ML19220B482
| ML19220B482 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/17/1977 |
| From: | Collins P Office of Nuclear Reactor Regulation |
| To: | METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7904260328 | |
| Download: ML19220B482 (46) | |
Text
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NRC FONM 157 U S.
- LEAR R ECULATO R Y COW 51 SON LICENSE
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RECCMMENCATION 00 NOT APPROVE FOR ALgI EXAMINER CATE SENIOR LICENSE C
CPERATOR Lf CENSE C
SIGNATURE 9 '/
ISCUE SENIOR UCENSE C
OPER ATOR LICENSE
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j UCENSE
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ACTE. De coe, tor and orar esamination checx sneer must se cometered for each acchcant. If acchcant has kned the acerst:eg and oral rest at
- e Semor Levet. :he c.*eck sheer should be comotered to inoicare a:Cncant's evaluanon at me ocerator 'evel.
NRC FCRM 137 page i 7 904 '6 0 3M
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,m CPER ATING AND ORAL EXNitN ATION SUf.1?.i ARY REPORT Ev ALuaTtoN 1an SRO apptocant.s unsat stactory at tne SRO L,,et. carr piete he RO column to eva<uare sponcant's
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nsatisfxtorv re conses are esclaun&.
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- 1. CP:R ATING OEMONSTR ATION 1.1 P-estartup and Instrument Checks 1.2 Conscie Operation a.
N8anipulations
- b. Undentanding
- 2. F ACILITY ECulPMENT S
a.
Maior 3
- b. Aw ailiary c.
Engineered Saf eguards S ystems
- 3. INSTRUMENT AT;CN
- a. NucMar b
b.
Process
- 4. RE ACTOR PROTECTION
- 5. PROCEDURES C
a 4.
Normal v
- b. A bnormal S
c.
Emeyncy 0
- 6. RE ACTIVITY E FFECTS (Eucept Console Coeration) a
- 7. ACMINISTRATIVE REQUIREMENTS
- 3. EMERGENCY PLAN 9
a R ACI ATICN PROTECTION 2
- ve 2 p' [j
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NRC FCRM 157 n a:s>
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U. S. SUC.ZAR REG.'1ACRY C2 tCSSIOJ REACER CFER.CCR LICE.NSE EMCSATIO!
9038 Three Mile Isl and
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Facility Tp Fractor Type Oyo _ g!,y Date Ad:'tinistered a//7 77 __
Eminer,
J. J. Su:v fpplicant
/
Instructicns to Applicant:
3 D3 P
Use separate paper for the ars. vers; staple questien sheet Points for each questien are cn tcp of the a sser sheets.
An overall indicated in parenthesis after the questien.
j score of 70% or greater is passing.
r
% of Category
\\ of Applicant's Cat. Value_
Valm _
Total _
Score Principles of Reactor Cperatien 14 13.5 11.Cl A.
14 13.5 11.(
$0 B.
Features of Facility Design 17 16 17,.9 ED C.
General Cperating Claracteristics 16 15
}3,f
$6 D.
Instr m ent.s and Centrols IU I 9l_
E.
Safety and Energency Systens 16
!5 Standard and E.mergency Operating l3.l 7 1.
F.
16 15 Procedures G.
Radiatien Centrol and Safety b.
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12 11 0 ics Cr 6
Final Grade
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A.
PRIT;CIPLES OF REACTCR OPERATIOli (14)
A.
1.
a) What is meant by the term "delayec neutron fraction *?
(1.0) b) How and why does the delayed neutron fraction change during core lifetime?
(1.5)
A.
2.
The TMI procedure for ap;rcach to criticality requires the operating staff to plot a ene ever M (1/M) during red withdrawal.
a) What data is required for the 1/M plot?
(0.5) b) How is the data utili:ed? f;o te: a sketch may be used as cart of ycur ansaer.
(1.5)
A.
3.
Assumeycurreactorigenasable..SCPMstartuprateaspcwer level reaches 2 x 10-en the intermediate range monitor.
(Time is "0"),
a) Plog the trace on Figure Al until the pcwer level reaches 10.
(1.5) b) After a two-minute "hcid" to take data at 10-8 the operator withdraws the c$ntrol rods to obtain a 1 DPM startup rate.
Plot this trace for 5 minutes; include the period of red mo ti on.
Exclain any chances during the five minutes.
(1.5) c) Assume the reactor was at 10-6 at 3 minutes on figure Al and a reactor trip occurred. Plot and discuss the resulting trace for the remaining period of time on figure Al.
(1.0)
A.
4 Xencn conditicns include periods when xenon is at equilibrium and at peak conditions.
a) What is meant by equilibrium xenen?
(1.0) b) Hcw does peak xenen develcp?
(1.5) c) Assume a plant startup and pcwer escallation is in progress at the time of peak xenon. Mcw dces this differ from a cower esca11ation af ter a one-aeek shutdown? No te: a sketch ::ay be used as cart of your answer.
It is not necessary to include values for xenen.
(1.5)
A.
5.
Water is used as a ccolant in your reactor.
List three other incticns of water in the reactor.
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B.
FEATURES OF FACILITY DESIGN (14)
B.
1.
The nuclear services closed cooling water (NSCCW) system prcvides cooling water for the nuclear ccmponents during normal oceration and also provides cooling for ccmponents during a loss of coolant accident.
a) List those components which use the NSCCW during normal operation. List the compenents (i) Inside and (ii) Cutside the reactor building.
(2.5)
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~
b) What com;cnents are supplied during a LOCA? Note:
i ccmpenents a;cear in part A, mark them xith an asterisk".
(1.0) 1pt M 4o k M hr.[ 021) fA-p
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i B.
2.
Condensate and reedwater are heated in a number of heat exchanges before water enters the steam generatcrs.
Sketch a cne-line diagram of the comoined systems from the condenser until it leaves the last heat exchanger.
Include all pumps and source of heat for each heat exchanger for one string.
(2.5)
B.
3.
The fire protection system at ycur facility consists of a number of pumps located in unit ene and two. The fire main supplies a number of hydrants, hosereel stations, a wet pipe sprinkler and a deluge system.
a) List all pumps which supply the fire main; include cacacity and when each pump will start.
(2.0) b) Explain the difference between the wet pipe sprinkler and the deluge water systems.
(1.0) 3.
4 Cne of the components associated with the reacter coolant system is the R. C. drain tank.
a) List the equipment which discharges to the drain tanks.
(1.0) b) What'linits app'y to the tank? Include the reasons for these limits.
(1.5) c) If the drain tank level is tco high, where may the cperators discharge the excess liquid?
(0.5)
B.
5.
One of tne Limits and Preventions require that systems containing liquids should be free of air before they are operated to prevent water hammer.
a) What is water hammer and why shculd plant cceraturs be concerned about water hammer?
(1.0) b) Caes water hammer cccur only with systems containing liquid and air? Explain your answer.
(1.0)
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GENERAL CPERATING CHARACTERISTICS (17)
C.
1.
The ence-thrcugh steam generators (OTSG) at your facility will producr superheated steam during pcwer eteration, a) With the aid of a simple sketch, explain hcw superheat is accomplished in the OTSG.
(1.5) b) Hcw does superheat vary frca 15 to 1C0% power? A sketch may be.used as cart of your answer.
(1.0)
C.
2.
The folicwing are two of the limitaticns and precautiens in the
~
Unit Startup procedure, provide the value and reason for each of the limitations.
a) Co not attemot to start a reactor c:olant pump when ;cwer is greater than J t__.
(2.0) b) Ouring startuo and shutdcwn when reactor Ocwer is less than 5 percent, the OTSG water level shall be maintaineo between G 3 9and Q <F9(.
(2.0)
C.
3.
Unless specified by the reactor engineer, group 8 (ApSR) will be positioned about 32% withdrawn prior to an approach to criticality.
a) With the aid of a sketch, indicate how the APSRs worth varies between 0 - 100% withdrawn.
(Include approximate reactivity values)
(1. 5 )
b) Explain when it is necessary to move the ApSRs (i) in and (ii) cut.
(2.0)
C.
4 Two of the limits and precautions for the control rod drives concern the gas concentration in the RCS and the cooling water to the mecnanism.
a) What are the scurces of gas in the R.C.S.?
(1.0) b) Why is gas concentration a factor for the CRCMs?
(1.0) c)
Is it necessary to maintain a minimum concentration of gas in the RCS during pcwer cceration? Explain ycur answer.
(1.0) d) The temperature of cooling water to the CRCM may vary frcm 85 - 105cF.
Is there a preferred tem;erature range?
Ex pl a i n.
(1.0)
C.
5.
Assume rated power ccnditions and a trip cccurs to a reactor coolant pump.
a) provide the initial steady state values for parameters listed in figure C5.
(1.0) b) Shcw th res;cnse to these parameters until conditicns stabilize.
Assume the react:r does not trio.
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D.
INSTR'JE*iTS 40 CCMTRCL (16)
D.
1.
-~
Refer to the transient in questien C.5.
Discuss the action of the centrol system (s) for each of the follcwing param2ters:
a) Load MWe (1.5) b) Pressurizer level (1.5) c) Feedwater ficw A & B (2.0)
D.
2.
Theintermediatechannelutilizesaccmcensatedicnchabber'(CIC).
a) With the aid of a sim;1e sketch, explain the ccerating principle of a CIC.
(1.0) b)
Is c mpensation required at ICC%,15%, or icwer pcwer levelsi (1.0)
Explain cur answer.
s (h A c w Uw-FMM PM An (v U_"k c) -Tr :: ;1 :----:m the " overlap" requirements between tce other channels? Explain.
(1.0)
D.
3.
The pressurizer electromatic relief value (RC-2) may be operated in the manual or autcratic mcde. Discuss how the-valve may be operated n eacn mode. A simple sketch may be used as part of vour answer.
(2.0)
Pw n w sl'.J-D.
4 Lnit 2 of TMI utilizes a baron feed and bleed controller, g
a) What are the input signals to the controller?
(1.0) b) What are the output signals of the controller?
(1.5) c) 'ahen is the controller utili:ed?
(0.5)
D.
5.
Two of the alarms en the Diamond Centrol panel are the (i) motor rotation fauit and the (ii) asymetric monitor.
a) Discuss the p rpose of each alarm.
(1.5) b) What immediate action is required by the operating staff for each alarm?
(1.5)
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E.
SAFETY AND EMERGENCY SYSTEMS (15)
E.
1.
The pcwer distributien system at T?tI-2 contains two :C buses, 2-10C and 2-2CC.
a) Use a one-line diagram to shcw what ccmponents surely these busses and their source (s) of pcwer.
Include any cross ties between these busses and supplies.
(2.0) b)
If a station blackout occurs, what leads are supplied by the D.C. bus until the emergency diesel generators be<;in to. supply pcwer?
(~. 5)
E.
2.
List the conditiens that will cause a reactor trip with four R.C.
pumps in service.
Include present set points.
(2.5)
E.
3.
Assume a loss of ccolant accident (LCCA) occurs. The LCCA ir. creases reactor buildinc cressure at 2 PSI / minute and doubles each minute (2, 4, 3, etc.).
Utilize this rate for the following:
/
2 1,p 3K a) What autematic action (s) occur during the first fcur minutes?
(1.0)
~
b) What instruments will the operator observe to assure initial and continuina operaticn of equipment?
(2.5) c) What is the postulated peak reactor building pressure in the design basis LOCA? f(/
(0.5)
E.
4.
Three Mile Island Unit 2 is provided with a feedwater latching systen.
a) List the instrumentation asscciated with this system and explain the logic that will actuate the system.
(1.5) b) List the automat'c action that will c; cur when this system is actuated. Assume rated 'c',,er operat on.
(2.0)
E.
5.
Assume a primary to secencary leak devr. loos in one of the OTSG. The leak begins at one GpM and dcubics eveg 5_ ninutes (1, 2, 4, 8, etc.)
a) List the systems that will alert the cperator. Attemot to arrange them in crder of increasing leak rate.
(1.0) b) When the c;eraters have identified the leaking OTSG, should they trip the reactor or begin an orderly shutdown? Explain the reascn for ycur answer.
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STANCARD At:0 EMERGENCY CPEFATD3 PRCCI:URES (16)
F.
1.
Assume your unit is at the pcwer level cutoff point and waiting to meet the requirements before increasing pcwer. All systems are in automatic. A malfunction occurs in the circulating water system and vacuum begins to decrease at 1" Hg/ min, a) What is the pcwer level cutoff point and when may pcwer be increased above this level?
(1.5) b) Hcw will the ICS respcnd as vacuum decreases?
(1.0) c) What immediate action is required by the cperator (assume vacuum ccntinues to decrease)?
(1.0 )
d) List the heat sink (s) available if corrective action can not restore vacuum.
(1.0)
F.
2.
One of the Limits and Precautions of the Decay Heat / Injection Pumps states "When operating in the ES mode with suction en the building sump, pump discharge must be throtled to
~ GFM to limit runcut.
Do not throttle belcw GFri as this may result in cavitation and inadequate ccoling."
a) What are the above ficw rates?
(0,5) b) What is pump runaut and what are the consequences if the pump continues to operate in this condition?
(1.5 )
c) What is cavitation and how would the operator observe this condition on the D.H./ Injection pumps?
(1.5)
F.
3.
During reactor startup, limits are imposed on the amount of reactivity which may be added above and belcw the estimated critical positien.
a) What'are these limits and hcw "ces the operator convert them to red positicn? Assume an E.
- 50% cn Group 6/7.
(1,5) b) What action (s) are recuired if tnese limits are exceeded?
(1.5)
F.
4 In the event of a reactor trip, list the manual acticns required by the TMI-2 emergency precedure.
(3.0)
F.
5.
In the event of an unanticipated criticality:
a) What symptons would alert the operating staff?
O.0) b) What immediata action is recuired by the operatcr?
(1.0)
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G.
RADIATION CCNTRCL AND SAFu.Y (13)
G.
l.
Assume you have been working in a ccntaminated area. Af ter ycu remove your prctective clotning you monitor ycurself before leaving the area.
a) Should an ionization or GM device be used?
Explain.
(1.0) b) What type of radiation will the instrument detect?
(0.7) c) What is the criteria for deter.aining if a person is. contaminated?
(0.8)
G.
2.
After working in area for two hcurs, ycu discover your dosimeter is off scale and ycu leave the area. A survey indicates a hot spot on a drain line reads 1200 mr/hr gamma at two feet.
Your distance frem t.1e line was abcut five feet.
a) Estimate the dcse wnich you may have received.
(1.5) b)
If your previous exposure to the quarter was 900 mr, what exposure limits may have been exceeded?
(1.0) c) With the aid of a sicole sketch, explain the operating principle of a pocket dosimeter.
(1.0)
G.
3.
Three of the classified areas in the Radiaticn Protection Manual at TMI are (i) controlled area (ii) radiatien area and (iii) high radiation area.
a) Define each area; include any limits.
(1.5) b) May the NRC examiner enter each & ras? Exclain your answer.
(1.0)
G.
4 In the evcnt of a radioactive spill, cne of the basic rules is to administer aid to anycne who is injured. What other basic rules apply?
(2.0)
G.
5.
During a refueling cutage you are in the reactoi
a en the main fuel handling bridge. A spent fuel element
.c
.ato the core.
a) What instrumentation n: ll alert the personnel in the reactor building of any release of fissicn products?
(1.0) b) What automatic action (s) may occur if radiation levels exceed preset limits?
(1.0) c) 'lhat actice-ecuired by the centrol reca staff?
(0.5)
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1x10 h10 3x10 i 3x10
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9x10 1x10 l
g unich d:es not decay by x cr sp:ntane:us I
fissionj
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e:uivalent to i re -
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