IR 05000341/1985001
| ML20129A458 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/17/1985 |
| From: | Lang T, Levy I, Mcmillen J, Sly G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20129A453 | List: |
| References | |
| 50-341-85-01, 50-341-85-1, NUDOCS 8506040613 | |
| Download: ML20129A458 (76) | |
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i 4 U.S. NUCLEAR REGULATORY COMISSION
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REGION III
Report'No.'85001(DRS).
~ Docket'No. 50-341 License No.'CPPR-87 x ' Licensee: - The Detroit Edison Company 6400 North Dixie Highway-Newport, MI_48166 Facility Name: Fermi Nuclear' Power Station
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. Examination Administered At: Fermi Nucler Power Station-
-Examination Conducted: February 12, 1985 W /
Examiners: T. Lang J~' /4-83 Date
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. )] hk hLevy,PNL T
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Date / 3 1i jf.Sy,PNL h ' '
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gh. l s Approved By: ll. I.-McMillen, Chief 3//7/75
)perator Licensing Section Dpe /
Examination Summary
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Examination administered on February 12, 1985 (Report No. 85001)(DRS))
Examinations were administered to one Senior Reactor Operator, seven Reactor Operators and three Instructor candidate ~
Results: Two instructors'and all of the Reactor Operators passed the examinatio L.
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' REPORT DETAILS-
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L1 1 n Exami'ers . ,
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T. Lang, NRC I._ Levy,'PNL
- G.' Sly; PNL Examination Review Meetina
- R. Bouinet S.' Heard
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- E.-Preston G. Overbeck m
During .the. review very few comments were made to the SRO examinatio . They were.so few that they were incorporated directly to the answer ke . R0 examination comments are detailed in Attachment . Exit Meeting
' The exit meeting was conducted satisfactorily and the utility was
- : informed of the candidatesLwho clearly passed the oral / operatin examinatio .
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ATTACHMENT A
'Coment Why are U235 and U238 mentione RESPONSE References to U-235, U-238, and Pu-239 in-the answer key were incorporated to_more fully evaluate candidate's response. Candidate was not required to mention' individual isotopes, but only that Resonance capture was a negative' reactivity effec Comment. ' Answer: May take it out further than thi We would end up back in the source range until cooldown brought Rx critical agai RESPONSE-Facility coment has no bearing on grading of question. No action take Coment Answer: The process computer calculates Quality and-Boiling length based on Core Flow, Inlet Subcooling, Thermal Power and Pressure. ' You may get these as answers also based on " measurement"
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in questio RESPONSE The comment is valid, Quality and Boiling length are not measurable parameters but calculated. The answer key has been modified to read: Core Flow Inlet Subcooling . Thermal Pouer Pressur Any two (2) of the above will be acceptable for full credi Comment Answer: Setpoint 120 ga Allowable '160 ga for scra T.S. chang RESPONSE Since 100 inches is comparable to 120 gallons in the scram discharge volume, either answer is acceptable for the scram setpoin . REFERENCE: Tech. Spec.. Table 2.2.1.1 Final Draf Coment 2.-2 Answer: May see reference to maximum 3" level deviation if all reference leg inside drywell flashe .
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RESPONSE Comment was' accepted, but answer key not modified because facility conenent was; considered not the most optimum answer and. is 'self explanatory if given by a~ candidat Comment' Answer 4: May also see.' reference to-Iodine dissolution in wate ~RESPONSE e Comment _was considered a secondary effect of the condensation process and was not conside. red necessary.for full credi Comment Answer:. May see single coincidence or 1/18 for whole -
,. ' syste RESPONSE- iSince-question' requests the trip logic for EACH Neutron Monitoring-System. a response of " single coincidence" is incomplete unless accompanied with 1Lof,18 or listed once for.each system. Answer key was modified to include single coincidence and l'of 18 or single
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coincidence -'SRM, IRM, APRM.-
- Comment.2.6' Answer: Some may specify the power suppl RESPONSE Comment ignored because question requests " power supplies." If
. candidate provides bus -information it will be graded according to supplied reference materials as to correct or incorrec Comment . Answer: CST Level Setpoint Setpoint- Allowable value-3" In may see 24" above bottom 27"'above bottom or 0" indicated RESPONSE ,
Values were verified with references provided and answer key was modified to include either answer 10,000 gallons or 3" indicated or 27" above bottom for the CST level setpoin REFERENCE: -Tech. Spec. Table 3.3.3-2, pg. 3/4 3-27 Comment Answer: It means what you say, but you'may hear:
Rod not selected and odd picked up or even dropped ou Timer timed out and even not picked u A, _
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RESPONSE- Rephrase of answer key was found to_be acceptable and answer key was
' modified to.use:the " trained" terminolog Comment Answer: Lower-Higher Same 4 L-3 trip back to 195" -
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-RESPONSE Comment was considered, but should have no effect on responses because candidates were directed to respond with increase, decrease, or remain the same not higher, lower or same. No action take Comment Answer: Should say:
73% CV Demand We normally accept 5% if SRV opens at less than set pressur RESPONSE Values (i.e., 73% and 5%)'were included to provide clarification for the grader and not required for full credit. The answer key had, 78%
and 7% respectively and was modified to reflect the correct values even though they-were not require Comment Answer: 1) Downscale not operable unless companion-IRM upscal ) I interpreted question to mean which output functions generate trips:
Answers could vary from INOP 4 - fixed / thermal / / Rod upsc/ upsc/ /8 locks 24 - (same for each channel)
Answer: LPRM's not part of trip circui Input to trip circuit onl Fixed Upscale Answer: Fixed upscale / Thermal Upscale /INOP
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- RESPONSE _The question elidits a number of responses depending upon the interpretation of the term " trip circuits. Since this is true,
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4 ; fixed (neutron). upscale thermal upscal . s
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. Rod-block and alarms will not be considered trip circuits but no
" ' credit will be removed if give The answer key was modified to read fixed upscale or neutron upscal c. - The correct response'should be the same as.in part a. The downscale trip circuit will input into the RPS system. . Rod blocks and alarms do not and credit was taken off if give Answer: -Assume below LPSP.-
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Comment RSCS group notch contro One notch other wa RESPONSE 3.6
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If candidate stated that.he was below the LPSP an insert response of
. one notch'is correct. Answer key was' modified to reflect the response.
. Comment Answer: If only one heater drain isolates, No Action-
.if both, your answer correct.
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RESPONSE ' During the examination the candidates weru told that only one heater
- _ drain had isolated, therefore, no action is the correct response and l- the answer key has been change Comment Answer: On LOP 1) Load Shed f.
2) DG Start
L (6 and 7 BKRs open)
4) DG BKR closes (s and volts)
5) Non LOCA loads sequenced on:
L DW Cooling fans L DG Aux. ~ Valve Power L- ,
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'2) -DG continue to.run
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3) Load sequence RHR/CS et L Following LOCA if. LOP-
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_ 1) Load Shed
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TRESPONSE 'The facility comment'is just clarification of the proper respons No action take The 10 minute time delay was eliminated from the answer key. This changed.the answer key to "yes (0.25),.if you have adequate core cooling or indication of system malfunction-(0.5)."
Comment' Answer: Or voiding causes power reduction while injecting Boro Level
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RESPONSE' Facility. comment was considered correct.and added to answer ke Both responses were necessary for full credi No action taken " indicated zero" is throughout~ procedure and should be known.-
Comment Question should state during changing operatio RESPONSE Due to the vagueness of the wording of the question, credit was given for both 1) when the nitrogen purge gas remains constant, or 2) when no'more fluid'is observed exiting the drain valve.
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138*F Drywell 95'F Torus
. 1.93 psig
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' RESPONSE 4.8-Comment reflects recent Technical Specification changes and the answer
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key.has_been modifie ~
Additional _ facility' comments received February 26, 198 ~
Comment : Answer: This question postulates a loss of feedwater heating as the reason for the moderator temp reduction:of 75*F during a startu Areas of confusion: A loss of feedwater heating would lead the operators to believe that we are at
. power, as such no change in moderator temperature would occur due to pressure control. We would see a change in inlet subcooling which would cause a power increase. This would not be viewed as a heatup or a cooldown rather it would be a-
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power excursio , In another view a 75' reduction in moderator temperature would result in a depressuriza -
tion and resultant closure of the MSIV's
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-and subsequent scram again a transient but no net-heatup or cooldow Additionally, the second part'of the q"uestion asks "which limit is more restrictiv While we all know that cooldown stresses are more limiting, in actuality'our heatup limit is more restrictive than it needs to be since it-is equal to the cooldown limit. Either answer with an explanation would of course be correc RESPONSE 1.3 i
- The question'alicits a response of, heatup or cooldown not pressure or power as the facility comment suggests. It was also clarified
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-during the exam that the core inlet enthalpy changed in an amount
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< No change was made to answer key because during the exam it was
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Comment _Regarding the' question dealing with removal of a sourc from a level of 6000 cps:
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.The correct answer per'your key was a "non-linear
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. The change in neutron level would describe a linear
- 4 plot on the SRM's since they are log scale b. ~ The neut'ron populatidn will' stabilize at a very low
. level (probably below our indication) but it will stabilize due to intrinsic source RESPONSE The question requested the MOST correct answer which is "2" non-linear progression. It'did not soTTcTt what the' operator saw on his script
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recorder or meter nor did it solicit an evaluation of final countrate
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with " intrinsic" sources. . Sources and subcritical multiplication are
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very basic fundamental area of reactor theory and the candidate should be' cognizant in this area. No action take Comment 1.12 Regarding the question dealing with Thermal Neutrons and their ability to produce embrittlement:
Forstainlessgteelsexposedtoathermalreactor 2 2 fluence of 10 N/cm -sec, the tensile properties show some increase in ultimate strength, an almost three fold gain in the yield strength and a drop of about 1/3 ductilit Thought to be due to trace Boron impuritie Ref: Basic Nuc. Rx En Foster & Wright Pg. 355 Therefore while thermal neutrons do not cause direct damage, there is some indirect damag Also defect production as a result of thermal neutron irradiation is largely due to the recoil of nuclei that have absorbed a thermal neutron and emitted one or more T' Additionally while the damage due to fast neutrons is on the order of 90 times as great as thermal, the answer is
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RESPONSE.1.12 Since the question requested an explanation of why the statement was true or false, credit would have to be given for a true response if properly justifie The answer key was not. modified because, according to your own admission, thermal neutrons "....do not cause direct damage...." and
" damage due to fast neutrons is on the order of 90 times as great as
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Comment Regarding the question dealing with what four (4) things are logged in the NSO's log when the reactor is declared critical, you may find that the examinees listed only four (4) items, i.e., Rod Number Rod Notch Rod Sequence This was specified as one item Rod Group in your ke . Period Temperature Time Four of the above, they may not have lumped all of the Rod Parameters into one grou RESPONSE The question states which items MUST be logged, these are: Temperature Time Period ' Critical Rod Position The other items listed rod number, rod notch, rod sequence, and rod group are considered to be a subset of rod position (i.e., you could in the same token separate time into hour, minute, second). But due to the non-explicit wording of the question partial credit will be given if the candidate broke the critical rod into its subcomponent ,
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MASTER COPY QUESTIONS U. S. HUCLEAR REGULA10RY C0fflSS10N REACTOR OPERATOR LICENSE EXAlllNATION Facility: FERMI-2 Reactor Type: BWR
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Date Administered: 2-12-85 Examiner: GA SLY
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INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write ansvers on one side onl Staple nuestion sheet on top of the answer sheet. Points for cach question
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are indicated in parenthesis af ter the nuestion. The passing grade requires at least 70% in each category end a final grade of at least 80%. Examination
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. papers will be picked up s!x (6) hours after the examination start Category 1 of Candide W s % of Value Total Score Cat. Value Ca tegory 25 25 - 1. Principles of Nuclear Power Plant Operation, Therwodynamics,
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Heat Transfer and Fluid Flow 25 25 2. Plant Design including Safety and Emergency Systems 25 25 3. Instruments and Controls
25 25 4. Procedures: Normal, Ahnormal, Emergency, and Radiological Control 100 TOTALS Final Grade %
A11. work done on this examination is my own; I have neither given nor received ai .
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAlllHATION Facility: FERMI-2 Reactor Type: BWR Date Administered: 2-12-85 Examiner: GA SLY
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Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side onl Staple ouestion sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the ouestion. 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 start Category % of Candidate's % of A Value Total Score Cat. Value Category 25 25 1. Principles of Nuclear Power Plant Operation, Thermodynamics,
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Heat Transfer and Fluid Flow 25 25 2. Plant Design Including Safety and Emergency Systems
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25 25 3. Instruments and Controls 25 25 4. Procedures: Hermal, Ahnormal ,
Emergency, and Radiological Control 100 TOTALS Final Grade %
All work done on this examination is my own; I have neither given nor received ai ,
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l PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER
~ AND FLUID FLOW- U is a major component of the fuel in a BWR: What two neutron reactions . occur with the 238U during core Ufe? (0.5)
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dependence of Xe poisoning. All curves are to represent the-effects of a rapid power change. The' equilibrium value of Xe is 100% at full powe Curve 1: The curve that results for a power reduction frorr 100% to zero at time equal zero. Show the curve for 30 to 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (1.0) Curve 2: The trace that results from the reactor being returned to full power 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdow (1.0) Curve 3: The curve produced, if instead of at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,'the reactor was not restarted until 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after shutdow (1.0)
alt s-During causes thea core reacto[r ' "tartup, temperature the loss of a feedwater to change heater 7.5'F in a ten (10) string
minute periW( Assumi the coolant temperature was stable from that point on), State whether this was a heatup or cooldown concern and if the heatup or coolcown limit was exceece (1.0)
' Which 1 is more/'hn't/[ctive restr heatup or cooldown?Why? (1.0)
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1.4 .The reactor. is stable with K effective equal to 0.95 and a stable neutron countrate of 6000 neutrons per generatio ~
a.- If all sources were instantaneously removed would the neutron level (choose the most correct answer): (1.0)
1. -decrease in a linear progression 2. decrease in a non-linear progression .
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3. _ stay the same due to continuous subcritical multiplication of existing neutrons 4. . decrease to constant neutron level equal to the maximum neutron level minus the source strengt If the source strength were doubled the critical rod .
position .(Keff=1) would increase, decrease, not chance?
Explai (1.0) During a startup, a 90 second period is establi she Ideally how much time would be reouired for an IRM signal to go from 50 on Range 2 to 20 of Range 57 (1.0)
g- If no action is taken by .the operator (other than changing rg ranges), what would happen to the power level? Include in your oiscussion the reasons for any changes in the power level 'or the reactor period, and the IRM ranges on which these changes would take plac (2.0)
yv yt Since MCPR is not a directly measurable parameter, what are two (2) measurement core parameters needed by the process computer to calculate MCPR? (1.0)
ki For each of the following conditions, state whether it will cause an increase, decrease or have no effect on critical powe ,
Assume all otner parameters are constan (2.0) decrease in reactor power increase in inlet subcooling increase in core flow location of axial power peak moves up in the cor .
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3 Arrange the following heat transfer flow regimes in order of
increasing heat flu (1.25)
l' . Mist Flo . Single Phas . Bubbly Flo . Annular Flo . Slug Flo Between which two regimes does DNB occur? (0.5) You have just completed a power maneuver from 70%~ to 90% reacter power by increasing recirculation flo State whether the- *
following would increase, decrease, stay the same. Why? Pressure drop in the main steam line (1.0) Average void fractio (1.0)
1.10 The steam jet air ejectors and the recirculation system jet pumps operate on the same principl Briefly explain their operatio (1.25)
I If the' main condenser and associated systems were absolutely
' air ' tight, would there be any need for the SJAE's during (high) power operation? Briefly explai (1.25)
1.11 Choose the most correct answer for control rod density if the-reactor reaches criticality at a point where 63 control rods have been withdrawn to notch position 48 and 4 control rods have been withdrawn to notch position 2 (1.0) % % % %
1.12 Explain briefly why the statement.is True or Fals High levels of thermal neutron irradiation of the reactor vessel causes changes in the mechanical properties of the steel (1.0)
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1.13 Explain whether the overall plant efficiency (MWe/MWt) will
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increase, decrease, or not change due to an isolation at 100%
power of the RWCU syste '
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i l.14 Assume ~ that .the reactor is being started up from' cold shutdown and a rod drop accident occurs. Of doppler, _ void, and moderator coefficients, , which will 'act first, second, and third to limit the rapid power rise?
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5 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS L - 2.1 L With regard- to .the Control Rod Drive (CRD) System: What design ~ feature allows a reactor -scram if the reactor pressure exceeds the . accumulator pressure or charging water pressure?.
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~b.- During normal operations the " Accumulator Trouble" ' alarm illuminates. List the two (2) physical parameters which can cause this alar (1.0)
- 9N Three (3) distinct trips are initiated based on the level in -
3 .. the scram ~ instrument volume. List the trips, setpoints and the actions which occu (1.5) State whether each incident would increase, decrease, or not change indicated _ leve Provide acequate support _for your respons i Incident 1: Leak in reference le (1.0)
- Incident 2: Flashing of'the condensing po ( 1.0)
c.. Incioent 3: Temperature increase inside drywell .
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'(1.0) Concerning the effgas Systen:
c a.' State the function (s) of the following offgas components: (2.0)
.., Ring water buffer tank Hold-up pipe
, Sandfilter Condenser (not main condenser) Arrange the above components in order of flow from the main conoenser to Reactor bldg. ven (1.0)
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6 Concerning the Neutron Monitoring System: What is the range (of output indicator) for the SRM system? (0.5) If an SRM detector loses detector vol tage will the sensitivity increase, decrease, or stay the same? (0.5)
. What minimum . number of LPRM inputs are required to an APRM channel to keep it from being declared inop? (0.5)
VM What ~ is the RPS trip logic for each Neutron ' Monitoring system (SRM, IRM, and APRM) with the shorting links removed and tha mode switch in Refuel? (1.5)
. How would the RWCU system respond to: high non-regenerative heat exchanger discharge temperatur (1.0) high differential flow (1.0)
Include system actions and components in your respons .- Concerning the Reactor Protection System (RPS): List the normal and alternate power supplies to the Reactor Protection System distribution cabine (Incit:de any major differences.) (1,0) The backup scram valves are energize or de-energized to operate in the scram mod (0.5)
( Explain how your answer to part "b" would accomplish the scra (1.0) Explain how decay heat is removed from the core following a reactor scram beause of a complete loss of the grid and the diesel generators have failed to star (1.0)
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7 If the SLC system keylock' is turned - to the " START Sys B" position: How many squib valves fire? (0.5) List six (6) control room indications that indicate the SLC system is initiate (2.0)
2.9 _ Concerning the HPCI system: What is the purpose (s) for the drain pots in the steam supply to turbine line and the turbine exhaust ifne? , (0.5) Upon low CST Level or high Suppression Pool level the HPCI pump suction auto shifts from the CST to the Suppression Pool . What is the CST level and Suppression Pool level that this switch occurs? (1.0) What is the valving sequence for switching the valve line up between the CST and Supp~ression Pool? (0.5)
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2.10 List- the three (3) Rod Block Monitor System bypasses and describe the purpose of eac (1.5)-
2.11 Concerning the Emergency Equipment Cooling Water System (EECW):
' Choose the gas that is used to pressurize the Emergency Equipment Cooling Water System makeup tan (1.0) air nitrogen argon hydroge Why is the tank pressurized? (0.5)
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_ INSTRUMENTS AND CONTROL -
. Concerning the Reactor Manual Control System (RMCS);
f' t - What causes a rod drift light to annunciate? (1.0) Can an overtravel alarm be received if the control rod is connected to its drive unit (yes or g)? (0.5)
' While operating at 80% power and in 3-element control (Level . A Controlling), would the final vessel level increase, decrease, or remdin the same in response to the following events: *
gv'g failure of an SRV full ope (0.5)
y failure of feed flow transmitter (downscale). (0.5)
i failure of a BPV full ope (0.5) level A indicator fails full scal (0.5)
., As condenser vacuum decreases from a normal operating vacuum to atmospheric pressure, what interlocks, trips, or alarms are expected and what are the setpoints for each? [Five (5) of the seven (7) reglared for full credit]. (2.0)
k . What would be the response of the Governor Pressure Control System, if a SRV were to fail open from 80% power? Include in d your answer initiating events, responses, and final stable condition (1.5)
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. Answer the following questions about the Power Range Monitoring v
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% 6 0s,, How many trip 5ciretrits are there? (0.5)
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p .' Which trip circuit is calibrated to ensure the maximum LHGR S4 '
is not exceeded? (0.5)
L'nop l M b Which trips are input to RPS? (1.0)
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time without tripping the RPS channel? (0.5)
' .When-a select error occurs on the RWM:
Y:g Can the operator still move the Rod? (RWM is NOT bypassed and NO rod blocks existed prior to selecting the Rod.)
AnsweT YES or (0.5)
Yh If so, how far, and why? If not, why not? EXPLAIN YOUR ANSWER TUEL Consider both an attempted insert and withdraw actio (2.5)
. Answer if the following questions ~ concerning~ ADS logic, seal ins, and resets are True or False: The ADS drywell signal will always " seal in" the ADS logic any time the high drywell limit is exceede (0.5) If following an ADS actuation the control room operator presses both ADS timer resets the ADS valves will close and the . timer will rese .i (0.5)
. It on1 takes one ADS Division (Div. I or Div. II) to open i
the valve (0.5)
d .- If following an ADS actuation the control room operator manually stops all core spray and LPCI pumps the ADS valve will clos (0.5)
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The recirculation flow control system (RFCS) logic was designed e with the following flow limits: Limiter #1 limits speed to 30% Limiter #2 limits speed to 45% Limiter #3 limits speed to 80%.
For each of the following conditions state which limiter (if any) would control the RFCS circuitr (Respond with limiter number, speed, or no speed limiter actuation.) !
Condition #1: 100% power operation, feedwater control system
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discharge valve closes to the'90% positio (0.5)
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~ (0.5) Condition #3: 70% power operation, feedwater control system in auto and "A" reactor feed pump trip (0.5)
t Condition #4: 70% power operation, 'feedwater-control system
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in auto and the "A" motor generator scoop tube locks u (0.5)
3.9 ' Concerning the-load shedding and sequencing system: What are three (3) types of events that the sequencer will recognize and what is the difference in the sequencers response?
(2.25)
v- Following a load being resequenced to the proper bus can the operator override this action if the system is not needed?
Explai (0.75)
-3.10 Explain how the LPCI Loop Select Logic functions once an initiation signal is receive (Include in your explanation the effects if the Recirculaton Pumps are running and/or shutdow ,
Limit answer to selection of a loop only.) (2.5)
. .11 When is it allowable to fully withdraw the SRM detectors without causing a rod block? (1.0)
,
b
- Section 3 Continued Next Page -
c
- _ _ _ _ _ _ _ _ _ _ _ - ._
_ __- - __ - _ _ .
. .
.
-3.12 Concerning the Residual Heat Removal System: * Which RHR pump (s) can be controlled from the Remote Shutdown I
(0.5) State the function of the Pressure Switches located in the DTiEarge line on each RHR pum (0.5)
c. : State ~-the interlock associated with the minimum flow valves DWA(B)) . (1.0)
f
. .
--
End of Section 3 -
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G g d
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_ _ _ _
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . -
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4.0 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.1 List eight (8) immediate operator actions following a reactor scram. (Do not include verification or multiple action for same immediate action). (2.0)'
4.2 During normal operation the recirculation pump speed mismatch is administratively limited to: % of recirculation pump speed at greater than or equal to 70% of rated core flo . % of recirculation pump speed at less than 70% rated core flo Give three (3) reasons for the mismatch limitation (3.0)
4.3 Following a Failed Safety Relief Valve ( A0P-20.000.25):
, When must you manually scram the reactor? (1.0) According to'the procedure how would this be accomplished? (1.0)
4.4 What three (3b things are you suppose to do prior to entry into an area on a RWP? (1.5)
'
, 4.5 According to the Reactivity Control Procedure (E0P-29.000.08),
"
the reactor operator is authorized to start the Standby Liouid
,
l Control System: Under what condition (s) is this acceptable? (1.5)
p- Following the initiation of the SLC system all Lg water injection is stopped from entering the cor What is the reason for this procedural step? (0.5)
c. .Under what condition (s), (other thar4 directed to by S.S.),
would you manually override the SLC system once initiated? (0.5)
.
. .
A b
- Section 4 Continued Next Page -
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_- _ _ _
, _ _ - _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _
v ,
.
ll
' Fill in Table 4.6 with the appropriate respogse: (2.5)
,
MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE e Power Operation 1 6 l Startup 2 l
7 Hot Shutdown 3 8 Cold Shutdown 4 9
. Refueling 5 10
- With regard to the Control Rod Drive System:
' %' * When the opera [or' - discharoes the water side of the i accumulator, how does he know the accumulator is fully
,
drained? (1.0)
- Upon recharging the gas side of the accumulator why must the operator wait 30 minutes before certifying the gas l pressure?- (1,0)
u*
-
- List the entry conditions for the Primary Containment Control emergency _ procedure (EP-29.000.03). (Setpoints required.) (2.5) What four (4) items must be logged in the NSO log book when the reactor is declared critical? (2.0)
4.10' A caution in the Residual Heat Removal System (50P-2'3.205)
states "Do not secure or place an ECCS in Manual Mode unless verified ...": How many independent sources .are needed for verification? (0.5) What are the reason (s) for stopping an ECCS system? (1.0)
- What is the reason (s) behind the caution? (0.5). .
.
b
- Section 4 Continued Next Page -
F
_ _ _____ _______
F
.
_
4.11 In reference to GOP 22.000.03 "Startup. from Cold Shutdown to Rated Power," match the following startup :' activities with the pressures at which they should be performed, (ONLY ONE CORRECT ANSWER FOR EACH PRESSURE.) (2.5) Place HPCI in " standby" # Start a reactor feed pump e c., Place RCIC in " standby" #
. Place turbine sealing s steam In service
. Start SJAE's #
-
4.12 According to 22.000.03, (Startup from Cold Shutdown to Rated Power), why is use of the " Emergency In" mode of inserting control rods limited to NECESSARY RAPID POWER REDUCTIONS 7 (0.5)
!
- End of Section>4 -
'
- END OF EXAM -
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e
.y*
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.
FIGURE b 240-220 -
w 200 =
180 0 160 -
W X
-
140 -
120 - d 80 -
2 80 -
40 =
'
20 - *'
, 0-t-10 O to 20 30 40 50 60
.
TIME (hours)
TABLT0 el . 6 HODE SVITCll AVERAGE REACTOR CONDITI0ll POSITIO!! C00L Alli TEMP ER ATURE I I I I i POVER OPERATION I I I I I I I I . ;. I I I I STARTUP 1 I I I_
I l I _I I ! .1 1 HOT SHUTD0Vil i I
~.I I I I l 1 i ! I l COLD SHUTD0Vil i I I
't I I I I I I I
-
1 R E FU E L IIIG I I I I I I I
.
. ,o
.
______________________..........____..... ______..._________.....______ __ ' '
EQUATION SHEET
___.___...__.._____.......__.______.____.___.______._______.__......____ . .
Where ml = m2 (density)1(velocity)1(area)~1 = (density)
.....______________.__________________..2(velocity)2(area)2
_________ .....__.._____ ....____
KE.= mv2 PE = mgh PEl +KEl +P V
~l! 1 1 = PE +KE 2 +P 2 V22 where V = specific volume
.
P = Pressure
__....___________________________________________________________. _______
Q = dep (Tout-Tin) Q = UA (Tave-T Q = m(ht -h2 I
_____________________________________stm) __.._____________________________..__
P = Po lusur(t) p . p oe t/T SUR =_26.06 i
delta K = (Keff-1)/Keff CR1 (1-Keff1) = CR 2 II-Keff2)
M = (1-Keff1) SDM = (1.Keff) x 100% ,
.-
(1-Keff2) K
..
-_____......._________________eff
.
_____________________________________ . de.dy constant = In (2) = 0.693 A = A e-(decay constant)x(t)
c t1 t1
..________.___________/_2_________/_2 ___..._____ ._____________ .....__ .....
Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs
.
I ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr Density = 62.4 lbm/ft3 1 Mw = 3.41 x 106 Btu /hr Density = 1 gm/cm3 Heat of Vaporization = 970 Btu /lbm 1 inch = 2.54 centimeters Degrees F = (1.8) x (Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 Btu = 778 f t-lbf 1 Atm = 14.7 psia = 29.9 in Hg 9 = 32.174 f t-lbm/lbf-sec2
.___________ ..._______________________ __________________________......__
e O
_ M
5
t
_ _ _ _ __
i\:. .
MASTER COPY ANSWERS U. S. NUCLEAR REGULATORY COMMISSION ,
REACTOR OPERATOR LICENSE EXAMINATIDH Facility: FERMI-2
~
Reactor Type: BWR
'
Date Administered: 2-12-85 Examiner: GA SLY _ _ _ . . __
. .
Candidate: ANSWER KEY
.
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side onl Staple ouestion sheet on top of the answer sheet. Points for each question are indicated in parenthesis af ter the ouestion. The passing grade r,equires 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 Ca tegory % of _ Candidate's % of Value Total Score Cat. Value
. Category 25 25 1. Princip'les of Nuclear Power Plant Operation, Therwodynamics,
,
Heat Transfer and Fluid Flow 25 25 2. Plant Design including Safety and Emergency Systems 25 25 3. Instruments'and Controls 25 25 4. Procadores: Normal, Abnormal, Emergency, and Radiological Control 100 TOTAL Final Grade %
All work done on this examination is my own; I have neither given nor received ai Candidate's Signature
.
U. S. NUCLEAR REGULATORY COMMISSION
- -
REACTOR OPERATOR LICENSE EXAllINATION Facility: FERMI-2 Reactor Type: BWR Date Administered: 2-12-85 Examiner: GA SLY Candidate: ANSWER KEY
.
INSTRUCTIONS TO CAHDIDATE:
Use separate paper for the answer Write answers on one side onl Staple cuestion sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the ouestion. 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 start Category 1 of Candidate's % of . ;-
Value Total Score Cat. Value Category
'
25 25 1. Principles of Nuclear Power Plant Operation, Thermodynamics,
-
Heat Transfer and Fluid Flow 25 25 2. Plant Design Including Safety and Emergency Systems 25 25 3. Instruments and Controls 25 25 4. Procedures: Hermal, Abnormal, Emergency, and Radiological Control 100 TOTALS Final Grade %
All work done on this examination is my own; I have neither given nor received a f.d .
.
-
Candidate's Signature,
-
-
(
__ _ _ _ ,
-
.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERAT10N, THERMODYNAMICS, HEAT TRANSFER AND FLUlU FLOW 1.1 ANSWER Fast fission (0.25)
Resonance capture (239 Pu) ,_ __ (0.25) Fast fission has a positive reactivity effe.ct (0.5).
YvA Resonance capture: (overal1 negative due to less neutrons)
.
2350 has a negative reactivity effect 238U has a negative reactivity effect ..
239Pu positive if fissions, negative if produces 240Pu (0.5) ~" "~
Resonance capture .(0,5)
.
REFERENCE: 'NT/R248 (Reactor Theory) Chapter 3, pg. 9, 10 1.2 ANSWER (See Graph) Figure (3.0)
'
Grading criteria for each curv .5 curve shape (times, direction, etc.)
0.25 relative maximum 0.25 relative minimum REFERENCE: NT/R248 (Rx Theory) Chapter 10, pg. 1-10
. ANSWER Cooldown (0.25), limit not exceeded (0.75) (100*F. in one hour.) (1.0)
h... Cooldown,- (0391 thermal stress from cooldown results .in h additive tensile stress .at inner vessel walls (0,5). (1.0)
~~ ( L3) fe.. >
REFERENCE: NT/R248 (Rx Theory) - Chapter 9 pg. 11, 21,-17, 29 Fermi Thermo - Chapter 5 pg. 8 .
1.4 ANSWER
. (1.0) not change (0.5), K eff is independent of source-stren (1,0)
REFERENCE: NT/R251 (Rx. Theory) Chapter 11, Chapter 12 pg. 5-8 t - Section 1. Continued Next Page -
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1.5 ANSWER (0.5 eq. , 0.4 numbers, 0.1 math) Pf = po et /g 2000/50 = et/c t = 322 sec $22 (1,0)
. k0 . Assuming the reactor is initially supercritical , the doppler coefficient = adds negative reactivity before reaching the-heating range, producing a longer period and slower increase in the power rise.(0.5) - At the point of adding heat (fuel heat transfer is equal to the ambient losses) the moderator starts to heatup, adding more negative reactivity. Period gets longer and power rise declines even more.(0.5) When the reactor becomes critical (neg. reactivity coefficients equal pos. reactivity from fissions),. the IRM pen recorders should indicate a constant power, and the period meter indicates infinity.(0,5) ~
Power 'will start to decrease because of the increased effectiveness of the moderator coefficient, causing powe to stablize at a lower power level than in above.(0 5)
.
NOTE: (Bulk of answer for grader and clarification) (2.0)
. REFERENCE: NT/R251 ~(Rx Theory) Chapter 11 pg. 9 NT/R237 (Rx Theory) Chapter 13 pg. 1-3 1.6 ANSWER (>
Quality and boiling lengt (1.0)
REFERENCE: Fermi Thermo Chapter 9 pg. 5
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- Section 1 Continued Next Page -
_ _ _ _ _ _ _ _ __ . -__ ._
i: _
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- 1 7. . ANSWER-l
'
a. . increas (0.5) )
.b. increase (0.5)~
c. increase (0.5)
. decrease _( 0.5)
. REFERENCE: Fermi Thermo Chapter 7 pg. 9-11 Chapter'8 pg. 6, 7
'1.8 : ANSWE . , 3,.5,_4, 1 (0.25 for each order) -(1.25)
b~. - Annular, mist (0.5)-
- REFERENCE: Fermi Thermo Chapter 7 pg. 1-5
"I'.9. AN'SWER (0.5 equation,-0.4 numbers, 0.1 math)
s
~
- -a.- increase (0.25), greater head losses iJn piping (0.75) (1.0)
b.. decrease (0.25),- necessary - to compensate negative reactivity
.from doppler (0.75): (1.0)
-
-. REFERENCE: Fer1ni Exam Bank Questions 1.08 and 1.43 1.10 ANSWER f
.A ' A' jet pump converts T velocity heat .into a pressure head (0.25)~.
'
-Due to :the convergent section of ~ the driving flow _ nozzle,. the
' driving . flow ~ is accelerated. at a _ high velocity, _ this in turn creates L-allow- pressure .in the throat area (0.5). . Due to 'this
_ pressure differential,- - the driven flow - is accelerated and entrained 'with the ' driving flow steam. In the diffuser. section, 3 a - further reduction in velocity. is l achieved and the resultant discharge . pressure:.is developed (0.5). ~
(1.25)
- h Yes (0.25).' Even if the main condenser and associated. systems
'
'were-air l tight, gasses would collect in the main condenser in the
-
...- . form lof Hydrogen and 0xygen (from the radiolytic decomposition of water),'
.
- --
-claddingboundaryL (0.5) andthrough
> fission-
"useproduct related" gases (which escape - from cracks),(0.5.) '(1.25th )
~
REFERENCE: Fermi Exam Bank Q. 1.56
,
Fermi: Exam Bank Q. 1.14 -
)
- Section 'l Continued Next Page -
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.r--
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1.11' ANSWER %- ' (1.0 ) '
' REFERENCE: Fermi Exam Bank Question 1.38 (rewrite)
1.12 ANSWER False (0.25), fast only(0.75) (1.0)
REFERENCE: General.Rx Theory .
-1.13 ANSWER Increase (0.25), because more Rx - heat available .for turbine ~since less requirea 'for RWCU.(0.75) ~ (1.0)
(. REFERENCE: : Fermi lhermo Text Chapter 4 pg. 22 1.14 ANSWER (0.25 pts -for each response)
First-----doppler
!?cond----moderstor Third-----void ~ (0.75)
REFERENCE: Fermi Exam Bank Question 5.03 (rewrite)-
. l'
- End of Section One -
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A
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___..____________..__.__._________________ ________.___.....__________ .__
_
EQUATION SilEET
., ____ _____......__________.._____..______________________________ ________
. .
Where ml = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2
_______________________________________.._____________ .________________ .
KE.= mv2 PE = mgh PEl +KEl +P1V1 = PE +KE where V = specific Y
2 +P 2 V22 volume
. P = Pressure
.
Q = dep (Tout-Tin) Q = UA (Tave-Tstm) 0 = m(h3 -hp) -
__________________________________________________________________________
P = Po 10sur(t) p . p 9e t/T SUR = _26.06 i
______________________________.__________________. ________ ......._______
delta K = (Keft-1)/Keff CR i ll-Ke rri) = CR 2(1-K ergg)
M = (1-Keffi) SD!t = (1.Ke rg) x 100%.
.-
.(1-K eff2I Keff i decay constant = In (2) = 0.693 A = Ag e-(decay constant)x(t)
'
t
.
1/2 t1/2 Water Parameters 1?iscellaneous Conversions 1 gallon = 8.345 lbs- 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs
, .
1.f t3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr
- Density = 62.4 lbm/ft3 i g.tw = 3.41 x 106 Btu /hr Density = 1 gm/cm3 -1 inch = 2.54 centimeters Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8) x (Degrees C) + 32 lleat of Fusion = 144 Btu /lbm 1 Btu = 778 f t-lbf 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec2
______________________..___________.._____________________________________
J e m
=
- m
_ _ _ . . . . _ . . ._
.
.
FIGURE . .
300 280 260
240 220 !
Z 200 A s 10 Aer x/\ \ \-c<, 2:
160 140 120 \
100 \
$ B0 60 -
40 ONO 3I \
20 \ N O A-10 0 ' 10 20 30 40 50 60 TIME (hours)
.
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e e
G e
.-
E
. __ . _ . _ _ _ _ , _ - _ , . . . _ . _ _ _ . _
i e 1 i
5 l 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 ANSWER
' The ball check valv (0.5) Low nitrogen pressure (0.5)
liquid. leakage (0.5)
c. 3 gallons, causes an alarm in control room ~ (0.5)
f & gallons, causes rod withdrawal block (0.5)
100 inches, causes scram (0.5)
- .
Il L REFERENCE: C11 (CRDM) pg. 15, Rev. 3 PIS #C-11-50 (CRD) pg. 10, 12 2.2 ANSWER (0.25 for change, 0.75 for reason) Increase, due to reduced pressure on reference side of dP
. cel (1.0) Increase, due to reduced density in ference le (1.0)
'
)(g(; No change, al1 piping in drywell is horizontal, . reference leg is outside drywell . (1.0)
REFERENCE: PIS #B21 (RPV Instrumentation) Figure 2, pg. 4,16 2.3 ANSWER (0.5 pts for each response)
.~ . Separate the two-phase air-water mixture discharge from--
the water ring vacuum pump (0.5) Allow short and .long-lived isotopes to decay to acceptable ~ level (0.5) 'To remove the nongaseous decay daughters and A attentuate a transient pressure wave, thus providing to-protection for the vessels downstrea (0.5)
, Reduce the mositure content of the off-ga (0.5)
. ,2,3,1 (1.0)
- REFERENCE: N62, (Off-gas System) Pg. 14, 5, 12, 11; Rev. 3_
t - Section Two Continued Next Page -
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. , - , , . - . - - , - - - - .- -
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' 2.4 ANSWER a. '10-1 and 106 cps (0,5)
b. decrease (0,5) LPRM and at least 2 from each level (0.5) One'of four, one of eight, one of six (1.5)
REFERENCE: PIS #C51 (SRM) pg. 23, 7 PIS #C51-12, 13 (PRM) pg. 4, 19 PIS FC51-11 (IRM) pg. 13 2.5 ANSWER (0.5 'for each action and 0.5 for cause) - Outboard isolation valve (F004) closes (0.5), tripping the-RWCU pump (0.5) (1.0) Outboard (F004) and inboard (F001) isolation valve close, with either tripping the.RWCU pum (1.0)
'
1 REFERENCE: G-33-(RWCU) pg. 17, 16 .-
.
2.6 ANSWER (k1" . volts AC power (M/G set) norma (0,5)
480 volt AC via 480 VAC to 120 VAC transforme (0.5) energized (0.5) When either backup scram value is energized the air is vented from the lines. This opens the scram valves and closes the scram discharge volume vent and drain value (1.0)
~ REFERENCE: PIS-C71 (RPS) pg. 4, 5,-3 2.7 ANSWER By using - HPCI (0.25) or RCIC (0.25) systems and the SRVs in manual (0.5). (1.0)
- REFERENCE: Fermi Exam Bank Question 2.09 Section Two Continued Next Page -
_.
-
2.8 ANSWER
- Two (0.5)- . Continuity lamps extincuish Stanaby liquid control injection valve ignition continuity loss alarm is annunciate . Pump discharge pressure equal to-Rx pressure + 40 ps . Pump run light ON 5. Storage tank level decreasing 6. .Rx power level decreasing 7. Testable check valve indicating OPEN 8. RWCU isolated (Any s'ix 0 0.33 pt. each) (2.0)
REFERENCE: C41 (Standby Liouid Control) pg.11,12, Rev. 3 2.9 ANSWER
, Prevent water hammer by removing condensate. - (0.5)
KS c' . CST - less than 10,000 gallon (0.5)
Suppression pool - greater than 5" above normal (0.5) Suppression pool inboard isolation valve (F042) opens, Suppression pool outboard isolation valve (F041) opens, and CST isolation valve (F004) closes following F042 &
F041 full ope (e.g., Suppression pool valves open proir to closure of CST valve) (0,5)
'
. REFERENCE: .PIS E41 (HPC1) pg. 8, 14, 19, Rev. 3 2.10 ANSWER Joystick, bypasses one channel if INOP or for maintenanc (0.5) <30*. power, RBM not required because RWM and RSCS " enforce" flux distribution (0.5) Selected . edge ' rod , can never produce a condition that .
. approaches a thermal limi (0.5)
REFERENCE: Fermi Exam Bank, Question 2.40
-
CC51-14 (RBM) pg. 12, Rev. 3 -
u
- Section Two Continued Next Page -
- . _ - .
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2.11 ANSWER a. 2 (1.0)
b. Assures NPS (0,5)
REFERENCE: PIS-P44 (EECW) pg. 1 End of Section Two -
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v s
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+
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- INSTRUMENTS AND CONTROLS e
3.1' . ANSWER
'
- The drift alarm 'will activate when the rod passes' an - odd '
numbered read switch (0.5), or dual indication of two
~ different locations (0.5) (1.0) No (0.5)
REFERENCE: C11A-00 (RMCS) pg. 7, 4 j O' 3.2 ANSWER , decrease (upstream flow element) (0.5) increase (0.5) remain the same (downsteam flow element) (0.5)
d.- decrease (0.5)
REFERENCE: PIS N21 (Reactor Vessel Level Control System)
'
pg. 12, 13, Figure ;,
. 3.3 ANSWER -
t .
' Low Vacuum Alarm 26.3" Hg. decreasing /'r ' Low Vacuum Jurbine Trip - _ 25.5" Hg. decreasing /.? i Low Vacuum Bypass Valve Trip 16" Hg. decreasing c 7 , ._ Low Vacuum Group 1 Isol 14" Hg. decreasing rg RFP Seal Water Return Pump Auto Start 7 psia
' RFP Seal Water Return Pump Auto Stop 6 psia RFPT High Exhaust Pressure Trip 0 psig
(5 0 0.4 ea) -- ' ,
-
, r , (2.0)
REFERENCE: -INEL Exam Bank Question 5445 PIS #N61 (Main condenser & aux.) pg. 5, 6 N21-1 (feedwater) pg. 10,11,17 Procedure 20.125.01 pg. 2, rev. 3 210/R181/5.15 X 34 ANSWER
-
GL' . SRV failure 7% flow. reactor pressure falls
(0.5) . CV start to respond to pressure decrease (voids)- -
-(0.5) CY close to 78% demand and presure returns -to normal (0,5)
- .
REFERENCE: PIS #N30-12A (Gov. Pressure Control) Figure 1
'
,
- Section Three Contiunued Next Page -
. . _
. _ _. _
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c 10 L- '
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3.5 ANSWER- "
/ k (alarm, thermal, neutron, downscale(, Irpor 1
' "'
(0,5) '
(Cf# R
Neutron (LPRM) -
(0.5)
g'. APRM thermal power, APRM neutron power, APRM.downscale, INOP (1.0) (joystick) (0.5)
REFERENCE: C51-13 (PRM) pg. 13, 19, 23, 15 3.6 ANSWER Yes .(0.5)
.j( hi It can be moved one notch out (0.5), before a withdraw block will block furu.*a- movement.tu.5). If the rod was inserted it will move as far as the operator wants (0.5), as long as it is not the third . insert e rror. ( 0.5 ) If it were the third
-
inser error, it would only go Une notch.(0.5) (2.5)
REFERENCE: INEL Exam Bank Parent Question 5443,
'C11-08 (RWM) pg. 20, 21 .-
.
3.7 ANSWER true (K38 seal in) (0.5) true (K9B TDC opens / resets) (0.5) true (KGA,B & K7A,B close) (0.5) true (KGA,B & K7A,8 - Trip) (0.5)
REFERENCE: (821-04) (ADS) pg. 6, 7, Figure Logic diagram (03-15-42-01-A2)
3.8~ ANSWER Limiter el (30%) --
(0.5)
- g jec- . .
^' ~ '#
b._ l'-4+a- 73 (6GL) - ' -
(0.5) No limiter (0.5) No limiter (0.5)
,
REFERENCE: _B-31 (FWCS) pg. 10, 13, Rev. 3
. A b
- Section Three Contiunued Next Page -
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. . . . . . _ - - . - . _- __
3.9. ANSWER -
jg#. . Loss of off-site power (LOP); ECCS~ premissives, no
' star (0.75) LOP followed by LOCA; LOCA signals- override LOP load squencing, ECCS start automaticall (0.75) LOCA followed by LOP; CS permissive, auto start on LO (0.75)
h ;I; k.~-inter h is a cen
'1C) iiii.iuie inier kc' j,;j6-t- f !. 2c,~
- prechde Op;r:t0r
. % A,, p Ac, (0.75)
-
M: h J-j (0.z r}
'(0.25 No, 0.5 for reason) .
't
- Note: Yes also acceptable if reason corresponds properly REFERENCE: R30 (Emerg. Diesel Gen. and Aux. System) pg. 46, 47 3.10 ANSWER-
,
. If both recire. pumps are running the- logic isAelayed for 2
- -
second_s_ prior to selecthg-the loop for injectio (0.5)
. -
If either o[ both. pumps are NOT running then the logic is
-
delayed. for 0.5- seconds (0.25), after which a' trip signal .is
'given to both pumps.(0.25) (0.5)
. 'The logic then requires. that reactor pressure be below 5 !
psig (0.25), before continuing _to the 2 second time delay prior' to loop selection.(0.25) (0.5)
!
. Loop selection is made by . comparing the two riser
'
pressures.(0.5) If . loop ' A' riser -pressure is' greater than loop 'B' riser pressure by a-preset differential -then loop 'A'
~
is selected.(0.25) If loop ' A' pressure IS NOT greater than
- loop . 'B' ' pressure by the preset differential, then loop '8' is
~
selected for. injection.(0.25) (1.0)
REFERENCE: INEL' Exam Bank Question 5446 E11 (RHR) pg. 29, 30
~3.11 ANSWER = gg . As-long as the counts stay above 100 cps' .
(0.5)
2. All the IRM'.s are on range eight or above a
.(0.5)
!. ri r u w a. , a e ,
n-REFERENCE: Fermi Exam Bank Question 3.07 (rewrite)
t
- Section Three Contiunued Next Page -
--
I
,..
..3.12 ANSWER RHR pump A- (0.5) Provides signal of pump running to ADS logic (0.5)
- c. . F007A(B) will auto open if loop flow is <2200 gpm for.1 seconds if. either pump .is running in that loop as indicated
.
. by; closure of pump breaker.(0.5) Will auto close on ' loop flow is ~> 2200 gpm.(0.5) (1.0)
REFERENCE: Fermi Exam Bank Question 3.21
.
- Er.u of Section Three -
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4.0 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.1 ANSWER
- Verify automatic actions Depress both manual scram pushbuttons Verify reactor power is decreasing Place mode switch in refuel and verify all control rods are in Place Rx mode switch to shutdown Maintain water level between level 3 and level 8 Maintain reactor pressure between 900 psig and 1075 psig Insert SRM and.IRM detectors and range the IRMs Trip the main turbine 10. Verify recirculation pumps have run back to min. speed *
11. Run program 0D7, Option ~2
-
12. Notify S.S. and may require classifying in the EPIP EP-101
'
-(Any 8 for full credit) (2.0)
REFERENCE: AOP 20.000.21 pg. 1, 2
4.2 ANSWER .' Avoid possibility of " fooling" LPCI loop selection logic (1.0) Back . flow through idle or reduced flow jet pumps can cause jet pump vibration and therefore unnecessary stress on jet pump component (1.0) Provides for some coastdown flow from the unbroken loop during a long break acciden (1.0)
L r. . .
REFERENCE: 831 (RFC) pg. 9, 10 Rev. 3 T.S. 3.4. .3 -ANSWER . If the SRY cannot be closed within two (2) minutes (0.5) If torus temperature exceeds 95* (0.5) . Decrease recirc. flow to minimum (0.5)
- Place mode switch to shutdown (0.5)
REFERENCE: A0P - 20.000.25, Pg. 1 .
.
-
-
- Section Four Continued Next Page -
. . , , _
-
.
4.4 ANSWER .
.
1. - Read the RWP (0.25), and resolve question (0.25) (0.5)-
2. Sign (0.25), "RWP. Acknowledgement Log Sheet" (0.25) (0.5)
3. Review copy of RWP (0.25), at work area (0.25) (0.5)
REFERENCE: .T,5. 4,7.1.2. pg. 379 7-S AP 12.000.13 pg. 10
. ANSWER ~ . Reactor power cannot be determine (0.5)
. Reactor power > 6% (0.5) and torus water temperature reaches 110'F (6!5). (1.0)~
err') ;'Glt()
ggjpor To concentrate the boron liqui (d. oveo e d ',* <8~ (0.5)- When the SLC tank level indicates zero (0.5)
,.
REFERENCE: E0P'29.000.08, pg. 1-4
.,
1- 4.6 ANSWER -
C
'
MODE SWITCH AVERAGE REACTOR CONDITION - POSITION COOLANT TEMPERATURE 1. F0WER OPERATION Run Any Temperature (0.5)
2. STARTUP Startup/ Hot Standby Any Temperature (0.5)
3. HOT SHUTDOWN Shutdown > 200*F (0.5)
4. COLD SHUTDOWN Shutdown < 200*F (0.5)
5. REFUELING Shutdown or Refuel ][ 140*F (0.5)
(0.25 for mooe position and 0.25 for temperature)
REFERENCE: Technical Specification 4.7l ANSWER 1( h* ' When the nitrogen purging gas pressure remains constant. t r (1.0)
t .a. a d v . W- wn
- Gas expansion lower the temperature and you nust wait for it to come to ambient equilbriu (1.0)
.. .
REFERENCE: S0P 23.106 pg. 9, 10 *
-
-
-
Section Four Continued Next Page -
. . . - _ _ _ _ _ _ - _ _ _
.
. :~
'
4.8 ANSWE ,, Torus water level > +2" (124,220 ft3)
. torus water level < -2" (121,080 ft3)
.
f h,- 3._ - Torus water average temperature > 1-39Y-d^~, Drywell atomsphere average temperature > 135*F Drywell pressure > h88 psig
.Q, /A3 gas I.E3 wO -
(0.25 each condition, 0.25 each setpoint) (2.5)
-REFERENCE: E0P 29.000.03, pg. 1
- ANSWER * Time (0.5) Critical rod position information, rod secuence, rod group, rod and rod positio (0.5) Reactor coolant temperature (0.5)
'
4. . Reactor perico .i (0.5)
Jga REFERENCE: GOP 2'2.000.03, pg. 6 (Startup from Cold Shutdown)
4.10 ANSWER - (0.5) . Misoperation in autiomatic is confirmed (0.5) Adequate core cooling is assured.(0.5) (1.0) Will not auto initiate (0.5)
REFERENCE: 50P 23.205, pg. 14-4.11 ANSWER (0.5) (0.5)
.- (0,5)
- (0.5) (0.5) '
REFERENCE: FERMI Exam Bank Question 4.08
- .
- Section Four Continued Next Page -
,
- , - -. .-, ,--,-- ., 1m- -,
. .
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4.12 ANSWER High hydraulic. force (0.5)
REFERENCE: FERMI Exam Bank Question 4.12
- End of Section Four - ,
- END OF EXAM -
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U. S. NUCLEAR REGULATORY. COMMISSION SENIOR REACTOR.0PERATOR LICENSE EXAMINATION FACILITY: FERMI 2
_________________________
' REACTOR TYPE: BWR-GE4
_________________________
DATE ADMINISTERED: 85/02/12 l
_________________________
EXAMINER: LANG,T.^
___.________________L____
APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
__________________________
Uno separate paper for the ,swer Write answers on one side onl Staple question sheet on top of the answer sheet 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 tho examination start .
-
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL' SCORE VALUE CATEGORY
________ ______ ___________ ________ ___________________________________
25 00 24 THEORY OF NUCLEAR POWER PLANT
________ ___'88__ ___________ ________ OPERATION, FLUIDS, AND-
,
THERMODYNAMICS 25.50 25.37 PLANT SYSTEMS DESIGN, CONTROL,
________ ______ ___________ ________ AND INSTRUMENTAlION 25.00 24*88 PROCEDURES - NORMAL, ABNORMAL,
________ ___ __ ___________ ________ EMERGENCY AND RADIOLOGICAL CONTROL
,
25.00 24.88 ADMINISTRATIVE PROCEDURES,
________ ______ ___________ _-__---
CONDITIONS, AND LIMITATIONS
'2100.50 Il00.00 ___________ ________
TOTALS
________ ______
g FINAL GRADE _________________% [
All work-done on this examination is my own. I have neither -
_
given not received ai I APPL 5CIUT 5'5 GUITUR5-~~~~~~~~~~~~~
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4 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2
-------------------------------------- y'
__--__-------- ;
l-GUESTION 5.01 (1.50) $
For the following conditions, state if the FUEL TEMPERATURE COEFFICIENT -
becomes more nesitive or less nesitiv '.
A. INCREASE in fuel temperatur (0.5) :
l
,
B. INCREASE in moderator temperatur (0.5)
C. INCREASE in void fractio . (0.5) --
QUESTION 5.02 (1.00) h Define the following terms * 1 A. Latent heat of vaporizatio (0.5)
B. Steam Dualit (0.5) 3_
i =
QUESTION 5.03 (3.00)
Explain the effects of Increasing the followins core parameters on Z steady state critical powe A. Core Flo (1.0) e B. Inlet Subcoolin (1.0)
C. Reactor Pressur (1.0)
_.-
QU.ESTION 5.04 (2.50)
1r equalibrium xenon is obtained (the reactor has been. operated at {
constant power for many hours), and the reactor power is doubled, will the new equalibrium xenon consentration be twice as great? Explain your }
answe ;
00ESTION 5.05 (3.00) f; x
-
A.After making a rod notch with the reactor critical, you notice a 100 second period. How much reactivity was added by the rod notch? (assume i BOL). (1.0) ;
B.After a reactor scram from power the shortest stable period possible c (1.0) i is -80 seconds. Explain this statemen >
l C.Is the initial period immediately followins the scram shorter than 4-80' seconds? Explain your answe (1.0) ,
-!
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_ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
--------------------------------------
GUESTION 5.06 (2 00)
A common misconception regarding rod worth is that, 'if the neutron flux increases in the vicinity of a rod, the rod worth (differential)
of that rod increases.' Explain why this statement is incomplete in oxplaining rod wort ,
QUESTION 5.07 (2.00)
A.What are four indications inside or outside the control room, which could be used to determine if a pump is cavatating? (1.0)
B.What is cavitation? (1.0)
,
QUESTION 5.08 (1.50)
SBLC is designed to reduce reactor power to zero level, while allowing the neuclear system to cool to room temperature with the control rods ,
'
romaining withdrawn. To do this several positive reactivity effects must bo overocome. What are six of these effects?
DUESTION 5.09 (3.00)
.You increase-core power by pulling. control rods around the center fuel
- bundle. Assuming that recirculation speed is kept constant, would the flow through the center bundle increase, decrease, or remain about the ocee? Explain your answe '
00ESTION 5.10 (2.50)
The Reactor is on a 100 second period. Moderator temperature is 160 des F. With no operator action:
,
a.What is the associated reactivity at this period? (1.5)
l b.What will be the moderator temperature when the reactor is again on an infinite period? (1.0)
Assume BOL'for time in Reactor Lif State.any assumptions you mak SHOW ALL WOR .
1 -
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.
THEORY OF. NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 5o
____
g________..____________________________
______________
QUESTION 5 11 (3.00)
The reactor is shutdown by 5% dk/k and the SRM's indicate
'10 cp If K eff of the reactor is' increased to 0.981 WHAT should be-the approximate count rate? (Show all work.) (1.5)
b. What is the original K eff?- (1.5)
,
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. PLANT-SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 5
______________________________________________________
SUESTION 6.01 (3.00)
What parameters'will be indicated on the Rod Block Monitor meter
.uith the meter switch in each of the following positions: Input Count Reference * Block
. Flow f.- Average (3.0)
DUESTION 6 02 (2 00)
,Briefly describe the purpose of the following UPS switches:
A. Static Bypass Switc (1.0)
B. Static Trznsfer~ Switc (1.0)
.
QUESTION 6.03 (3.03)
With regard to the core spray system:
A. State the automatic initiation signals of the syste (1.0)
B. In the event that a Core Spray pump CMC switch is in DFF when an initiation signal is received, will placing the CMC switch in AUTO cause the pump to start? EXPLAI (1.0)
-(.
C. Which valve (s) receive a CLOSE signal from the Core Spray logie upon receipt of an initiation signal? (1.0)
D'JESTION 6.04 (2.00).
Explain how the Logic of the Reactor Protection System is altered by installing the shorting links in the Neutron Monitoring System i.e.,
links installed verses links remove .
-
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 , ______________________________________________________
GUESTION 6.05 (2.50)
If1during.a unit start.up you receive a spuricus Group It A. What valves would you expect to close? (1~.0)
B. What'are six_ signals you would check to determine the cause of the Isolation? (1.5)
, GUESTION .6.06 (3.50)-
Which Reactor Feedpump trips provide for protection from:
'
A. Potential cavitation damage? (Three required for full credit). (1.5)
'
B.; Potential' turbine blading damage? (Four required for full credit). (2 0)
- QUESTIO .07- (3.00)
.
Iar each of~the RCIC (Reactor Core Isolation Cooling) System-component
- the reactor vessei.IF IT WILL NOT INJECT, STATE WHY, AND IF IT WILL INJECT,' provide CNE. POTENTIAL ADVERSE EFFECT OR CONSEQUENCE OF syste _cperation with_tne failed component. Assume NO OPERATOR ACTION, and the
.
. component is in the failed condition at the-time RCIC receives the AUTO Linitiation. signal. Consider.each item seperatel A. The Barometric Condenser VACUUM PUMP fail (1.0)
B. The MINIMUM FLOW VALVE fails to AUTO open (STAYS SHUT) when system conditions require it to be ope _
(1.0)
C. The RCIC PUMP DISCHARGE' FLOW ELEMENT OUTPUT SIGNAL (to the RCIC flow controller) is f ailed to its- MAXIMUM outpu (1.0)
r ,
GUESTION 6.08 (1.00)
Staie four(4) EMERGENCY DIESEL GENERATOR trips that WILL NOT shutdown
_the engine with an AUTO start condition present.
>
QUESTIDN 6.09 (3.00)
What are the conditions which will cause Recirculation Pump Run Back
~
.cignals to be initiated. Be specific. Match speed limit to. condition-cnd state which runbacks must be reset manually. (Three required fo .
. full credit). _
i-
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PLANT. SYSTEMS. DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 .-----------------------------------------------------
QUESTION 6 10 (2.50)
If you receive the followins alarm:
XXXXXXXXXXXXXXXXXXXXXXXXX X APRM FLOW UNIT X X INOP X X X
'
XXXXXXXXXXXXXXXXXXXXXXXXX What AUTOMATIC action will result from the alarm? (0.5) B. Other than the INOP, what are two (2) signals associated with the flow units which will cause the SAME automatic action to occur as in(1.0)
part '
a' above?
C. What are two (2) conditions which will cause an INOP trip of a flow (1.0)
unit?
.
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70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE- 8
~~~~ ~ ------------------------
RA6E6E6656AL C6 TR6t
____________________
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QUESTION 7.01 (3.00)
What are the entry _ conditions for E0P 29.000.08 ' Reactivity Control'?
OdESTION 7.02 (1.00)
'
FA? cording to procedure 20.129 (Loss of Station and/or Control Air Pressure) when are you required to manually scram the reactor?
"
QUESTION 7.03 (2.50)
List five (5) indications / symptoms ~of a jet pump malfunctio #
.0UESTION - 7.04 (2.00)
If the Standby Liquid Control System is initiated, what are five (5)
,-
indications or parameters that should be checked or verified to insure
.the system is' operating properly? (0.4 each)
j QUESTION 7.05 (1.50)
If plant. conditions are such that the control room must be evacuated,
'what are three (3) steps which should:be performed prior to evacuating?
OhcSTION~ 7.06 (2.00)
What must be done if MORE THAN ONE withdrawn control rod has an INOPERABLE. SCRAM ACCUMULATOR while operating the plant at power?
GUESTION 7.07 (2.00)
'With regared to EOP 29.000.08, . REACTIVITY CONTROL
~
If during a transient the Reactor fails to scram by a VALID _ scram
,
signal,-WHAT condition (s) require the initiation of the Standby liquid Control System and the isolation of the Reactor Water Cleanup System?
.
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I ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 9 ,
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~~~~RI656E655EIE~E6UTR6L
____________________
,
l QUESTION 7.08 (1.50). !
Under'what conditions-is an operator allowed to " DEFEAT THE AUTOMATIC l ACTION OF AN~ECCS*?
QUESTION 7.09 (2.5';) ,
What are the entry conditions for EOP 29.000.03 ' PRIMARY CONTAINMENT CONTROL'?
QUESTION 7.10 (3.00)
A reactor water isotopic analysis for iodine and specific activity is roquired when any of.three-(3) conditions occur. What are the three (3)
' ecr.d i t ons?
QUESTION 7.11 (1.40)
How is control rod coupling verified?
QUESTION 7.12 (1.00)
Roactor level indicatrots can-be expected to___________.once drywell tcmperature reaches the. reactor vesseld saturation temperatur (Choose one) Read lower than actual vessel leve B. Read' higher than actual vessel leve C. fail lo D. fail to respond to level change .
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f 1 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10
--- EA5i5E55iEsc
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E6siE5E------------------------ l
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QUESTION 7.13 (2.00)
Rosarding the RHR SDP 23.205 when operating in the SHUTDOWN COOLING MODE:
A. You are cautioned to NOT use Outboard Injection Valve E11-F017A/B Jto control cooldown rate. What adverse effects could result, '
'
according to this procedure? (1.0)
B. This procedure also cautions you to exercise care during one portion of a valve realignment to avoid RAPID DRAINING' of the REACTOR VESSEL. Describe THREE FLOW PATHS associated.with the RHR-system that could result in RAPID DRAINING of the REACTOR VESSE (1.0)
.
?
.
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ADMINISTRATIVE PROCEDURES, CONDITkONS, AND LIMITATIONS' PAGE 11 8 .-
QUESTION 8.01 (2.00) gjf, g/ 9c:xY pt y ,: .. f, g
.
ucec g equal to or greater than ,3Ef RATED S#i'
A. If the reactor is operating a^.
THERMAL POWER and under a LIMITING CONTROL ROD PATTERN, what action
.would you take if a RBM were to fail? Be specifi (1.0)
(1.0)
B. What is a LIMITING CONTROL ROD PATTERN?
'
DUESTION 8.02 (2.00)
' If a temporary procedure change is required on the backshift, who~
must sign the ' Temporary Procedure Change Request *? (1.0)
B. What limitations are placed on making temporary changes?
.
(1.0)
CUESTION .8.03 (2.50)
lA. On shift operations personnel and security are granted UNLIMITED ACCESS to the control room. What are four (4) others (by title) who are also granted UNLIhITED ACCESS? (2.0)
B. Who has authority to EXCLUDE non-essential personnel from the control room when their presence is hampering the operation of the (0.5)
control room?
8.04 (2.00)
}"ESTION What are the Technical Specification limits for Reactor Coolant System loakage?
C:lESTION 8.05 (2.00)
Dafine the followins ter*S* (1.0) Core Alteratio (1.0) Shutdown Margin.
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8. - ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 12- l
l GUESTION 8.06 (1.00)
Rogarding the TECHNICAL SPECIFICATION curves for MINIMUM TEMPERATURE FOR CRITICAL operation:
A. What could happen if critical operation was conducted LESS THAN the minimum temperature specified? (1.0)
B. Is this temperatture expected to change over core life and if so, why? (1.0)
GUESTION B.07 (3.00)
Briefly explain WHY each of the following-RECIRCULATION PUMP STARTING-LIMITATIONS are necessary:
A.-The pump in an-idle recirculation loop shall not be started ,
-
unless the temperature of the coolant within the idle and operating loop are within 50 degrees F of each othe (1.0)
. B. When in one pump nperation, the idle pump shall not be started unless the active pump speed is reduced to less than 50% rated spee (1.0)
.
C. Recirculation loop flow mismatch shall be maintained within 10%
of rated recirculation flow with core flow less than 70% of rated(1.0)
core flo GUESTION 8.08 (2.00)
As per Technical Specification definition, what is the difference 1 batween Hot Shutdown and Hot Standby?
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13 8.
__________________________________________________________
l QUESTION 8.09 (2 00)
The statements below define each of the four (4) emergency SITE AREA EMERGENCY, classifications. Match the classification (i.e.,
UNUSUAL EVENT, GENERAL EMERGENCY, OR ALERT) aplicable to each statemen A. Does not require activation of the emergency centers (TSC, or (0.5)
B. Major failures of plant functions needed for protection of the public, but the release is not expected to exceed the upper (0.5)
PAG's (5 rem whole body, 25 rem thyroid.)
C. Limited release of radioactivity in excess of Tech. Specs. limit is probable. Does not require protective actions be recommended (0.5)
to the state and county authoritie D. A release of airborne activity which results in off-site
'
exposures in excess of the limits specified in the USEPA (0.5)
Protective Action Guide .
QUESTION 8 10 (2.00)
Ragarding Limiting Conditions for Operation, 3.6.2.1, for Suppression Chamber The maximum average temperature during Operational Conditior.s 1 or 2 is ____ dest ees F , except that the maximum average temperature may be permitted to increase to____ degrees F during testing which adds heat to the suppression chamber or to ____ de3rees F with thermal
.
power less.than or equal to 1% of RATED THERMAL POWER or to ____
degrees F with-the MSIV's closed following a scra QUESTION B.11 (1.50)
The reactor at Fermi 2 has several power distribution limits. One of these is Linear Heat Generation Rat (0.5)
A. What is the limit for LHGR?
B. When does this limit apply? (0.5)
.
(0.5)
C..What must be done if this limit is exceeded?
.
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_ _ - - - _ . _ - . - . _ _ _ _ _ _ . _
- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14
__________________________________________________________
QUESTION 8.12 (1.50)
Answer the following questions as they pertain to Containment Systems Toch. Specs.:
A. How many Suppression Chamber-- Drywell Vacuum Breakers are required to be operable in Operational Condition 3? (0.5)
B. How many Reactor Building -- Suppression Chamber Vacuum' Breakers are required to.be operable in Operational Condition 3? (0.5)
C. When must the Drywell atmosphere be less than 4% 0xygen'by volume? (0.5)
QUESTION 8.13 (1.50)
.Mosarding PLANT STARTUPS:
A. Who must REVIEW the REACTOR STARTUP MASTER CHECKOUT 22.000.01 upon completion? (0.5)
B. In the REACTOR STA'RTUP MASTER CHECKOUT, who determines which
. systems are required to be lined up? (0.5)
C. Followins a reactor SCRAM, who has responsibility for authorizing a return to power? (0.5)
.
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- . . THEORY OF NUCLEAR POWER PLANT 0PERATION, FLUIDS, AND ', PAGE 15
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ANSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 5.01 (1.50)
A.less B.more
- C.more .5 each .
. REFER'r:N CE Formi-Question Bank (5.02)
ANSWER 5.02 (1 00)
A. Energy necessary to convert a-unit of a substance from a saturated
^
' liquid to a saturated vapo (0.5)
B.The ratio of the mass of steam to the mass of the total flui M(steam)
X=---------------------------
M(steam)+M(liquid) (0.5)
REFERENCE Formi Question Bank (5.08)~
ANSWER 5 03 (3.00)
< A.Increasiris core flow will cause critical power to increase due to the increase heat removal capabilit (1.0)
D' Increasing subcooling causes an increase in critical power.-The inlet
.
enthalpy will be reduced and the heat which can be removed will increas (1.0)
C. Increasing Reactor Pressure reduces the energy to be at critical ]
quality therefore critical power will be lowe (1.0) ,
,
REFERENCE '
Fermi Question Bank (5.09) l
..
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-. - -. . . _ . . . _ . . - - _ _ . _ -. __ _
. _ _ _ _ _ - -
..
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. THEORY DE NilCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE' 16
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_-
ANSWERS -- FERMI 2 -85/02/12-LANG, ' ANSWER 5 04 (2.50)
No. The production rate is directly proportional to power level, but removal rate is preportional to xenon concentration and it contains a ppwer dependent ter m, thermal neutron flux. Since flux is directly-proportional to power level the burnout term becomes more significan This results in an equalibrium xenon value which is higher than the original value, but not twice as hig REFERENCE Formi Question Bank (5.14)
A'NSWER 5.05 '(3.00) T=B p/xp.so p =B/xT+1- x= Lambda assume B=.0072 (BOL)
p=.0072/(100)(0.1)+1=6.545 x 10E-4 delta k/k (1.0)
B.After'the initial prompt drop, power can not decrease faster than the longest lived delayed neuton appear (1.0)
C.Yes The initial drop in power will only be due to the prompt neutron (1.0)
.
REFERENCE G3neral Theory ANSWER 5.06 (2.00)
The worth depends on the neutron flux in its location compared to the overage neutron flux in the core. Thus, as power is increased, the flux 01so increases. This will increase the worth of the rod if the local flux is increased more than the core average flu REFERENCE G3nerdl Theory
-
,
.
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k
b THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17
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CNSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 5.07 (2.00)
A. Vibration, noise, lower or fluctuating amperage / flow /ditcharge pressur (1.0)
B.The formation of bubbles of steam in the suction of a pump due to operation with insuffiecent pressure. The bubbles collapse in the volute or high pressure regio (1.0)
REFERENCE General Theroy ANSWER 5.08 (1.50)
Any six for full credi A. Complete decay of rate,d power xenon inventor .B.The elemination of steam void C.The changins water density from hot to col re. Reduced doppler effec i.The reduced neutron leakage from boiling to col F. Decreased control rod worth as moderator cool G.Suffiecent reactivity added to insure shutdown 7 *aC margin met. [~MMM Futt Y m ;WMO '/ -
pas;v v6 rum rivt ' ~ O f Lc-REFERENCE Goneral Theory
' ANSWER 5.09 (3.00)
Once the control rods are pulled steam forms in the coolant channel, it forces a volume increase in the coolant, the more steam the greater.the volume. The greater the volume, the greater the frictional pressure drop
. cost be in the channel. But the pressure drop across the core can not be changed by any one bundle. The pressure drop across-a bundle is locked botween the pressure in the lower plenum and the pressure inside the chroud. This means that as power and consequently steam production are increased in the bundle, the pressure drop can not change. If the pressure drop can not change then flow must decrease. Core orificing was installed to minimize the effects of two phase flow.HAs the pressure drop due to two phase flow increases bundle flow decreases. However, as bundle flow d: creases the pressure drop across the orifice decreases. The total effect io that bundle flow will remain about the sam _-
NOTE Shorter e>:plainations would be excepte .
- _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _
! THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18
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ANSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 5.10 (2.50)
Assumptions: Lambda = 0.1 see -1 (given)
Beta Bar = 0.0070 (0.25) *
Alpha T = -1 X 10 E-4 (0.25)
Period = 100 see (given) (0.5)
T= (B-p)/(.p): solve for p p = B/1+T.:
p = 0.0070 / 1 + 100 X 0.1: p = 6.3 X 10 E-4 delta K/k (1 0)
NOTE: The second part will be graded independent 1v of the first par dalta T mod = p / alpha T
= 6.3 X 10 E-4 delta K/K / 1 X 10 E-4 delta K/K des F
= 6.3 des F ,
Moderator Temp = 160 des F + 6.3 des F = 166.3 des F (1.0)
REFERENCE Roactor Theory ANSWER 5.11 (3.00)
formulas!
p =-(Keff - 1)/Koff CR1(1-Keff1) = CR2(1-Keff2)
.05 = (Keff-1)/Keff 10(1 .952) e CR2(1 .98)
.05 = 1 - 1/Keff 10(1 .952)/1 .98 = CR2-1.05 = -1/Keff .48/.02 = CR2-1.05/Keff = -1 24 = CR2 (1 5)
Koff = -1/-1 05 Keff = .952 (1.5)
(The two calculations will be graded seperately)
REFERENCE Reactor Theory
_
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_ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19
. ANSWERS -- FERMI 2 -85/02/12-LANG,T.
ANSWER 6.01 (3.00)
o. Input - Any of the LPRM inputs Count: -The number of LPRM inputs which are operable Reference: The reference APRM input * Block - The trip level reference
.o. Flow: The flow input to the slope and bias circuit
' . Average: .The RBM channel output (0.5.each) (3.0)
REFERENCE RBM Lesson Plan f
ANSWER' 6.02 (2.00)
A. Transfers load to an alternate AC supply should the inverter voltage dr op. below a ' pr e-set ' level . Mv+4-bewanuall" ee et4eet-oce-nomi power-suppl Q m. >> o eg. y FL- DC /IW' N ,mr New A 's " UL B. Automatically transfers the load from the inverter to the voltage r e g u l a t o r i-f- e n y --e f--the--f ol-low i ng-e x i st4 w/:. t i nen r 1.Undervoltage a) slow @bWN CU Am M W@ 3 W
WHM co w C> t T i p o { -b) fast 2. Abnormal Frequency 3 0verload
'EFERENCE Formi Question Bank (6.36)
ANSWER 6.03 (3 00)
A. Rx low low low level-level 1 3 UwF CC uA e (v: .. J (1.0)
B.~No. The CMC will have to be placed in RUN to cause the CS pump to sta'et. The ' Manual Override Seal-In' indicator above the pump CMC will be illuminate (1.0)
C.-Full flow test valves, F015A/ . (1.0)
e REFERENCE u-Formi Questin Bank (6.37)
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~6 . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20
..-----------------------------------------------------
. ANSWERS ---FERMI 2 -85/02/12-LANG, ,
.CNSWE .6.04 (2.00)
With shorting links installed:
A.SRM Hi-Hi. scram is bypassed .
(0.5)
B. Neutron Monitoring System for RPS is one out of two-twice'los,ic. (0.5) "
With shorting links removed:
Any one of.the 4 SRMse ori any one of the 8 IRMs, ori any one of the
.
6LAPRMs is capable of imposing scrams if their trip points are exceeded i.e. non-coincident logi (1.0)
REFERENCE-
- Formi Guestion Bar (6.28)
ANSWER 6.05 (2.50)
a MSIV's -
(0.5)
'.. (0.5)
- MSL Drain' Isolation valve ' . RPV low-low level Any six for full credi . Main steamline high radiation-
, 3. High steamline flow 4. High steamline tunnel temperature 5. Low-steamline. pressure 6.-Condenser high pressure 7. Turbine building high area temperature (tunnel) (0.25 each)
,
REFERENCE Formi Question Bank (6.19)
ANSWER 6.06 (3.50)
.A. Main condenser hotwell level lo All heater feedpumps of .Feedpump suction pressure Io (0.5 each)
B. Exhaust hood temperature hig Exhaust hood pressure hig Overspee Res~ctor vessel water level hig (0.5 each)
- -
REFERENCE Fermi Question Bank (6.20)
,
, ~-,r,e - - - , - ,v, , ,- - -.v.+n- ---n-----c. ,--.-.-c-w --_- - - v,-, ,.~,--r,.,,~w,m,- --,--,c,- --n-~ r-n- - , - e---.: - .- -.
.
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E PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21
CNSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 6.07 (3.00)
A. Will inject. Turbine seal leakage resulting in potential airborne activity in the RCIC roo (1.0)
B. Will inject. Pump overheating and seal damage result during low or
~
'
non-flow' condition (1.0)
C. Will not inject. Maximum signal from the flow element will result in the flow controller keeping the turbine speed at minimu (1.0)
REFERENCE Formi Question Bank (6.16)
ANSWER 6.08 (1.00)
(.25 each)
W A or for full credi . Lo ibe oil pressur i. Etr% few f%
og C #ch v+tic . Overspe 2 High crankca ressur , y ,,,f,f.o#t m .A Generator differe ' a' q d u M.pd 1 f urd b . Overcrank (time delay . g g , g (;, , ../%
REFERENCE b " ~"' N " '
' W s'E
Formi Question Bank (6.17)
WM. blU 7, !/wl Oml. jr.wa.6- M*t .
H,io k
" * "r. ," " " P ' ' "s ,
e I' s ANSWER 6.09 (3 00)
A. Feedwater flow decrease below 20% or discharge valve not full open, runback to minimum speed, 30% -- signal--euto-reset. vwi met s (1 0)
D. Reactor level decrease below level 4 and RFP trip,, runback to,45%
speed,.si 3 nal manually rese *
, j,:-[. g' pg (1.0)
C. Loss of heater drains, runback to 80% speed, signal manually reset.(1.0)
REFERENCE Formi Question Bank (6.15)
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i 6.- PLANT SYSTEMS DESIGN, CONTROL', AND INSTRUMENTATION PAGE 22
ANSWERS ~--- FERMI 2 -85/02/12-LANG,T.
ANSWER 6.10 (2.50)
A..An alarm and a Rod Block will bu generate ". > 10% difference between two flow units. 4 - ' h t ce<,,Ye Nua t:. ' > ..
High flow (100%). om i a r._ feu ,
C. Module unpluge Mode switch out of operat REFERENCE Formi Question Bank (6'.08)
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~
R565UE665C5L 66NTR L
____________________
CNSWERS -- FERMI 2 -85/02/12-LANG,T.
CNSWER 7.01 (3 00)
A. Receipt of a full scram signal and sustained APRM indication-is greater than or equal to 6%. (1.0)
B. Receipt of.a full scram signal and more than.one rod does not insert to position 04 or les (1.0)
C.-Recept of a full scram signal and reactor power cannot be determine (1.0)
REFERENCE Formi Question Bank (7.44)
ANSWER 7.02 -(1.00)
When the scram air heade'r low pressure alarms annunciates and more than ene control rod drift alarm is receive REFERENCE Formi Question Bank (7.42)
ANSWER 7.03 (2.50)
Any five for full credi . Unplanned change in core flo . Unplanned change in recirculation loop flo . Unplanned decrease in core differential pressur .A~ Jet pump percent differential pressure deviates excessively from the average of the remaining pump . Unplanned decreased in reactor powe . Unplanned decrease in MWE outpu REFERENCE !
.Formi Question Bank (7.39)
~
ANSWER : 7.04 (2.00)
Loss of squib valve continuity /A p ^l- b
- B. In]ection check valves indicate ope *
C. Pump run indication (run light and amps). ~'
D. Tank level decreasin E. Power decreasins l y ,G"' k . (0.4 each)
p, 74 , , 'c;.o (. 'e n t ,, M w e $r* :?
$, fp) * ' .! ! A* * v! ja t-kk t e ' ? N' t
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7.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~
R5656L66565L 66UTR6L
____________________
ANSWERS -- FERMI 2 -85/02/12-LANG, REFERENCE Formi Question Bank (7.36)
ANSWER 7.05~ (1.50) ,
A. Place the mode switch in SHUTDOW B. Depress both manual scram pushbutton C. Arm and depress the main turbine trip pushbutton REFERENCE Formi Question Bank (7.34)
ANSWER 7.06 (2.00)
A.~ Declare the rods inop and immediatelyi (0.5)
B. Verify one CRD pump i.s operating by inserting a withdrawn control rod at least one notch, (0.5)
OR Place the Mode Switch in the SHUTDOWN positio (0.5)
l C.. Insert inop rods and disarm electrically and manuall (0.5)
REFERENCE Formi Question Bank (7.20)
ANSWER 7.07 (2 00)
A. If Rx power cannot be determined (0.5)
OR B. If MSIV's closed--or--condenser not available (0.5).
AND Rx power greater than or equal to 6% (0.5)
AND Torus temperature reaches 110 degrees F (0.5)
l REFERENCE Forni;Guestion Bank (7.16)
.
ANSWERu 7.08 (1 50) .
Ccnfirm by at least two independent indications that! (0.5)
(0.5)
1. Misoperation in auto initiated; or 2. . Adequate core cooling is assure (0.5)
,
, , - .
. . . . _ .. - - ..
.
. . ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25
~~~~ ------------------------
RA656L G5 CAL"66NTR6t
____________________
-ANSWERS - < FERMI 2 -85/02/12-LANG, REFERENCE
,
Forni Question Bank (7.14)
ANSWER -7.09 (2.50)
.
Five for full; credi A. Torus water level above +2, (434t40^ T t 0 )
B. Torus water. level below -2, (4204 800-f44 )
' Torus water average temperature above 95 degrees F D. Drywell atmosphere average temperature above.130' degrees F-E. Drywell presure above 1 S3'psi (S N (0.5 each)
M* "T' "
/ REFERENCE Formi Question Bank (7.10)
ANSWER .7.10 (3.0V)
LCO exceeded and: Thermal power changed by more than 15% of rated thermal power in one hour, or B. The off-sas level, at the delay pipe, increased by more than 10,000 microcuries per second in one hour during steady state
~ operation at release rates less than 75,000 microcuries per 1 second, or C. The off sas level, at the delay pipe, increased by more than 15%
in any one hour during steady state' operation at release rates grater-than 75,000 microcuries per secon REFERENCE Formi Question Bank (7.05)
ANSWER 7.11 (1 00) l Ar--Verif y ' FULL-0UT*"indicatur--ts iTIUnsirruted 1snd position indicator indica te s- 8-482 . . M M vC., .(ASP]
Jr. Attempt to withdraw each control rod to the OVERTRAVEL position. , c/-J2 Verify control rod is coupled by: , ,
,
'-
~1. Observing ' stall flow' / , 2t g' ) .
~ g\, {[ c,
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2. Rod settling back to position 4 ( ,
-,,
_Lortrs;4 3. '0VERTRAVEL" light extinguishe ( .3 5)
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~
.
w r PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~RI656L665C5E~66ATR6L
ANSWERS -- FERMI 2 -85/02/12-LANG, REFERENCE Forni Question. Bank (7.02)
ANSWER 7.12 (1 00) ,
B. Read higher than actual vesses leve ' REFERENCE-Formi Question Bank (7.50)
ANSWER 7.13 (2.00)
,A. Thermal stratification, improper mixing, loop temperatures-appear normal, could result in exceedin3 200 degrees F and possibly pressurize the vesse (1.0)
p. Through.the shutdown' cooling suction and out through the suppre.tsion pool suction. Through the shutdown cooling suction and out through the
. RHR pump minimum flow line. Through the shutdown suction and out through the full fimw test valv (1 0)
REFERENCE Formi Question Bank (7.12)
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~ , ADMINISTRATIVE ~ PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 27
__________________________________________________________
ANSWERS -- FERMI 2 -85/02/12-LANGr ANSWER 8.01 (2.00)
A. For failure of a single RBM channel 1. Verify not operating on a Limiting Control Rod Pattern (0.5)
AND 2. Restore the inop RBM channel to operable within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (0.5)
'
OR 1. Place the'inop RBM in a tripped condition within the next hou (1.0)
B. A rod pattern which exists that should a control rod be moved out of the core, the core would be placed on a thermal limi REFERENCE Formi Question Bank (8.03)
ANSWER 8.02 (2.00)
A. Two cognizant members' of plant staff- one with an SRO licens (1.0)
B. 1. Must not change the intent of the procedur (0.33)
2. Must be approved by OSRO within 14 day (0.33)
3. <=5 changes per procedur (0.33)
REFERENCE Formi Question Bank (8.05)
ANSWER 8.03 (2.50)
A. Plant Superintenden Assistant Plant Superintenden Operatin3 Enginee Assistant Operating Enginee (0.5 each)
B. NSS may exclude personne ( d * C)
WSO -may -r equest NSS _to-exc4vde-pereennel . (0.25 each)
'~- "'E M *
REFERENCY DN
~
Formi 0uestion Bank (8.07)
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a ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28
-ANSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 8.04 (2.00)
A. No. pressure boundary leakage.
g B. 5 spa unidentified leakaSee L. 25 Spa total leakage average over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> perio , D .. 1 spa leakage at an RCS pressure through valve .
- E' 2 spm increase in Unidentified Leakag (0.4 each)
' REFERENCE'
Formi Question Bank (8.10)
ANSWER 8.05 (2 00)
A. The addition, removal, relocction, or movement of fuel, sources, incore instruments or reactivity' controls within the reactor pressure vessel with the head Temoved and fuel in the vessel. (Suspension of-Core Alteration shall not preclude completion of the movement to a safe conservative positio (1.0)
B. SDM shall be the amount of reactivity by which the reactor is or would be subcritical assuming all. control rods are fully inserted except for the single control rod of the highest reactivity worth which is assumed
~
to be fully withdrawn and the reactor is in the shutdown condition,
.
cold, xenon fre (1.0)
REFERENCE Formi Question Bank (8.11)
ANSWER- 8.06 (1.00)
A. Brittle fracture of reactor pressure vessel could occu (0.5)
B. Yes. Due to neutron embrittlemen (0.5)
REFERENCE Forni, Question Bank (0.13)
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.
. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29
ANSWERS -- FERMI ~2 -85/02/12-LANG, ANSWER 8.07 (3.00)
A. Prt: vents undue stress on the vessel no: les and bottom head region.(1.0)
B. Prevents creation of abnormal conditions in the idle loop jet pumps (vibration)' (1.0)
I C. To insure an adequate core flow coastdown from either recirculation pump followins a LOCA -- prevent "foolins' LPCI Loop Selection Logi (1.0) l REFERENCE
~
,
FGrmi Question Bank (8.10)
ANSWER 8.08 (2.00)
HOT S/D --MSS in S/D and >200 destees (1.0)
HOT S/B---MSS in S/U - , HOT S/B and any temperatur (1.0)
REFERENCE Formi Question Bank'(8 22)
l
. ANSWER 8.09 (2 00)
1. Unusual Even %nt ' * ' ' t* ' ' ' ' '
. 2. Ater . Site -Emergency. F ut t 4
'
4. General Emergenc (0.5 each) l l l l
REFERENCE Fermi Question Bank (8.27)
I
- ANSWER 8.10 (2.00)
.
... 95... 105... 110.... 12 r REFERENCE Formi-Question Bank (8.35)
.
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__ _ _____ _ __ __ .____5____ _ __ $___ _ _ ___ __ I CNSWERS -- FERMI 2 -85/02/12-LANG, ANSWER 8.11 (1.50)
A. 13.4 kw/ft B. Operational Condition 1>= 25% of RTP C. Take actions within 15 min. to restore LHGR to within limits (within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or-reduce power to < 25% of RTP).
(0.5 each)
t
- REFERENCE Formi Question Bank (8.40)
l ANSWER 8.12 (1.50)
'
l A. 10'
, B. 2' .
'
C. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 15% reactor power following a S/U to within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to decreasing power below 15% on a scheduled
'
Rx. shutdow (0.5 each)- 3 REFERENCE Fermi Question Bank (8.36) .
ANSWER 8.13 (1.50)
... NSS (0.5)
B. OE (0.5) (
C . C e v e s ' r.sw r, =-NSS (.,.'
~~
'# ^' #'
. . .. ,.. '
- " ' ' " ' "
(0.25)
I 'Cause unknown---Plant Superintenden (0.25) !
I REFERENCE I Formi. Question Bank (8.12)
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