ML19220B466
| ML19220B466 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/05/1978 |
| From: | Collins P Office of Nuclear Reactor Regulation |
| To: | Withheld METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7904260248 | |
| Download: ML19220B466 (35) | |
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1.
CPER ATING CEMCa.STR ATICN 1.1 Pscartua and instrumem C'ects 12 Conscie C;e<atron a.
Vanipulations b.
Uncerstacc ng 2.
F ACIL:'Y E CUIPYENT e
a.
Vat er b.
Au m.li a ry
}
c.
E";ineered Safeguards S. stem i
1 (NSTRLYENTATICN i
a.
%c ear b.
Process 4 REACTOR PRCTECBON
- 5. PRCCECURES a.
Nor at
- b. 1 bncr ai
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Erne ;eacy y
- 6. RE ACT VlW EFF:C*3 r Except Conscie Operation)
- 7. ACVlf4tSTR AT1vE REGUIREVENTS 3
- 3. EME ACENCY PLAN
- 3. S A CI A TI C*4 PR CT EC'ICN
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3 U. S. NUCLEAR REGULATORY C2dISSICN REACTCR CPERATOR LICENSE EXAMINATICN facility:
Three Mile Island Reactor Type:
PWR-B&W Date Administered:
11/
/78 Examiner:
B. A. Boger Applicant:
__W
/ /M I.NSTRUCTICNS TO APPLICAliT:
L'se separate paper for the answers.
Stapic questien sheet cn top of the answer sheets.
Points for each question are indicated in paren'.heses after the questien. An overall score of 70% or greater is passing.
Category
% of Appli can t's
% of Value Total Sco re Cat. Value 13 13.3
/8 3
[dd A.
Principles of Reactor Cperaticn 14.5 14.9
/3 /
[A3 S.
Features of Facility Cesign 12 12.3 7d_
73 [
C.
General Cperating Characteristics 15.5 15.9
/I NI[
D.
Instrucents and Centrols 14.4 14.8
/2.2 7
[f/
E.
Safety and Emergency Systems 17 17.5
/f,4 N[
F.
Standard and Emergency Operating Procedures 11 11.3 N.6 f/4 G.
Radiatien Centrol and Safety 97.
25Cff'
~-
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Final Grade
'i8-115 i
o A.
PRMCIPLES OF REACTOR OPERATICN (13)
A.1 Exclain anat is meant by a "l% Shutdenn Margin."
(1.0)
A.2 a.
Cescribe hcw and why a Tcderator temperature change will affect the worth of a centrol rod.
(2.0) b.
Describe hcw and why a change in bcron ccncentraticn will affect the worth of a centrol rod.
(2.0)
A. 3 A reactor startup is in progress. Tne coerator has determined the estimated critical position. Hcw would each of the following condi-tiens or events affect the actual critical rad position (i.e., less than calculated, greater tnan calculated, or no change)? Explain ycur answer.
a.
Tne turbine bypass valves actuate 30 psig lower than nor al.
(1.0) p: cira, >
.,, j b.
Curing a batch feed to the RCS, the boric acid f.lcw is Ogem
" ' =
rather than the desired 1Ga tm.
(1.0) s' e
./
c.
Xenon is changing due to extended pcwer cperation 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> pre-viously.
Consider that criticality is to be achieved 30 minutes later than estimated on theECP.
oc;% fc._ A )
(1.0)
-8 A.4 'dould the insertion of +.1% 0.k/k to a critical reactor at 10 imps s
cause the same change in power level as the insertion of +.1%._.'.k/k to a reaccer at 80% pcwer? Explain your answer.
(2.0)
A.5 a.
'Jhat are delayed neutrons?
(1.0) b.
Cescribe ocw delayed neutren can be "seen" during reactor operations.
Ccnsider a reactor startuo and a reactor shutdcwn (or trip).
(2.0)
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3 4
3 3.
FEATURES CF F.3CILIT. SIGN (14.5) 3.1 List five different methods of detecting RCS leakage while cperating at full power.
(2.0) 3.2 a.
Why is it desirable to limit RCS pressure when RCS temperature is icw (at cold shutdcWn)?
(1.0) b.
Discuss all of the mechanical ccmpcnents available to prevent excessive RCS cverpressurization while at cold shutdcwn.
Include system lineups and/or administrative controls that ensure the varicus syste.cs are available.
(2.0)
, 3.3 a.
List eignt (3) different compcnents supplied by the Nuclear Ser-vices River Water (NSRW) System during LCCA ccnditions.
(2.0) b.
Discuss the difference between NRSW System operaticn after an SFAS signal with and without less of offsite power.
(1.0) 3.4 a.
Cescribe how the RCP cn your unit is sealed to restrict RCS leak-age.
Include ficw paths, seal descriptions, normal pressures and ficw rates and any design features that would help identify RCP seal failures.
(3.0) b.
Assume seal injecticn ficw is lost to cne RCP.
Identi fy the restricticns under which continued cperation is permitted.
Cescribe the design feature that cculd allow continued operation.
(1.5) 3.5 Sketch a one line diagram of the emergency feedwater system at TMI.
Include all normal and alternate sources of water, major ccmponents, pumps and centrol valves.
(2.0) 44r r-t e
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C.
GENERAL CPERAT:NG CHARACTERISTICS (12)
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86 3 C.1 a.
Why is it necessary to degas the maks=co-tank during a reactor shutdown?
(1.0) b.
Outline the process folicwed to perform this operation. IJdric)
(1.0) c.
Why is it necessary to perform a hard bubble degassification on the pressurizer during an RCS heatup?
(1.0)
C.2 Assure several fingers of a control rod remain in the core when the rod is withdrawn.
Explain hcw you ceuld discover this situation:
a.
Curing a reactor startup.
(1. 0) b.
During rated pcwer cceration.
Nace three (3) ways.
(1.0)
C.3 Feactor power is decreased frem 70% to 50%. Cescribe the plant response if rods are not moved as the Xenon transient takes place.
Assume all ICS substations in automatic except the Diamond rod control.
(2.0)
C.4 Explain why an APSR can add positive reactivity as it is inserted belcw 40%.
(2.0)
C.5 Curing power operation, the OTSG level is determined by feedwater ficw; hcwever, under certain conditions, level is automatically maintained or restricted to specific values.
a.
List these conditions.
(1.0) b.
Indicate the level limit.
(1.0) c.
State the basis for each limit.
(1.0) a g a.ls O:
a 1
4 D.
INSTRL.ENTS AND CCNTRCL (15. 5 )
0.1 With reference to the nuclear instrumentation:
a.
What is meant by discrimination?
(1.0) b.
What is meant by ccmpensaticn?
(1.0) c.
Mcw are the pcwer range channels adjusted to read total pcwer?
(1.0) d.
Mcw are the pcwer range channels used to generate a pcwer imbalance
( fi ) signal?
(1.0) 0.2 The Reactor Protecticn System has two key operated bypasses.
Cescribe the pur;ose and effect of each bypass upcn the protecti:n system.
(2.0)
D.3 a.
Describe the two means of red position indication.
(2.0) b.
If a control rod is suspected of being stuck, whicn scurce of indication would you rely ucon? Why?
(1.0)
D.4 The ICS imposes STU limits under certain cperating conditions.
a.
List the parameters mcnitored and, (1.0) 5.
For each, indicate why and under what conditions (e.g., high or icw) STU limits would be imposed.
(2.0)
D.5 a.
What interlocks must be satisfied to feed and bleed?
(1.5) b.
What ccnditicns will initiate a centrol rod withdrawal inhibit?
(1.0) c.
What ccnditions will prevent rod control frca being swit:ned to autcmati c?
(1.0)
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E.
SAFETV AND EMERGENCY SYSTEMS (14. 4 )
E.1 a.
Cescribe six features af the ECG's that will ensure prccer
, s -, g - -
starting (2), leading (2) and icng-term cceraticn(2) after a ce""
LOCA with concurrent loss of offsite pcwer.
(2.4)
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s b.
List the sigrals tr.at will stop an ECG after an autcmatic start under the conditions noted above.
(1.0)
E.2 'he turbine bypass valves are used to serve four functicns (each with a different setpoint). List each of these functions and its setpoint.
(2.0)
E.3 List the ccnditiens, with setpoints, that will cause a reactor trip.
(3.0)
E.4 Answer the fcllow;ng questi ns with rescect to ; cst-LCCA aceraticn:
3.
Name the M) fluids used in the reactor building scray system and why each is used.
(1.5) b.
Discuss the method (s) available to remove hydrogen frcm the reactor building atmosphere after a LOCA.
(1.5)
E.5 a.
Describe the response of the Emergency Feedwater System to:
i.
Loss of all Reactor Ccolant Pumps.
(1.0) ii.
Loss of both Main Feedwater Pumps.
(1.0) b.
'4hy is the respcnse different for the above two emergency ccnditions? (1.0)
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4 F.
STANCARD AND EMERCENCY CPERATING DROCEDURES (17)
.1 While cperating at 95% pcwer with all major systems in autcmatic control, a group 5 control rod drops into the core.
a.
Mcw is this dropped rod "seen" by the protective systems?
(1.0) b.
What automatic actions should take place?
(1.0) c.
What are the operators' immediate actions?
(2.0)
F. 2 The folicwing limits and precautions should be followed at TMI. Give the reascn for each ene.
a.
- o not atteret to startup RCps when pcwer is greater than 30%.
(0.75) 5.
?rior to a 5A dilution, source range instrumentaticn shculd be 2 cps.
(0.75) c.
When reactor coolant tercerature is less than 500*F, no more than 3 RCPs shall be run at one time.
(0.75) d.
Verify that pressurizer bypass spray ficw is maintained whenever RCS temperature is greater than 200*F.
(0.75)
F.3 Curing a reactor startuo (pull to criticality), an erroneous signal causes engineered safeguards channel A to actuate.
a.
Mcw do you terminate this erroneous actuaticn of safeguards equipment?
(1.0) b.
Are any acprovals necessary prior to taking this action? Explain.
(1.0)
F.4 Ouring rated power operation, the control rocm fills with dense smoke.
According to the emergency precedure for this situation:
a.
List the immediate coerator actions.
(3.0) b.
For four (4) of these actiens, describe hcw the same result could be achieved cutside the control room.
(2.0)
F.5 L'nder certain ccnditicns, it is advisable to initiate emergency bo ra ti cn.
a.
What is the normal ficw path for emergency baration?
(1.0) b.
What conditicns require emergency baraticn at ycur f acility?
(2.0) m 121 f
-.... - -. - -. - - -.- i
)
N f s G.
RADI ATICN CCNTRCL A14D SAFETY (11.0)
G.1 a.
List the four (4) requirements established at TMI for control over each High Radiation Area.
(2.0) b.
Name one area of the plant that becomes a High Radiation Area during pcwer operaticns.
Explain why it changes.
(1.0) c.
Describe the additional control that is placed over areas that exceed 1000 mrem /hr.
(1.0)
G.2 A suspected leaking fuel element is removed frcm the core after one year of pcwer operaticn.
a.
What three different radiological prcblers will this element rep resent?
(1. 5) xO b.
In ceneral, hcw will this element be tested and transported to the spent fuel pit?
(1.0)
G.3 a.
A worker is required to perform maintenance six feet from a small crud trac which reads a r/hr at 4 inches. Mcw lcog can he work without exceeding his quarterly administrat'ive exposure limit (witncut extensions)?
(2.0) b.
At TMI, what is the maximum quarterly dose a worker is allcwed to receive ? Include any requirements which must be satisfied p ior to exceeding routine limits.
(1.0)
G.4 What autcmatic acticns occur for the follcwing radiation monitors upcn a high alarm? Se specific.
g;~ o r a.
Stack Mcnitor Auxiliary and Fuel Handling Exhaust Duct (F3-A8).
(1.0)
C g-b.
. dis-eetitres-4emp Discharge Monitor (FPM).
(0.5)
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