comment Question 3.01 Comment accepted. Added " Acoustic Monitors" and " Tailpipe Temperature" as acceptable response Question 3.02 Comment accepted. Added parentheses around % power in the answer ke Question 3.03 Comment accepted. Answer key modified to include (1) Refuel Area High Radiation; 2.0 mrem /hr, and (2) Refueling Area to Outside Atmosphere Low Delta Press'ure; (-) 0.1 inch wate Question 3.06 Comment accepted. Removed answer "f. None" from ke Question 3.07 '
Comment accepted. Added 1. NSSSS" for level 3. to answer key. This makes each response worth 3.00/13 = [+0.231] with a maximum question value of [+3.0].
Question 3.15 Comment accepte "480" will be accepted as equivalent to 440. Answer key modifie Question 3.18 Comment accepte Added the Remote Shutdown Panel Room and Computer Room to the answer ke _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -
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- Page 3 Limeridk R0 Written Exam-
- Facility Comments.and Resolutions-
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M' Question'3.21'
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Comment accepted. Modified answer. key to add "or Plant Manager."
. Question 3.23 b.-
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Comment! accepted. Answer key is' modified to include " Room Supervisor" or-i! " Shift Supervisor" as a correct respons Comment to Facility: It is a poor administrative policy to allow two or'thre <
- different designations (titles) apply:to one position.
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Ku, , ' Limerick ~ Exam 1--6/12L .6/16/89 R b .SR0 Written' Exam' ' j (' ? Facility Comments and Resolutions l lx [
y . Question 4.11'
,
Comment Accepted. Correct answers will be changed to a.3, b.5,' key,is modified.-
Answer l
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Question 4.12~a. .
-
Comment is not applicable to the SR0 exa g
!
. Question 4.12 b./ .
-Comment; refers.to'4.12'a. and.b.'~of SR0 exa The question is-seeking three' factors which affect the magnitude'of the impulse !
force.. Time'is given. The facility' training material formula for impulse j force is F=M x V/t. 'The correct response for full credit, therefore,.would- l be " mass" and'" velocity." , . .' l
.
Partial credit [+0.25];for " acceleration" will.be given only if the second .)
' response is " mass" completing the formula F=M Answer key is modifie i
-
{
. Reject the facility recommendation to allow the additional response " area."
'
V.B.1,is:a' review o.f the calculation of mass of water in a-pipe in a discussion
'of.the momentum of,ate w Questfon4.13 Comment accepted. Answer to include alternate response " indication.of' ion =
'
exchange depletion." ' Answer key is modifie '
Question 4.13. ;
'i Answers similar to " particulate clogging" only will be accepte !
-Question'4.14 As, agreed to in the. pre-exam review, the candidate: response will be graded
.as per the answer key. Other responses such as facility recommendation 2 and
>.
3 will be considered for full credit if assumptions are stated and the response is correct, i.e. flow change is stated AND the reason for the change is stated and correc Question 4.16 Comment accepted. Answer key is modified to change the equation equivalent for NPSH to NPSH= Hp-Hf-Hvp-H I
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? Limerick.SRO Written Exam y facility Comments and Resolutions .
. Question 5.08 Comment accepted. Answer key is chanpd to "17 (inches)" for correct response.
1, LQuestion 5.10 Comment accepte'.d Answer key is changed to "8" (-161") for correct. respons Question 5.13-1 Comment accepte Responses in agreement with SE-1 steps-4.5.1, 4.5.2, and 4.5.3 will be accepted. . Answer key modified to include comment . Question 5.14 Comment rejected. LOT-1560,' p. 20 of 1 and T-115 Bases, Step AK-14 1(Discussici), state that the reason for maintaining reactor pressure 50 psig above' suppression pool pressure.is to assure the SRVs stay ope Question 5'14 b.-
.
Comment accepted. ' Answer key will be modified to reflect the chang Question 5.16 Commept accepted. ' Answer key will be modified to reflect the change. This does hot affect th'e point values for he questio . Question 5.20'
Comment' rej ecte The explanation of Step 2.4.1 in ON-119 makes no mention
'of a reactivity transien there is no mention of concerns about reactivity transient being caused by the closure of the MSIVs anywhere in the procedur A reactivity transient would only occur on a fast closure of.the MSIVs'which is-not addressed _in the Bases. The only item.in ON-119 that concerns the MSIVs is_a note on page 8 of 1 Note: MSIVs may begin to drift clos This. note applies'to the ability to maintain the main condenser as a heat sink. The MSIVs are closed in Step 2.7.2 a and b to protect the turbine and
<
condenser from overpressurizatio Quest' ion'6.02 Comment accepted. Added 1. NSSSS" for level 3. to answer key. This makes each response worth 3.00/13 = [+0.231] with a maximum question value of [+3.0].
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Page 3 l Limerick SRO Written Exam L Facility Comments and Resolutions Question 6.07 Acc'ept the change of T.S. 4.8.1.1.2. Answer will be modified to reflect the chang Reject the. suggestion of T.S. 4.8.1. . There is no mention of generator load in that surveillance requiremen Question 6.08 Comment accepted. Added " Acoustic Monitors" and." Tailpipe Temperature" as acceptable response Question 6.13 Comment accepted. Answer key will be modified to reflect the change.
l Question 6.30
! Comment accepted. Modified answer. key to add "or Plant Manager."
Comment to Facility: For exam preparation in the future, it would be helpful if the Bases for the TRIP Procedures could be included in the reference materia \
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. Attachment 5
SIMULATION FACILITY REPORT Facility Licensee: Limerick Unit-1&2 Facility Docket No.: 50-352/50-353
'Operat'ing TAsts Administered on: June 13-15, 1989 This form is to be used'only to report observations. -These observations do
- not constitute audit or inspection findings and are not, without further-verification'and review, indicative of non-compliance with 10 CFR 55.45(b).
,These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observation During the conduct of the simulator portion of the operating tests, the following. items were observed (if none, so state):.
r ITEM DESCRIPTION ATWS During the ATWS scenario Reactor Power was 63% and
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turbine was tripped, only 8 bypass valves were open (out of nine) and the SRV were closed after only going open for a short period of tine.
L Dry Well With both Reactor Recirculation Pump. seals lost and the recire. pump secured Dry Well pressure never increased.
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