IR 05000354/1985048
| ML20153C721 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 02/06/1986 |
| From: | Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153C719 | List: |
| References | |
| 50-354-85-48, NUDOCS 8602210575 | |
| Download: ML20153C721 (96) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 85-48(0L)
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I FACILITY DOCKET N0. 50-354 FACILITY CONSTRUCTION PERMIT NO. CPPR-121 LICENSEE: Public Service Electric and Gas Company 80 Park Plaza - 17C New&ck, New Jersey 07101
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FACILITY: Hope Creek Generating Staticn
EXAMINATION DATES: October 14-23, 1985
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CHIEF EXAMINER: .
jQ Dave lange, Le
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eacto Engineer i 17 Dake k
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REVIEWED BY:
Robert M. Keller,
[ /!5/[b i f, Projects Section iC Dite'
APPROVED BY: .
U6 [
H/rrf B. K s er, Chit . D8e ( ~ '
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Projects B nch No. 1 SUMMARY: Operator and Senior Operator Initial Cold License Examination were 4 conducted at the Hope Creek Generating Station from October 14 to October 23, i 1985. Eight (8) Reactor Operator, Thirteen (13) Senior Reactor Operator and j one (1) Instructor Certification Candidates were examined. All candidates
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passed the oral examination. One (1) Senior Reactor Operator Candidate and
two (2) Reactor Operator Candidates failed the written examination. One (1)
i Senior Reactor Operator Candidate failed the Simulator Examination. The can-i didate failing the Simulator Examination showed deficiencies in communication,
- plant direction and control and the use of emergency operating procedures.
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REPORT DETAILS TYPE OF EXAMS: Initial EXAM RESULTS:
l R0 l SRO I Inst. Cert l l Pass / Fall IPass/ Fall l Pass / Fall l l 1 l l
I I I I I I Written Examl 6/2 1 12/1 1 1/0 l l l l l l l l l l l l Oral Exam l 8/0 l 13/0 1 1/0 I I I I I I I I I I I ISimulator Examl 8/0 1 12/1 l 1/0 I I I I I I l l l l l l l Overall l 6/2 1 11/2 l 1/0 l l l l l l
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1. CHIEF EXAMINER AT SITE: D. Lange, USNRC 2. OTHER EXAMINER: F. Crescenzo, USNRC a A. Howe, USNRC (Trainee)
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L. Banavitch (Trainee)
G. Robin,on, USNRC (Censultant)
- B. Hajek, USNRC (Consultant)
T. Morgan, USNRC (Contractor EG&G)
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1. Summary of generic deficiencies noted on oral exams:
Generic strengths were recognized by all examiners in the candidates systems knowledge, plant component familiarization and responsible attitudes.
Generic weaknesses noted by examiners were in the areas of knowledge on flow signals to APRM/RBM, ADS logic, Feed Water Level Control, and fire protection. Also observed was the overall lack of use of abnormal procedures and operation /use of bailey controllers.
2. Summary of generic deficiencies noted from grading of written exams:
Generic weaknesses noted in grading of the written exams were an overall weakness in Section 6 and Section 8 of the SRO exam. Specific weaknesses were noted for the administrative requirements on fire brigade manning exemptions and power limitations. Also noted was a lack of knowledge in use of Tech Spec 3.05.
3. Personnel present at Exit Interview:
NRC Personnel D. Lange, Chief Examiner F. Crescen:o, Operator Licensing Examiner L. Banavitch, Operator Licensing Examiner B. Turner, Operator Licensing Examiner R. Blough, Senior Resident Inspector NRC Contractor personnel T. Morgan (EG&G)
8. Hajek (NRC Consultant)
Facility Personnel D. Hanson, Manager, Nuclear Training B. Gott, Hope Creek Operations Training R. Salvesen, General Manager, Hope Creek Operations G. Mecci, Pr.ncipal Nuclear Training Supervisor, Hope Creek Simulator
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4 4. Summary of NRC Comments made at exit interview:
1. Now that Hope Creek has determined the type of ARM system to be used the NRC wants training on this system started.
2. The surveillance tests identified as ready to be used for the sim-ulator exam were not all available. The next set of candidates
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should be trained on the procedures that will be available.
3. The simulator performed well and the number of malfunctions not available for the exam was small.
5. Changes made to written exam during examination review.
(See Attachments)
i Attachments:
l 1. Written Examination and Answer Key (RO)
2. Written Examinatics and Answer Key (SRO)
3. Facility Comments on Written Examinations made after Exam Review
4. NRC Resolution of Facility Comments on Written Examination i
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" A3 t. b3 9pn ph fyr twHeb (0 2 5 e.a.)
2. Meh. twitch i n EuN M cerwp W m co,23 Arg6 nd deutWeah.
e
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3. INSTRUMENTS AND CONTROLS
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PAGE 30 ANSWERS -- HOPE CREEK -8 5 /10 /15 - B o rJ A V I T C H , L / L AN C E . D AN5WER 3.09 (2.00; 1. 3 Indicates LPRM suitch posttien 3 (1.0)
B indicates the c::161 position of the detector, c e t o n t, f r o n. the b o t t o n.
24-33 indicates the rad:al .: y coordinates cr the d e +. c c t a r 2. The &mplification of the LPRn s i g n a l s n.u s t it.c r ea se oser core (1.0)
life to a c c o n.c d a t e for the dec r e a s ing sen s i t t '. i t y due ta urantun, depletion.
REFEFENCE 302HC-000.00-015-01 Guasticn +3. 4 LF 15 paje 14
_. -
AN5WER 3.10 (3.001 1. a.
,
.. b. c. -
3. a. c. d.
f.J.
- A > c- ;-
L8 toll 5 -
~ g o,py 5. a. * ~ ~
-
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if G-O f Tach)
R E F E R E tJ C E ~
HCC2, LP, ~RM, 302 HC- b .0C-014-0,1.'pg 19, F;j 6. and A .: ; h , 30: H C - G t 0 . 2 0 --
016-01 T a b i r- II. Fig 5
+ .w
.c.t J - n- a . ,. .,
.
..
.. ..i Monitori s *. a r. t c toverter cutput anc g tches to b;c< u ,- AC :pply w h c r.
loss of Inverter outcut is 2ndlesteo. * * * 1m e erT~. .
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' .25)
' LC 10(19 REFE:ENCE LP 66 1E At power supp1v
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4. PROCEDURES - NCRMAL, ADNCRNAL, EMERGENCY AND PAGE 31
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~~~~RAU5UEUG5C L'UUNTRUL
ANSWERS -- HOPE CREEK -6S/10/15-EANAVITCH,L/LANGE,0 ANSWER 4.01 (3.00)
3. 1. Reacter power drops below 5% (0.5)
2. RPV water level resches TAF :0.5)
3. All SRVs remain closed (0.25s and crywell pressure r e.t a i n s '
iO.5)
below 1.68 psis (0.251 b. Concesstrate tha bcron (0.51 and enhance void generatico 10,5) (1.0)
c. You wouldn't ter minate injection until reaching 5*. of the SLC (0.5)
tant volumg and only if it c i c r.' t autenatically. trip.
REFERENCE Coh5(dtr KC/ d-l ,10 t (%r CMit LU 18ttS OP-EO-ZZ-307, OP-EJ-ZZ-207 Lesson Plan Objective 13 g(Asctsv- rm.kcrt'iTc d or ridt ( rs yatt 0 2. g g:(7lg ANSWER 4.02 (1.50) M LN MI b M M MC * W N'
Follcuing the ficodino of the vessel (0.25), the flow path would be*
MAm Sithm Lim ( to /gi LE)
6 's to suppression pool via SRVs ( 0. 251 - ~ ..
-
Suppgession pool t,c vessel via Core Spray (0.25; or LFCI (OctS)
.. H e s t is teroved from supp,resu on Pool by suppression chanber~rooling
--
mode of RHR (0.5) --
.
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REFERENCE OP-EO-ZZ-205, page 5 AN3WER 4.03 (2.00)
1. S c r -s e. . o r.c i t i o n (0.~5) a r. d r -setor power Qor undetcrm: nod (0.25;
.
2. RFV water level below -35' (0.25) or e n d e t e r m i r.e d (0.25)
3. Reactompr e s sure above 1037 ps:3 (0.5)
4. Dryuell pressure cbove 1.68 ps29 (0.5)
REFERENCE Reactor fressure Vessel (EPVi Control (OF-EG-Z'-101: . p a g ,; ;
, .,- -
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4. PROCEDURES - NORhAL, ABNORnAL. EhERGENCY AND PAGE 32
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~~~~EdD55L5555dL C5sTR5L
ANSWERS -- HOPE CREEK -85/10/15-BANAVITCH,L/ LANCE,0 ANSWER 4.04 (2.50)
a. Following verification by twc independent indications; 40.5)
1. hisoperation in a u t o n.a t i c n. ode is confirmed, or (0.5)
2. Adequate core cooling is assured. (0,5)
b. 1. haie frequent cheels of in2t: sting parameters. (0.5)
2. Festore the system to automatic / standby mode if possible. (0.5)
_
~
kO R,tr t e:ENCt M C83 h ALCC)> d f0r- I2. c.d( , LG u(7l6 Scram procedure (OP-EO-ZZ-100) Note 10. .
AMSWER 4.05 (3.00)
a. The CFD pump will increase water level (0.25) anc there is no so.5)
outlet flow path established. (0.25)
b. Because cooling flow is Icst to the re3cnerative heat e ::c h a n g e r (1.0) __
(0.25) increasing the outl tt temperature to the NRhX 4 0.25 ) . ~
' '
possibly.
-.
causing an isolation of the system (0.5) .
.
_
c. Hot shutdown * .; with no recirculation pumps operatin3 . ..; L3
.
(1.5)
tc n.i n i m i c e thermt1 strattf;;ation of vessel water '~ ! ;
l\fzo[IT REFERENCE .
Prep f.; r I s n t Starto;; (ZZ-002> pa3e 5 RWCU LF ::;-01) pages 56 57 AN5WER 4.05 i2.02)
,
1. Suspeno all refueling opersticos. (0.5~.
2. E.scoste all uneces s ar y per sonnel f r om the E c- f u e l Floc: . (0,5)
3. Ensure the Reactor Euticinj V.'n t i l a t t e n . Iss l4.% t,LT u ("I fi r (C.5)
4; Ensu e th m the Filtretion+ Recirculation. sr.d Ven t : l at: or. 0.5)
s y s t .e . F. F P3 ) eute starts.
REFERENCE Irradistec Fuel Osmage White Refu.'line (OP-AB.ZZ-101)
y. . .<-
.
.~.
t 4.
-
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 30
RADIOLOGICAL CONTROL
.._ - - - - _ - - - - - - - - - - - - - -
ANSWERS -- HOPE CREEK -E5/10/15-CANAVITCH,L/ LANCE,0 ANSWER- 4.07 (2.00)
1. All operable APRMs are indicating between 4 and 12%. (0.5)
2. All APRM DNSC lights are off. 2t. $* L.y u pf ry- (0.5)
3. The Main Condenser vacuoci is greater than (W) HgV. (0.5)
4. Reactor pressure is greater than ( Igr ) psis. (0.5)
'TTC REFERENCE Startup f r o c. Cold Shutdown to Power (OP-IO.ZZ-003) page 16 AMSWER 4.08 (3.00)
0.73 a. 1. Scram the reactor. M . S c. ,
2. Trip the main turbine. W 10W s o _5 9.W
.(-0.05 for the auto actions; these address contrcl roca.
ventilation isolation and the start of the control room (0.T3)
. emergency filter uni ,.,
32 VC.Vih AM SAAh d4hmt. i h&vt occu.vt.d.
)
. _ . , b. HPCI wiil continue t.o operate as is. (1.0)
^::C ~_;. : ;,_ r n " m--2 2: es=+ < :: *: - .s. . _ (1.0)
IECtC Wtn cmti* br cye.M<_ aJ R.
REEERENCE .. .?
Control Econ. Evacua, tion ( -AB.ZZ-130) page 1 Shutdown f rom Outside tue}30 Lontrol R e+ m (C+-1022 N 00) page 3 and Atte'ch. 1 ANSWER 4.07 (1.00)
1. Lasve the area immediately via ncemal access procedures. (0.5)
2. Notify the Radiation Protection scetion of the clare. letation. (0.51 REFERENCE SA-AP.ZZ-024(G). page 33
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4. FROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
- - - - - - - - ~ ~ ~ - - - - - - - - - -
~~~~E D5UL5EiCit E5srR5t
ANSWERS -- HCPE CREEK -E S / l G /15 - C AN A'.'IT C H , L / L A N G E ,0 ANSWER 4.10 (3.00)
a. If an SRV were opened with the supression pool level b31cw the (1.0)
top of the SRV T quenchers. rspid p r e s s u r i c a t i ori of the supression chanber and drywell wauld result. It i; possible that the desi3n pressure f c. the containment would be exceeded.
b. M a i r. Concenser r, S L D r a i r.s (1.0)
Head Vent HPCI RC:C S.in E RFPi 4 0.25 each, 4 neeced far full credit)
c. 1. 50 psig is the lowest d i f f e r e r.t ; a l pr essure at which an SRV (M)
utli remain fully open with its contrc' switch placed in the 0.77 open position, g gg(g 2. Selow 50 pstj, RPV Depr.essuricatien is consic+; red complete '.5 and additional steam paths are not necessary. o.25
.A -
REFERENCE .
OP-EFM Z-20: LP : ages
^
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a. o, 7 / b. page'S~/ c. p .3 3 g' E
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AMSWER 4.11 _
(2.00) '
-
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AGE s. ADMIN b. r;FC'
~ ~ .
g ;' c. r e r / q t t 1:5 a t e n. / : t :
g :X m r e n .' : . r 1250 . r e n e q +. r g 1000 :. r e .s / a t r 1250 r!cw qti 36 *, r 107: c. r e n.
( l. 0 ) 5 ( n ~ 10 ) = 2 R E .1. ' y r (l.0)
or 500 rw-o r. . ' q t :
REFEREtJCE SA-AF.ZZ-02J pg. :0-43 and ICCFR C (g gg l 3 l g.y 0 ~ \ E~ \1 SYC y p<.h-ftdA t.vt.d(t fav-con-ccb anrwtvi ccmcernig 36 tye.r old .
l l
_=- --_ - - -.
.,O I.
_ , Qu.eshon 3 Jl Aqpe (Ai4mx4<.youe.v)
[Normalyower) 125 V bc.
( 5"fPyl j e
4fo Vu 12sv ot -
1 y g.6e r i
'> Auctim '< Izs v 'a V '
6 tic lavtykr
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( Fuky Yewey fuyy\g) V 120 voc
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.
490 V g . . gQt. Sgg
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I 20 Vg (dUIMbution yd)
,_
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9_t Lk @ , yacjo is-17
- .. . . . _ _ -
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t TEST CROSS REFERENCE PAGE 1 DUESTIDN VALUE REFERENCE
________ ______ __________
01.01 1.50 DJL0000474 01.02 2.00 DJL0000475 01.03 2.00 DJL0000476
"
01.04 .00 DJL0000477 01.05 1.00 DJL0000478 01.06 1.C0 DJL0000479 01.07 1.50 DJL0000480 0*.08 1.00 DJLO3004S1 01.07 1.00 DJL000045:
01.10 3.50 DJLC0004C3 01.11 3.00 DJL0000404 01.1; 2.00 DJL0000455 01.13 2.50 DJL0000486 01.14 2.00 DJL0000467
______
25.00 .
02.01 2.00 DJL0000483 C2.0 2.00 OJL000048?
02.03 2.25 DJLOOOO4?O 02.C4 2.50 DJL0000491 --
,_
02.05 _.2.00 -D-JLOOOO496 -
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02.06 2.50 DJL0000477 ..
-
_02.07 .. .3!00 * DJL0000513 '
02.05 3.00 DJL0000514 02.07 1.75 DJL0000515 02.10 3.00 G463C00516 02.11 1.00 DJL3000517
__-___
25.0C 03.01 2.25 O J L O C 00 4 .' 3 03.02 3.00 DJL0000404 03.03 2.CC DJL0000495 03.04 1.00 DJL0000475 C3.05 2.00 DaLG000516 03.06 .7 . 5 0 DJL0000515 03.07 3.00 DJL00005:0 03.05 2.00 DJL0000521 03.09 Z.00 DJL00005:2 03.10 3.00 DJL0000523 03.11 1.25 DJL0000524
______
25.00 04.01 3.00 DJL00004??
04.02 1.50 DJL0000500 04.03 2.00 DJL0000501 04.04 2.50 DJL000050:
04.05 3.00 DJLOD00504
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t TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE
_ _._____ ______ _ ________
04.0c 2.00 DJL0000505 04.07 2.00 DJLOC00506 04.0L 3.00 CJL0000507 04.07 1.30 CJL3000511 04.10 3.00 DJL0000512 04.11 2.00 OJL0000527
__- . _ _ -
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______
he-p___h 100.00
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U. S.
NUCLEAR REGULATORY COnttISS10rJ SENIOR REACTOR OPERATOR LICEt4SE E X AhitJ A TION FACILITY: HOPE CREEK
_____________-_____-_--__
REACTOR TYPE: E. W R - G E 4 DATE ADhIt4ISTERED: 85/204,15 EXAnINER: CRESCEHZOi F.
APPLICANT: _ _ [_ _ ____________
ItJ 5 TR UC T ION E TO APPLIC At4T .
Use separate pacer for the answers.
Staple question sheet on top of the answer Writ.esheets.
answers an one side oniv.
Poirits for each cuestion are indicated in parentheses after the cuestion. The c a s s i n -a grade least 6 0requires
- . . at least 7 0 *. In each category and a final 3rade of at E:: a m i n a t i o n papers will be picked up s t ;' tc) hours after the enamination starts.
CATEGORY '.OF
' -: 0$ ...
AF'FLICANT'S._ CATEGORY -- ~
VALUE TOTAL SCORE ~
________ _ _ _ _ _
VALUE CATEGOR.
___.________ ________ '
m e.
""*U".
.
'c" * O '5 MW
___________._,____________________ _
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5.
TiiE O R Y OF NUCLEAR POWE6 P L A rJ i OPE R A T I 0tl . FLUILL, AND TiiEn h00 f tJ An :L 3 25.00 25.06" 6. Pla t4 T Sr5 TENS DESIGN, C0tlih 0L e A r4 D IflS TRUnE tai A!!ON 25.00 2 5. M
.
FROCEDUEES - NORnAL, AE tJ0hn Al e LnERGEtJCr A t< D RADIOLOGICAL C 0 tJ T 6 0 L 25.00 25.00 6.
ADt1ItJ15 TE A T IVE F RUCEDUPE5 e COrJDI T 10N L e Ar:D LIhI T A T 10rJL 100.00 100.00 TOTALE FIflAL GRADE _________________%
All work 32ven nordone receisec on this euamination is mv own. I havo neithor aid. 1
,
APPLICANT'L SI G t4 A T U R E
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PACE ?
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GUESTION 5 01 (2.25)
As the reactor is taken f r o n. COLD SHUTDOWr4 to RATED OPERATING CONDITI0t45, HOW :re the followinq affected and WH'(?
a. The hAGNITUDE of the h0DERATOR TEnPERATURE COEFFICIEt4T.
b. DIFFERENTIAL CONTROL ROD WORTH.
c. Tne m A GriITUD E of the FUEL TEMPERATURE LOEFFICIENT (Doppler).
QUESTION 5.02 (2.00)
During a reactor startup, criticality is achievec when a cositive perlod is maintained without further positive reactisity additions.
The definition of critical states Keff equals 1.0 and reactivitv equals 0.0 and period would therefore be infinite. WH1 then is the reactor declared critical when the period is cositive?
_
GUESTI0t4 5.03 x ^4 24 -
-
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_ ,
The Reactor is on a 100 second_ period; hoc e r a to r tamperature'is 160 deg F. With no-operator action, WHAT~will*ne tae r..o d e r a t o r temoerature when the reactor is again on an infinite oortod?
Assume; E:0 L for time in Reactor life.
State anv a s s u n.p t i o n s vou make.
SHOW ALL WORK.
GUESTION 5.04 (2.00)
The Reactor has baan operatinc at 45% cower for several cava. An oper6 tor RAPIDLY reduces resttor power to :,0 7 by : educing the speed of the recirculation pumps. During the next 2-3 mINUIE5 tar noerator notices that the reactor poaer slowly incteases a p p s o ;: 1 ro a t e l ,
3*. EXPLAIr4 the cause of tnis effect.
(ass ** CATEGOR) 05 CONTINULL ON N E ;:1 PAGE Aaaa*)
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5. THEORY 0F NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
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GUESTION 5.05 (2.00)
Regarding nCPR (hinimum Critical Power Ratio);
a. What FHEN0nENON could exist in a fuel bundle if it wore operated at a HCPR LESS THAN ONE (s 1.0) arid WHAT w o u lrj very likelv de the CONSEQUENCE of the phenomenon?
b. WHY aiu s t the T e c hr. i c a l Specificatiori hCPR limit include a k factor wnen core flow is LESS THAN RATE 09 OUESi10N 5.06 (3.00) ,,,
,
Attached Figure 2 shows selected clant carameter responses for a IUR E:IN E TRIF transient initiated froa rated conditions with NO OPERATOR ACTIOrt. Answer the following:
-..
NOTE: (1) Use of all traces may be required to de t e r m i tie the correct answer.
(2) Another malfunction..may be present. -
a. Whv does core flow decrease EFoint 13 and whv doesn't it decrease to zero [ Point 2]?
b. Whv does reactor pressure increase CPo t tet 32 and remain high EPoint 437
) c. Wnv does reactor level decrease initiallv [Foint 53 and what is c a u s t rig the peaks in level later E P o i r. t 6]?
e GUES T I0 re 5.07 (1.00's HOW ts concensate depresston .s f f e c t e d (ItJCRE ASED or DECREA;CDs bv the followin3 changes I r. the circ weter flowirij throu3h the condenser:
CAssun.e all other parameter remain constant.]
ECoris a de r each enan3e seper a tel y.]
a. Flow decreases, b. Temperature dect eases.
(***ss CATEGORY 05 C0tJ TINUED ON NEXT FACE *****)
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND FACE 4 THERh0DiNAMICE QUESTION 5.08 (1.00)
A centrifuqal pump is forcino water throuqh a civen lenqth of 2-inch diameter line. A r. operator revalves the system so that it now puraps through the same length of 1-inch otameter line.
Choose one of the follow 1 rig that best describes WHAT happens to the p u n. p ' s DISCHARGE HEAD, PRESSURE DROP in the line, and the FLOW.
a.
Discharge head I n c.r e a s e s , pressure drop increases a rid flew I r.c r e a s e s .
b. Discharge head decreases, pressure drop decreases ano t' low cecreases.
c.
Discharge head i r.c r e a s e's , pressure drop t rec r e a s e s a rid flow decreases.
d.
01scharoe nead decreases, pressure drop decreases and flow tnereases.
,
QUESi10N 5.09 (1.00)
In terms of entracting enerqv from steam, jgICH of the followinq statements best describes WHY the condenser le operated at a 19P'
VaCoum. -
.,~
a "w. u... *
r. n - . -
a. Less enerqv is e:ITr a c t ed from t h e -s t e a m but overall piant efficiency is increased. *
^
b. As the vacuum is increased, the saturation temperature o t' the steam is Cecreased, allowing more erie r g y to be entracted.
c. As the vacuum is increased, the saturation temperature at t rae steam 15 1 ric t e a s e d . allowing more erie r gy to be entr ac tec.
d. The amount of enerqv extracted from the steam is not d e p o n c e n t, on condenser vacuum.
(****s CATEGORf- 05 CONTINUED ON NExt PAGE eeaes)
[ .
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE D THERh00iNAhlC5 GUESTIOt1 5.10 (1.00)
The fission process in a commercial reactor requires the neutrons that are ' born' by f i s s l o ri to be 'thermaltred'. The interaction in the reactor core which is most efficient on thermalizinq neutrons for fission occurs with the....
a. C X 'r G E N atoms in the water molecules b. E.ORON atoms t re the c o r.t ee l rods '
,
c. ZIRCONiUh a t o ce s in the fuel claddinq v
d. HYDROGEN a t o ai s .1 r. the water molecules
. -
QUESTION 5.11 (2.00)
The temperatine sens r on theqdhgcharge of the Reactor Water Cleanup 5 y s t e n. (RWCU. NIcurren ly reading 100 F but is suspected of being inco r r ec t .- Given that 133,000 lbs mass per hour ofaceagtor fluid enters the r M M entative heat e x c h a rs g e r of the RWCS atT temperatur e of 260 F. also 256,000 lbs mass per hour enters the.shell side at '"' L-80 F and exits.at 140 F. . _ ~
a. WhAT should the discharge temperature readinq ba? (1.001 b. SHDULD En isolation have occurred? ( I ric l u d e setpoint.) (1.001
. --
GU E 5 T IOr> 5.12 (1.50)
A reacter startup is in progress. How would each of the cIlowinq c orio t t i oris or e v e ri t s affect the actual critical roi posittori (more rcc withorswal, less rod withdrawai, or no sionificant etfect'3 s. D r. e r esctor r e c i r c u l s t l a r, pumF .5 5 topped. (0.00'
b. Aenor. 13 c n a r. 21 n g ( e n t e rid e d power operatton was terminated 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> prr,1ously). (0.50)
c. nocerator temperature tf araduallv decreastnq. iO.50.
tesex: CATEGORY 05 CONTINUED ON NEXT PAGE ***ss)
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6
T H E R n 00 ! r4 A N IC S
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GUE5 TION 5.13 (2.00)
With regerd to nAPRAT.
a. WHAT is the RELATIONSHIP betweer. HAPRAT & nAPLHGR9 (0.75 b. Tne process computer prints out a hAPFlAT of 1.05. Is this acceptable?
(0.50)
c. WHAT physical corisequer ce could occur if the MAPRAT 11 pil t is e::c e ec ed ? '" -
(0.75)
DUESTION 5.14 (2.00)
Concerning the Linear Heat Generattor, Ratto (LHGR), match a nun.bered tern. to an apptoprIEte' lettered staten.ent. '4 answers a-d)
1. m a ::1 m u n LHGR _ 5. t o t a l -p e ak inq factor
,2 local peaking factor 6. local power spite 3e clad spik inq 7. average LHGR 4. pellet clad interaction D. NFLPD a. The power that would ce produced in every foot of fuel rod lenoth if al' fuel rods produced the s a ra e fraction of total power.
b. Determines the highest powereo rod at a particular node and is a
,f u n c t i o r. of ivel t,pe, e,:p o sur e , vold fraction, contr ol red conriquratton, and surroundinq fuel t v rs e .
c. The c oric ent r a t t ori cf stress, 21rcaloy e mb r i t t l e n.e rit , &nd pellet cra:kinq leadtnq to stress corroston crackinq a n r' subsequent clad failure.
d. Gap oetween two fuel pellets to e :: p o s e a greater surface area to neu t t or. f l u, .
s***** EtJD OF CATEGORf 05 as4*s)
/ . .
6. PLANT SYSTEnS DESIGN, C 0 tJ T R O L , AND INSTRUMENTATION PAGE 7 00ESTION c.Ol (3.001 With reqard to thq Htqh Fressure Coolant Insection (HPCI) Svstem:
a. WHICH HF CI ,7rsp -
.s
, signal will be automatically r eset a rid retntttate HPCI tf required? (Without operator action.) (0.50)
b. WHAT would be of t r n.e d i a t e c oric e r ti If the HPCI n. t n i a. u n. flow control valve FAILED OFEt4 followinq a HPCI turbtne trtp? (0.50)
c. With the system t ri * standby,' HOW is the pump discharge pipinq maintained full of water) (Include source of water.) (1.00)
d. WHAT would be the status of the HPCI System if the r a nip cenerator failed at its low l i n. t t ? (1.00)
GUESTION c.0I s3.00)
The reactor ts operattnq at 50% power when there is a sudden electrical load decrease o r. the grid of app r o;;i ma t e l y (~15%).
HOW will t h eemma t n Steam F r e s s u r e -Cth t r o l svstem CEHC] c o n.n e n s a t e for this loss of lord? (Refer to the attached EHC Lostc Diagrim.)
CAssume that all reactor svstems are in their normal a t- p o w e r lineups, 3 rio r e ac t o r s c r a n. occurI a ri d ri o operetor a Rion is taken.]
Limit vaur e ::p l a n a t i on to the R E S P 0 rJ S E of the ri a t n S t e a m Pressure Control s y s t e n. [EHC] c r.d state toe f i r.61 steady state corid t t t oris o f the followine par.36 ters: 6.- Fower, R: fressorer Gen. Fower, Gen. R P ri , and OPV's. . N8
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CATEG0Fr 06 C 0re T ItJUED O rd rJExT PAGE ***ss)
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6. PLAT 4T SYSTEh5 DESIGN, CONTROL, AND INSTRUMENTATION PAGE B GUE L T10t4 s.03 (3.00)
a.
What is (are) the source (s) of operattnq qas to the Inboard and Outboard nain Steam Isolation Valves? Ir.clude NORM AL and E: AChUF'
supplies as applicable.
(1.00)
b. The p rie um a t i c a c c u n.u l a t o r associated with each MSIV 1s stred to allow HOW nUCH valve operatton?
(0.50)
c.
The two normal operatin3 s o l e reo t d s for each MSIV are powered bv WHAT TYFE(Se of power (AC, DC, 24V, 43V. 120V, etc.)? (0.50s d. How would the hSIV respond to 1. A total loss of power to the valve (0.50)
2. A total loss cf pneumatics to the valve (0.50)
GUESTION 6.04 (2.00>
Regarding the Restdual Heat Removal (RHR) Svstem while operatinq in the Shutdown Cooling (SDC) hodei Sr -
~_--= .
M
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a. EXPLAID whv'It . ts nece3sarv to prevent the RHR pumps discharoe
--flow from decreasing below 1400 3pm.
6. EXPLAIt4 whetner the tollowinq actions WILL or WILL NOT occur tr vessel pressure were ts t r.c f e s s e s to 90 ps29 1. Shutdowr. coolinq suction salve (F006) auto closes.
2. All r u n r.: n g R H F. pumps trip.
3. RHR s. ump suction valve (F006 A < E. > auto closes.
4. C,utboa r c nese spr ay .alve s F O .' 3 s euto clo:,es.
i aaaa* C6 {[CQf. t 0 r, C uti { [titJ L D Ori N[AI fACl *****)
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6. PLANT SiSTEh3 DESIGN. CONTROL. AND INSTRUhENTATION PAGE 9 GUESTION 6.05 (2.00)
TRUE or FALSE a. On the Rod Worth hintminer (RWn) S v s t e n. , If an insert block is present, ther. three cor trol rods insert errors have occurred and all three rods are positioned two odd notches past their insert limits.
b. When decreasing Reactor oower into the RWn noorable range the Rod Group Window will display the highest 3r cup which has less than three insert errors AND at least one rod withdrawn past its m i r.1 m u m 11a.it.
c. The select error l a c. 9 , on the RWn displav o a ri e l , wtli illuminate w h e r.e v e r the selected rod is r. o t r e sp o r.s t b l e for the current red block.
d. If Reactor power is decreased such that the RWM svstem becomes op e r a t t oria l with greater than the naximum a n.o u n t of Iriser t and withdrawal errors allowed, n o n o r n.a l rod movement ts oossible, unless the latched group c o n t a t ris a c o re t t o l rod c a u s t rig a re t
error.
in s eg;y/wo lt~t h d r a w a l~- .
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QUESTION o.06 (2.50i
.
Answer the followinq ovesttons based ucon the sttiration Gescribed below. The RRC5 15 fuli f o p e r e t i o r.a l . Thc RRCL receives a reactor Hlan Fressure i1071 ostas stanal t r. both complementarv loqtes of a RRC 5 ch ar.rie l a r. d r ett.a.tr E in f or c. 4 s e c orio s . It taies L7 seconds tram the tnttial reactor Haan Fressure stanal before the APRn levels are dowr.s c a l e .
a. WHICH or the four laatcs inteqrateo into 66CG are actuated 1 : 0 s e t o rid s "-
b. WHICH loq1cs are a c t u a r. e i2 at T=10 s.concsi c. WHICH 1031 s ar e a cti>a te c et is3 seconds d. WHICH loatcs are actuated at Td3 wenn.1x-e. HOW LONG t r o n. T0 sec or ds 13 at t.e t o r e the RRCb c ari be reset?
(aaens CATEGGhr 06 CONTINUED ON rJE A T PAGE aaa**)
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6. PLANT SiSTEh5 DESIGN, CONTROL, A tJ D INSTRUMENTATIOt4
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I' AC E 10 GUESTION o.07 (2.00s The rectreulation system wtll respano to solocted plant condtttons by automatically reducing the r e c t r e u l ti t i o n n/G set generator output.
For each condtttons listeo below, identtfv WHAT wtll occur, a FULL r u rib a c k , INTERhEDIATE r u r.b c cl . or NO runbac6 a. Secondarv Condensate ocmp tr15 w/FW 65%
b. P r i n.a r y Condensate pump trtp W/FW , C S *.
c. RFFT trip W/RPV level at +25*
d. F e e r;w a t e r flow at ICL for 20 s e c o rid s e. Loss of stator c o o l t rea 4- 3 L u. t o u i. . i.L..$_
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h. RPV level = 10' -
GUESTION 6 06 (3.00)
For each of the Reactor Protection isstem (RFS) t r t c, s in column 'A'
ider tif y which of the $ m(s) in c o!vr n ' E. ' will b/pess thz.t tr.p functton.
. /
( r40 T E : There may b e* m o r e than-ono wa. to b v r. a s s each t r i r, f airic t t on . i
'A' TM, IRIPL 'L' aE Ily)s e DF D i i' A L ;
1. ATEn Upscale Thertal Fwr a. Affin bvpa=3 twttch t re ovpass 2. IRh Upscalc n. n o -; e Sw tch in Fun s
3. AFRn Downscale .
I n r1 ovpiis swttch in Dvnis3 4. IF:n Dowr.sc e i o <; .
node Ou t it h in 51attop 5. APAN Inop e. It n 6an h ?, w i t c h on fi o n l e l i nodi L w 1 1 t. , i in iefve1 1. [ 1, r1 ' s o r, scale (88888 ChlE6CI: 1 O f- C O fi i l l! U f. D Oil il[ A I i' AI.E 88***A
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60 P L ArJ T SYSTEMS DESIGN, CON TROL , AND INSTRUMENTATION PAGE 11
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00ESTION 6.09 (3 00)
Asssume the FEEDWATER LEVEL C0rJ TROL SY5 ten ts beinq or>erated'in 3-ELEMENT corit r ol uscir.3 r eactor LEVEL DETECTOR C H A t4 t4 E L 'A*.
Reactor power is at 85%, STEAD 1 STATE.
For each of the t ri s t r o m e ri t or c o rit r o l s t 3rie l failures listed below, STATE HOW REACTOR LEVEL WILL INITIALLr RESPOND (increase, decr ease, or r em a t re s con s tarit ) and BRIEFLt EAPLAIN WHY t re ter ms of WHAT is happening in the Feedwater Control Svstem IhhEDIATEL) AFTER ..
1pE FAILURE. -
(FOR ExAnFLE, voor answer should include tne rollowtna detail,
'Causes error reactor stqnal, steamlevel to decrease due to a steam flos/ feed flow flow ,
feed flow, resulting in a stonal to increase the speed of the reactor feed pumpts),' If APPLICALLE.) ,
a. ;jg' FEE 0 WATER line FLOW stqnal FAILS HIGH. ~"
b. Channel 'A' RE AC TOR LEVEL- de tec t c r s t g ria l FAILS LOW. '""
c. LOGS of CONTROL SIGNAL from
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'O' Feacter Foec Pumrs GOVERNOR.
GUESTION 6.10 (1.50)
Desertbe the condittens necessarv~to cause the followinq alarn.s on the Iriter loc k 5tatus Dtspicy nodule esbottetsd wi tii the r e f ue l t r.3 p l- t f or m.
(use the attached picture for_ reference #
a. '6rtdge Reserse 5 top 4 l'
30.LO)
6. ' Lack Up Holst Limit'
(0.50)
c. ' Rod Dlock Interloc6 i l'
(0.50)
(ms*** END Of CATEGORr Oc saast)
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7. PROCEDURES - NORhAL, ADtJOR h AL , EriERGEtJC) AND P($C E 12 RADIOLOGICAL CONTROL GUESTION 7.01 (1.50)
Concerning the use of Emeroencv Doerattno Procedure flow enarts a. If, while in the process of executing a r. erae r genc y proccoure, another entry condition for that procedure occurs, what must the operator do and why?
(1.00)
D. Whtle in the process of anecutinq an emergency procedure. an a c t i o re step cannot be p e r f o r n.e d . C a r. the oper ator c urettrive in the procecure? ~ '"* '
(0.%0)
GUESTION 7.02 (2.00)
Procecure GP-EO.ZZ-102 (Coor,atna.ent Control) directs the operator to
' r or batt recirc and n.anually scram * If suppression pool t e e.p e r a t u r e
cannot be maintained below 110 F or if an S R 'J naa been stuck onen for 3reater thari 2 a.inutes.
a. Whv is rectre runback prior to scram? **
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(0.50)
b. 110 F was chosen because, acions otho8 a sons , It is ',h e ' Do r or.
Insection T e n,p e r a t u r e . ' Euplain what is meant ov this t e r n. . (1.50)
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(T STIGN 7 sO3 (O.00)
Follcwing a requitec initiation of the SLC system vou ,, r o directed bv the level / power c o ri t r o l procedute (OF-EO. 7-207) to 'loxr RT V w i. t e r level bv t e r n, t i . a t i n q ano preventinq all in metton e.:c e at. r r o n. CnD and borori t r. j e c t i o r. 3, s t e r . urit t i e i t he r . . . '
a.
Wh t are three conditions when you may reinttiate c o o l a ri t t rijec c 1oni ( 1 . '; O s b. Wost is the pornose for lowertra RF V water level at tot; t, t a, e <
t 1.00/
c. Ur Ge r what c or.d t t t oris w o u l e, it be p e r r. t s s i b l e.
to t u r r t ria t r tne etC inseetion once tt has been manuallv or automattcalIv intttated- (O.Su)
<>sa4a CATEGOR) 07 LON1IllulD ON N L ), i FM.E sa4s4)
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7. PROCEDURES - NORMAL, AE.r4uRr1 AL e ENERGENCi AND PAGE 13
-- s35i5E55iCEE C5siE5E---- -------------- --
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CUESTION 7.04 41.50)
According to the Alternate Shutdown Cooling Procedure, (OP-00.7Z-205),
what would De the alternate heat removal flow path 1or per f or ma rig the shutdown c o o l t re q functton?
t1.50)
GUESTION 7.05 (3.00)
Concerntnq the blowdcun and rectreviation modes of the r:e a c t o r water Cleanup L y s t c. n. :
a. Durinq plant startup, the operator ts cautioned to olace the RWCU s y s t e ar t rit o the blowdown mode prior to starting the C is b p u n.p . What is reason for this caution? (0.50)
b. When opersting t re the blowdowre n.o d e , why shouldn't you divert all the RWCU flow to Itquid radwaste or the main condenser) i 1.00)
c' ' . UrtQ t what op e r a t i o ria l conditiores woulo the r ec i r cul a t t ori mode of the RWCU svstem most i t k e l h,b e used and why? '* ( 1.50)
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GUESTION 7.00 (2.00)
__
L Hope Creek ts in the process,.of refuelinq when the crare operator drops e fuel asi,en.biy i n t o -t u storage area. Ac t o r d i r.g to I t r a d a s. t r o Fuel Damate While hefuelinq procedure (Of'-AO.ZZ-101) uhat are t.ie i n. c. e d i a t e oper ator tnd a u t o s. a t i c a c t i o ri s which ni v u t de perforred at verifted) t;.00)
DUELIIJt4 7.07 (1.25e A fire occurs t r. the control r o oth r o sio t r t n q tmmediate control roon.
O v a C O di t 1 O ra .
4. What are the immedtate ictions ortar to control room e v a c o ,, t t o ri ( Of - Ai. . Z : - 15 0 i i i O . t,0 i o. HfCI ano 6CIC auto t rii t t a ted prior to estabitshtoo c a rit r o l at tan s hu t d o w r. pene1. /. t tnr s u t d o w r. panel you tr ar sf ei e11 t r ar.s f e r switches to ' E r1 E R ' . What h6ppens to the operatton of botn HICI in: f.CI C T (0.75s ieeaae CATEG0hr 07 CONTIrJULD Or1 NEAi f'til; C s****>
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7 0 PROCEDURES - NORMAL, ADNORNAL, EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 7.08 (2.00)
A tour group of hich school students are currentiv vistttna Hope Creek during your shift. The gr oup corisists of a 17, 18, a r. d 19 year old, and a 36 vear old instructor. The instructor is an ex-radiation worker, and has completed an NRC-4 form with a total enposure of 88 REh. All of the visitors have completed the GET and RWi courses a r. d as sesch their exposure is not limited due to their 'sisitor' status.
a. For the 4 Individuals, what would be the administative ovarteriv dose limits? (no entensions) (1.00)
b. Li.L would be the NRC mantmum Quarterlv dose limits; (1.00)
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OllE S T I0t4 7.07 (2.25)
During operation at power, a f e edwa t ep %ra l f un c t i on occurs caustno water lovel to increase rapidly. According to OP-AD.ZZ-117 ' Reactor
% 9h Level' procecure:
. .a 93 . At what level must the operator close the MSIV's, termir a te all i r.j ee f ro n into the RPV, erd verafy the reactor has scrammed? (0.25)
b.
Assuminq level exceeds 110 inches, as indicated on the voset ranie,
what Edditional p r,o,p l e n. could exist? What alarms n13ht agurclate as a result of this problemi What addittonal pr ecaut t ons TUs t the cperttor t,a v e as a result of this p r o t l e n. ) . .%., s2.0u)
GUES TIOrJ 7.10 (1.50)
Procedure Of-AL.ZZ-1v6, ' L F h r1/ A f h r1 Failures,' usqqests three methods for d e t e r c. i r, t r. 3 If an LPf:h a; arm is due to high neu t r or, flu, or J r. Lfhn fativre. Erterlv describe these methoos. it.t0)
G U E L T IU ri ~.11 (2.00#
What are all the entry c o n d i t, t o n s for the 6eactor/ Pressure Vessel Cor.t r ol Fracedure, OP-EO. Z-101; (2.00s t***** CATEGORr 07 C0tJ ilrJUED OrJ rJE A T FAGE seaes; l
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7. FROCEDURES - NORNAL, ALNORNAL, ENERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 7.12 (3.00)
Use the attached fioures from Op-FO.77-102 to a n s w .p r the following'
o. Determine the minimum suppression pool lesel 31 v e ri a RPV pressure of 700 psis and suppression pool teaperature of 160 F1 (1.00)
b. Following a LOCA within the drywelle suppressiosi c h e m b t- r pressure is 50 psiq and suppression chamber water level is 17 ft.
What could this be indicative of? (1.00)
c. Explain the basis for the curve OW-F -1, Orvwell Sprav Inst 1ation Limit Pressure ^ Limit. (1.00)
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(***as.CND OF GATCG06/ 07 eaema, l
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8. ADNINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS l' AGE 16 OUESTION 8.01 (1.50)
The reactor is operattnq in mode 2 at 8% reactor power. A check of conditions /surveillances required to enter neode 1 andtestes that orie trip systesi of the EOC-RPT trip instrumentation is inoperable.
Using the attached technical spec 1ftcations, d e t e r ni t rie af the startop i nay continue and the mode switch placed in RutJ . Site specific references to tech specs for your answer. (1.50)
GUESTION G.02 (2 00)
The reactor is operatinq at 100% reactor power and 100% rectreolation t
'
flow with the two rectreulation p u nip speeds within 5% of each other.
During a routine survetltance, it is found that indicated total core flow is 12% greater than established total core flow derived f r o n. tecirculation loop flow. Using the attached technical speciftcations, can the pl. ant continue to operate under these condit tores? Ref er ence any sections of the
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T.S. used in vovr answer. (2.00)
4.
QUESTION 8.03 (2.00) -
Using the attached M ehnical Spectftcations, d e t e r n. t n e the m a :: t mum
)'
time that the recetor rh o y carittnue ope r a t ion ;31 ven the followin3 malfunctions. Reference the sections of tech specs used in determintna ,,
your answer.
a. It is discovered that valse F040A ( A H F- 'A' Heat Euchanaer Ovpassa 15 fsiled open arid c a ritio t be closed.
tl.00) s b. Susequent to the malf' unction in tai aoove, it ts toond that FHh pump E. 15 anoperable.
(1.00)
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8. aDnlNISIRATIVE PROCEDUhES, CONDITIONS, At4D LIMITATIONS PACE 17 OUESTION 0.04 (2.50)
State whether the folle=1nq events would or would not be a one-hour reportable event.
(2.50)
a.
While performinq rounds, it was observed that the HPCI pump mechanical speed overtrip had not been locally reset following system testing.
b. A maintenance worker looses one week of work after recetving a strained back while repairing a feedwater turbine.
c.
A maintenance worker receives 125 REn to his rtant h a re d .
d.
HPCI ramp generator cirevit.
fails to start f ollow1r 3 a reactor s c r .;.n. due to a faulty
e. A bomb 15 found in the reactor building.
QUESTI0t1 3.05 (l.50)
Concer,rans . shif t complement and shift turnovers; c. How many SRO, RO, STA are reovared in oneration condition 17 <!.00)
b.
You come on shift a r.d f i nd that c,rily one (or -c oming) NCO 15 pr eseret.
Under what conditions can voor crew accept shift responsibt11ttes in this condstion?
~ (0.50e GUESTION 6.06 (2.50)
The reactor operator is p e r t o r n. i n q a surm illance of the Stanobv Ltquid Control Systen. and due to systen, m c.d t f t e s t i o n s . c pr ocedur cl e. t e p becomes impossible to perform.
a. Under this condition can an o r, the spot chanac be 1350ed?
(0.50)
b. Wnst four 'vey pot $t's' n.ust be 4"J h e r e d to wher. 1520:o3 n O C C '< (2.00)
<eaaea C A T C G 0 fi r OL CutJ ritJUE0 U" tdL A T IAGE *aaae>
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8. ADnINISTRATIVE PROCEDUREL, CONDITIONGs AND LIh1TATIONS PAGE 10 QUESTION 6.07 (3.00)
Ertefly enplain whv each of the following Technical S p e c a t' t c a t i o n rectreulation systen. LCO's are necessary.
a. Rectreulatton pomp speeds shall be maintained within 5% of each other with core flow greater thari or equal to 707. of rated core flow. (1.00)
6. Two reactor coolant system rectreviation loops shall be in operatton with tot,1 core flow greater thare or equal to 45% of rated core flow. (1.00)
e A ri tdle r ec t r evl a t t ori loop shall rio t ce started ure l e s s the tenperature differential between the reactor pressure vessel steam space coutant and the bottom head di a t rol inc cool t.n t is lest there of equal to 145 degrees F. (1 00)
GUESTI0ti L.00 (2.00)
The Divtsion 1 G$esel is operatina a rid as 30 minutes into a su r ve t t l aric e test where the att starting system fatis. The m a t rit e r.a ric e repair team -
estimates a 2 day minimum repair time. (use the attached tech specs to enplain your areswe r )
a. Is the Otesel Generator anoperable accordtnq to tur,; sueco? (0.50)
6. Are all the D i v i s t ori 1 E C C L s y n t e ni s a noper ab.e t.e c a u s e of the Diesel Generator problem? E.:p l a t n .
it.00s c. If at the ;. a m e tin.e the D 1 v 151 ore 2 core apray pump as out nf <. r r v t e r>
what added impttcattons does thts nave o ri voor tech s r. e c s posittorJ (v.LOJ GUESTION G.09 (l.00)
Of the shirt complemerit e 4 andtviduals c a t.no t be constu r d a portion of the fit e br1 3ade. Idcrit i y the 4 oy tit 1o or o tia r u s ,e . .;.00)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PACE 19
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GUESTION Q.lO (2 25)
a. What is rated thermal power for Hope Creek? (0 25)
b. According to OP-IO.22-006, ' Power Changes Durins Operattori', rated thermal power may be enceeded under certain conditions. What are these co ridi t i on s ? (1 00)
c. According to the same procedure, what checks must be performed following a thermal power change enceedina 15% of rated? (1 00)
GUESTION G.ll (2.00)
TRUE or FALSE!
3. Anytime a motor operated valve that is reqvtrod to change position to fulfill a safety function is manually seated or m a riva l l y bac6seatede that salve will be declared inoperable. (0.50)
b. During a valve alteriment, a locked valve must be verified. The lock should not be removed if the valve is kndth to be in the
- rcqvtreo position.(No locil or remote position indication entsta) (0 50)
c. During a valve alignment, a throttle valve must be vertfied. It should first be fully closed, then opened the rqquared non.bor of tnr ris .
(No local or remote position indicattore outsts) (0.50)
d. If a circuit breaker ttsps due to actuation of a protsettve device, or falls tc close, one additional attsmpt to recicoe toe etrcutt brealer 16 allowed before ari t rispec t i ori o f the circutt trcaler is conducted. (0 50)
DUESTION 0.12 (1.50)
Technical Specification 3.7.2 reqvtres two Control Room Emergen:y Filtration sub s ys t emt. t o b t- operable arid p r c v 16es cup!tc1t ac t 1or, r eq u i r e rrie n t s if one subsystem is t r.op e r a b l e . If coth of the s u d . y s t o c. s were to boccime n r oper able, no specific actiori stetement uuold epply. How should an operator interpret toch Spocs in this t ris tar c 4 i n.1 in other sis.11ar instarices riot dir octly pr ovided for in tne actnon s t a t e merit s to t ri s u r e the intent of the spectrications are mot 9 s1.50)
<sasaa CATEGORY 00 CONiltlUED ON tit y i P M.L seees)
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G. ADMINIG TR ATIVE F R0f.CDURES , CONDITIONS, AND LIMITATIONS PAGE 20
Q U C S T I0r1 0.13 (1.25)
Authort:stion fcr any personnel to receive a radiation dose greater than regulatory limits is considered an Es.ergency Caposure Author t:ation.
4. What are the recomn. ended upper limits for enier gency eapo sur e to save station equipa.ent and to save a life? (0.50)
' _ . What requirements revst be n.et prtor to author 1:tr3 an En.ergency
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Euposureo (0.75)
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3. THE0R( OF NUCLEAR F0WER PLANT OPERATION, FLUIDS, AND PAGE 21
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ArJSWERS -- H0f E CREEK -85/10/15-CRESEENZO, F.
ANSWER 5.01 (2.25)
M 5TE TL a. INCREASES CO.252. Cecause the change in density of water per degr ee F change in temperature increases with i recr ea s t ng temperature C O . dd1. (1 0)
.S b. INCREASES [0.253. Because neutron leakage from the fuel cell to the v o l o rt. e around the control rod increases c::posinj the rod to a higher thermal neutron flu >: [0. 753 . (1.0)
c. CECREASE3 to.253. Eecause the amount of resonance broadening pet degree F change fuel temperature decreases OR at higher fuel temperatures most of the broadening takes place at the higher energies where fewer and fewers to u t r o re s e >:i s t ( 0 Ql .
(Either reason e57' rect for full credit.) 'J (1.01 e
REFERENCE CRF LP 1, pp. 27, 31, 34, & Figure 45.
HCCS, Gtudent Handout, Moderator Temp Coefe See 27, p g .2t Doppler Coef Sec 30, p g 1, 2, a 3, Centrol Rod Worth, Sec 31, pg 5 s'
ANSWGR 5.02 (2.00)
. ~en'
Crtticality 15 achieved low in the power range ti.e. Source or *
t r. t e r r.4edtate range). It is difficult a rid tinc consuming to d i s t i n g u t s h t., e t w e e n oractly critical (Heff = 1.0) and subsctttc31 n.u l t i p l i c a t i o n and/or the effects of source neutrons. [1.03 If 4 posittve pertod t5 miintained wtthout forther por.1tive r e a c t i v a t ', addtttons. it is assured that ersticality has been achieved. The reactor 17, in fact supercritical. C1.0] (2.0)
REFERENCE So aqueMoria Reactor Theorv 3C023 A-4, Rev 0, 01/20/03, pg 14 a 15 HCC Otartup Procedure, OP-IO.Z2003. pg 0 note 5.2.15 r1 cacter Theary, R e a c t i v i t, y . Gec 10, pg te Gubcritteal tiv l t a p l i c a t t an See 21 p3 4 f. 5 v
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22
________________________________________________________
THERN00YNAhICS
______________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
AMSWER 5.03 (2.25)
Assumptions: Lambda = 0.1 see -1 (given)
B e t e. Dar = 0.0070 Alpha T = -1 X 10 E-4 Feriod = 100 see (gtven) (0.75)
T = (1:-p ) / ( . p ) : solve for p: p =,D/1+T.: [0.53 _
p = 0.0070 / 1 + 100 X 0.1: p =
6.3 X 10 E-4 delta K/k [0.253 (0.75)
tJ O T E : The second part will be graded independently of the fit $t part.
dolta T mod = p / alpha T CO.253
=
6.3 X 10 E-4 delta K/K / 1 X 10 E-4 delta K/K des F
= 6.3 deg F CO.253 -- .
Moderator Tem = 160 des F + 6.3 deg F = 166.3 deg F E0.253 (0.75)
REFEREt1CE Susquehanna Reactor Theory SC023.A-4, Rev 0, 01/28/83, pg 3, ,_
SCO23 A-7 Rev 0, 08/31/83, pg 7
,
HCGS Reactor Theory, Period Equation, Sec 24, pg 8 - 10, and Reactivity C o e f f i r. t e n t s ird Defects. Sec 26, pg 3
-
ANCuER 5 04 12.00)
The reactor :s now producing less steam to go to the turbine. T h e r s.
will be less entraction cteam as+d_I " -r - d' r d , going to the feedwater heater.(1.0) Therefore less feedwater heating will occur r e sul t t rig in colder feedwater enter t rig the vessel (.5) which util c wsc reactor p o w.e r to increase about 3% f r o rn the posittve reactivity eddition (clphe n.s.(.5) (2.0)
FEFERENCE rid f i , E i n.v l a t o r rialfunctions Causes and Effects d510 HCGS, Reactor Theory, Moderator Temp C o e f f i c i e ri t , Sec 27, pg 1 a rid Simvlstor 'J a l t d a t e d Nolfunction, FW-12 du
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23
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MJSWER S -- HOPE CREEK -85/10/15-CRESCENZO, F.
.
ANSWER 5.05 (2.00)
a. Transition boiling may occur CO.5] which can result in clad failure [0.53. (1.0)
b. To make the NCPR l i ra i t more conservative [0.53 to account for the possibility of a sudden flow ir crease and a correspondir:3 power increase CO.5]. (1.0)
C'To assure that the Safety Limit MCPP will not be violated *
acceptable for 3/4 credit) -
REFERENCE WNP Systems Training hanual, Vol. 1, Ch 3, and Tech Spec Power '
Distribution Limits Bases.
HCGS, Tech Specs, 53/4.2.3 Minimum Critical Power Ratio, ps C3/4 2-4 & 2-5 Ther.mo s L.P. Critica ower Sec 12 pg 3 E 4 ANSWER 5.06 (3.00)
a. RPT cn tur bine trip [0.5].
~ #
tJ a~*'
t u r a l circulation from decay heat E0.53. (1.0)
'
b. Turbine CPV's fail to open E0.5]. SRV's control pressure at higher value EO.5]. (1.0)
c. Void callapse due to Fressure increase and the scram CO.5].
Level swell from SRV's lifting E0.53 (1.0)
REFERENCE BRF TRANSIENTS. DXY-7 ANSWER 5.07 (1.00)
a. Decreases.
b. Increases.
REFERENCE OFNP Thermodynamics, p. 5.3-2.
of (de General Energy Equation and Heat HCGS, Exchange Thermodynamtes, Appitcation O p et a t i o n , Sec 8 pg 22 &_23
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24
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..--------------------
-_------__----
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
_
ANSWER 5.08 (1.00)
C.
REFERENCE HCGS, Fluid Flow Student Handout, Sec 3, pg 9. and Sec 6, pg 1 ANSWER 5.09 (1.00)
b.
REFERENCE HCGS, Thermo Student Handout, Sec 14, pg 20 ANSWER 5.10 (1.00)
d. .3..
REFERENCE HCCS, Re_ actor Theory Student Handout, Sec 13, pg 3
- apr ANSWER 5.11 (2.00)
a. M Cp T =M Cp T 133,000(1) T = 256,000(1)(60)
T = 115.5 F Tout = Tin - 115.5 F Tout = 260 F - 115.5 F = 144.5 F (0.5 equation, 0.3 calculation, 0.2 answer) (1.00)
6. Yes (0.5), setpoint is 140 F 'C.5) (1.00)
REFERENCE HCGS, Thermo Student Handcut, See a, pg 21 3 22
,
%
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25
- -- -
-- -- . .------.
ANSWERS -- HOPE CREEK -65/10/15-CRESCENZO, F.
ANSWER 5.12 (1.50)
a. No significant effect b. Less rod withdrawal c. Less red withdrawal REFERENCE Brcuns Ferry L.P. 11 HCGS, Reactor theory student handout, Sec 31, pg 2 ANSWER 5.13 (2.00)
a. NAPRAT is the, ratio of APLHGR(act) to MAPLHGR(LCO).
,
(0.75)
b. NO (0.50)
c. The clad temperature can exceed 2200 deg. F during a DDA LOCA. _.
(0.75)
sc - ,
REFED,5NCE SSES Thermal Limits HCGS, Student Handout, Heat Transfer, Sec 16, pg i & 14 ANSWER 5.14 (2.00) --
a. 7 b. 2,[
C. 4 d. 6 REFERENCE HCGS, Student Handout, LHCR Lia.it a r.d Bases, See 15, pg 1, 2, 3, & 6
. . . . .. . - - . . .- . _ . . _ - = _ _ . -.
.
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,
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26
______________________________________________________
ANSWERS -- HOPE CREEM -85/10/15-CRESCENZO, F.
,
e
~
ANSWER 6.01 (3.00)
a. " i f- or : '. : r '!; t s , Lm.el All 76<k'd 7"*P3 (0.50)
b. Draining the CST to the suppression pool. (0.50)
c. The Jockey pump taking a suction from the CST via the HPCI Booster pump suction. or Taros if HPcr o n.wed v p (1,00)
'
d. The HPCI turbine would be operating at its idle speed setting. (1.00)
REFERENCE oA //* CAwddste sfwiryi ixAkasie cMn's.
BFNP Hot License Lesson Plan 42, pp. 18, 19, & Figure 1.
HCGS, L.P. HPCI, 302HC-000.00-026-01,.Pg 28, 33, 34, 58, 70, Table 1 l
- Figure 2.
! '"""
ANSWER 6.02 (3.00)
! 1. The_ turbine.will start to~ increase in speed developing a mismatch between turbine speed and speed select signals.
! 7.~~This (neggAive) signal to the LVG is then passed-to the F
IV regulation subsystem Ethe error should not be large enaggh to cause the IV's to close] and the CV regulation ,
, .-&ubsystem.
t -
j 3. The error signal to the CV regulation is multiplied (by 1.11)
and then~
summed with the output of the load set signal (of 60%), decreasing the signal to the pressure control subsystem.
4. Once the Load Set signal is adjusted below the pressore
'
~ control signal the CV's will start to close. (~9 rpm) !
5. When-the pressure control signal overcomes the CV and the BPV close bias the DPV's will start to open (~11 rpm)
6. Final Steady State Conditions * Rx Pur ~50%, Gen Pwr ~35%,
1811, Ex Press > initial.
,
C6 0 0.5 ea]
REFERENCE Monticello, System Description, B.5.9, Main Steam Pressure Control, pg 9 & Figure 3 HCGS, L.P., EHC Control Logic, 302HC-000.00-051-01, pg 31, 32, 48 - 53, L
and Figure 10
i i
)
~ , . - - - r._ . . . . , _ . - . . . . . , _ , _ . - . _ . _ _ . . _ _ _ _ _ _ _ . _ . _ _ . _ _ . . _ _ . . _ , - _ _ , , . . - . _ _ _ . _ - _ . _ ~ _ _ , . - - _ - - . - , . - , _ _
Y, i . .
.
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 6.03 (3.00)
a. Inboard MSIV - normally operated with nitrogen (Instrument Gas)
Outboard nSIV - supplied by instrument air (1 00)
6. The accumulator volume is adequate to provide full strokin3 of the valve through one-half cycle (closing). (0.50)
c. 120 VAC (g.9 5)
- 22
. 'J : - '
iO.50)
d. 1. Valves 30 shut 2. Valves go shut (0.50)
REFERENCE CNS nain Steam System Description, pg 3-4 HCGS, L.P., hain Steam, 302HC-000.00-046-01, pg 20, 21, 22, AC and DC Elect Dist, 302HC-000.00-071-01, pg 17'and 302HC-000.00-066-01, table 5 N ..
. ,
--
- ANSWER 6.04 T2.00)
a. To prevent loss of inventory to the torus [0.75] thrcogh the
, minimum flow valve E0.253. (1.00)
6. 1. will
,
.. . . - v .f .\.\, ... 1, ,C
. . . .
.
,,
I d, oi N,.. I. - , \ w,ll,Wf 3. will not 4. w i l l -ete-' , 's ~2, -r.bcs : x: l'- (1.00)
REFERENCE HCGS, RHR LP 28-02 pg 24, 32, 21, 29, a 34
--
_ _ _ _ _ _ _ _ .
r
.
.
6. oLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28
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ANSWERS -- HOPE CREEK -85/10/15-CRESCENZU, F.
ANSWER 6.05 m '
a. m= T r e e- slie t. True c. False de kt
~
%. .s w -
<
REFERENCE HCGS. RWn LP, 302HC-000.00-009-01, s. Obj. 7, 106, d, b. Para VI.F.1 and :
ANSWER 6.06 (2.50)
a. Alternate Red Insertion and Recirculation Pump Trip (RPT)
e b. None -
c. Feedwater Runback d. S t ar,gspy. L i qu i d C o n t r o l - *
,
e. 10 minutes, 53 seconds.
REFERENCE **
HCGS E::am bank LP 24-0 question 10 and L.P., 302HC-000.00-024-01, ps 15, 22, 25, E 30
_. w
~&.
a
.
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( "NT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION FACE 29
_________u____________________________________________
ANSWERS -- HOPE CREEL' -85/10/15-CRESCENZO, F.
At4SWER 6.D7 (2.00)
a. intermediate 6. full c. intermediate d. full e. full f. i4h edi+4.e b SPC g. full h. full ,, C8 0 0.25 es]
'?kEFERENCE
--.ar,GS Recire LP (No. 302 HC-000.00-020-01) Obj. 7 & 8, figure 25 & 26
-
ANSWER 6.08 (3.00) ,
1. a. 5. m.
2. b. c.
3. s. c. d. --
f . ([f_ ,
- '
4. c. c. e. C12 0 0 3 ;l es]
M REFERE.4CE HCGS, LF, IRn, 302 HC-000.00-014-01, pg 19, Fig 6, and APEh, 302 HC-000.00-016-01, Table II. Fig S s
,
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION FACE 30
______________________________________________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 6.09 (3.00)
a. Causes reactor level to DECREASE due to the Level Control System having a STEAh FLOW / FEED FLOW ERROR, STEAM FLOW s FEED FLOW CO.5]
resulting in a SIGNAL to DECREASE thq SPEED OF Til REACTOR FEED PUMPS E0.5]. (no Chwge 8 cunM dwte "5S W ) 2) is B lN' de # 4*
b. Causes reactor level to INCREASE due to the Level Control System having a LEVEL ERROR, with N0 compensating FLOW ERROR CO.5] reso; ting in a SIGNAL to INCREASE the SPEED OF THE REACTOR FEED PUNPS E0.5]
c. Reactor level should REhAIN CONSTANT because the ' E: ' FEED PUMP Governor E0.5] will LOCK-UP E0.5].
REFERENCE ,~
BF LP 12, pp. 16-19.
HCGS, Rx Water Lvl Cntr1, 302HC-000.00-059-01, pg 10, 11, & 14 ANSWER 6.10 (1.50)
a. Indicates a concition which prohibits bridge motion towards the - --
-
reactor will be present if a rod is withdrawn and the platfort. is 6.
about to move over the reacter with a load on one of the hoists (0.50)
This lamp lights only if the normal maximum up limit fails and the hoist is stopped by the backup hoist limit (0.50)
c. This occurs when a fuel assembly load is on any, hoist and the ,
bridge is over the reactor 7' s0.50)
REFERENCE ,
G.E. Training material ' Refueling Tools Familiart:stion'
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PACE 31
- ---~~----------~~~------
~~R565ELUU555L E5sTR5L
____________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 7.01 (1.50)
a. The operator must return to the begir ning of the procedure and
. follow all the way through. This is because some of the decision
'S' criteria may have changed.
(1.00)
b. Ye. (0.50)
REFERENCE EOP LP pg. 12 ANSWER 7.02 (2.00)
'
a. To minimize the transient. (0.50)
b. M a >: . temp. at which SLC initiaticn will result in injection of hot shutdown baron weight before the supp. pool reaches the HCTL in a r.
ATWS, i.e. assures shutdown prior to emergency depressurization. (1.50)
REFERENCE LP OP-EO.ZZ-102 pg. 9 _,
ANSWER 7.03 (3.00)
a . '5 . R : power drops below 5%
4. RPV water level reaches TAF ,
3. All 3RV's remain closed and drywell pressure remains below 1.68 psis. (0.5 eacn>
b. Concentrate boron (0.5) and enhance void generation (0.5)
c. When reaching 57. cf the SLC tank volume and only if it didn't . . .
auto trtp. g gclqg go g p p g A,. . (0.5)
REFERENCE OP-EO.ZZ-207 a r. d 101 -
/
/
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7. PROCEDURES - NORMAL, ABNORMAL, EMERCENCY AND PAGE 32
~
~~~~Rh65ULU52E5L EdsYRUL'~~~~~~~~~~~~~~~~~~~~~~~
____________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 7.04 (1.50)
Following flooding of the vessel,(0.25) the flow path would be :
hSIV's to suppression pool via SRV's(0.25)
supp pool to vessel via core spray (0.25)or LPCI(0.25)
Heat is removed from the suppression pool by SP cooling mode of RHR (0.50)
REFERENCE OP-EO.ZZ-205 ANSWER 7.05 (3.00)
a. The CRD pump will increase water level (0.25) and there is no outlet flow g h established (0.25)
_
b. Because cooling flow is lost to the RHY (0.25) increasing the outlet
~
temperature to the NRHX (0.25) possibly causing an isolation of the system.(0.50).
c. Hot shutdown (0.5) with no recirculation pumps operating (0.5)
minimizes thermal stratification of vessel water (0.5)
REFERENCE Prep. for plant startup OP-IC.ZZ-002
.
ANSWER 7.06 (2.00)
a. 1. Suspend refueling ops.
2. Evacuate unnecessary perconnel from the refueling floor 3. Verify R.B. vent isolation 4. Verify FRVS auto starts. (0.50 each,'
REFERENCE OP-AB.Z2-101
.
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PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 33
- ------------------------
~~~~E 5YUL55iEAL C5s?R5t
ANSWERS -- HOPE CREEK -65/10/15-CRESCENZO, F.
ANSWER 7.07 (1.25)
a. 1. Scram the reactor (0.25)
2. Trip the main turbine (0.25)
6. HPCI will continue to operate (0.25) '
" RCIC will trip (0.25) and must be manually restarted from the RSP (0.25)
REFERENCE Control room evac (OP-AB.ZZ-130) pg. 1 Shutdown from Outside the Control Room (OP-IO.ZZ-008) pg. 3 att 1 ANSWER 7.08 (2.00) _
AGE a. ADMIN b. NRC-17 '10 mrem /qtr 125 mrem /qtr 18 300 mrem /qtr g 12s arem/qtt 19 1000 mrem /qte 36 gtS 1.50 mrem /qtr
,,
1000 mrem /qtr 5(n-18)=2 REh/yr or 500 mrem /qtr REFERENCE SA-AP.ZZ-024 pg. 40-43 and 10CFR20 ANSWER 7.09 (2.25) .
a. 90 inches (0.25)
b. The MSL floods beyond 118 inches which will flood the HPCI/RCIC steam lines. HPCI/RCIC turbine trcuble alarms due to flooded steam drain pots. Delay operation of HPCI or RCIC until level drops between levels 2 and 3. (2.00)
REFERENCE OP-AB.ZZ-117
,
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L66565L C60TRUL
____________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 7.10 (1.50)
1. Select a rod which will display the LPRM in the four rod display.
2. Compare with adjacent LPRH's 3. TIP trace near-LPRM (0.5 ea)
-
REFERENCE OP-AB.ZZ-108 pg.2 ANSWER 7.11 (2.00)
1. Scram condition and power >5% or undetermined (0.50)
2. RPU water lever below -38 inches or undetermined (0.50)
3. Rx pressure above 1037 psis (0.50)
4. Drywell pressure above 1.68 psis (0.50)
REFERENCE
'
OP-EO.ZZ-101 ~^
ANSWER 7.12 (3.00)
a. 12 ft. _ _
-
(1.00)
6. Pressure above curve DW/P-2 indicates a bypassing of the pressure suppression function of the containment i.e. failure of the downcomers. (1.00)
c. Spray initiation above this limit may, result in a containment depressurinstion rate that e::ceeos the relief capacity of the d ywell and reactor building vacuum breaker. (1.00)
REFERENCE OP-EO. Z-102 and .a s s o c . LP
. . . _. - -_ .. -. -- - . . _ . .. . .-. -. - - .
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PACE 35
__________________________________________________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCEN20, F.
,
. ANSWER 8.01 (1.50) i
'
!
Per tech spec section 3.3.4 2.d. the inoperable trip system must be
] restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be less than 30% in 6 hrs.
The vmu ma Lion a.v.4 w. . d i
,
pe:'ibi'.: startupcr*-- may ieM+rproceed-r".m.'.t'
- ^ .,a:t- e; ; c- r. d i ' , c -
'
-.r. : LC: 1 -t
,,4.t . _ _ , , . _ , __ 1 1 . . _ . . _ - _ ,
(1 39)
b) W Way
~
h p p li N E. h $ I ~ f . 3 . I .'3 , N .~ )_ . s's Ncn[e I, REFERENCE conbMvc 73o[o fowce j Hope Creek T.S. 3.0.4. and 3.3.4.2.d. FJC 51 i
ANSWER 8.02 (2.00)
,
Yes. As per T.S. 4.4.1.2, operation may contir.ve since an inop jet pump i is verified by TWO indications of T.S. 4.4.1.2 a-c. !f candidate assumes other indication exists then must be in hot shutdown in 12 hrs. ( 2. 00 )
- REFERENCE m_
l Hope Creek T.S. 3.4.1.2 and 4.4.1.2 FJC 121
.
ANSWER 8.03- (2.00)
a. Rastore within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in Hot Shutdown in 12 hrs. ~-
,
T.S. 3.6.2.3 (1.00)
o. Se in at least HOT Shutdown in 12 hrs. T.S. 3.6.2.3 (1.00)
'
REFERENCE T.S. 3.6.2.3
'
..
I t
i
I t
!
l
,
i
. . - , . . , , - . ~ _ - , - , . . , . . . . . - . - - . . . ,-n.,,_,,,,-_<. .-
n. _ .- , ,..,,-_n-- - - - - - . . , . . - , . . .
. .
.
i l
8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36
!
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
't"~ O ANSWER B.04 =(2.',J
--
- i. .
m
,m--
c. n o.
1 50 dd d. yes e. .yes (0.5 each)
REFERENCE Incident Report and Reportable Occurences Program (SA-AP.ZZ-006) Att. I ANSWER 8.05 (1.50)
a. SRO-2, RO-2, STA-1 (if SRO lic. STA, may serve dual role) (1.00)
b. The~off-going NCO must stay on shift. (0.50)
-%.
REFEREMtE Conduct of Operations (OF-AP.ZZ-002), pps. 14, 15. FJC 136 ANSWER 8.06- (2.50)
a. Yes - -
(0.50)
'b. 1. The intent of the ori 3inal procedure cannot be changed.
2. No more than 10 OSC permanent changes can be made to a procedure without revising the procedure.
3. The proposed OSC must be reviewed and approved by the IUpervisor in charge.
4. The SNSS must review the OSC prior to imslementation. (0.50 each)
5. See 75 '.9 3. p *r t' 'c ' -f o .- add.hoa41 revie w s REFERENCE Use of Operations Dept. Procedures (OF-AP.ZZ-102) FJC 137 ANSWER _ 8.07 (3.00)
a. Ensure adequate core flow coastdown from either loop following a LOCA. (1.00)
b. Single loop operation is outside the range of FSAR evaluations (1.00)
c. To preclude excessive thermal stresses of the Rx vessel. (1.00)
I oc ll& f t-l'
,gtre n .#W t 5
.. .tt s iva
.-
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37
__________________________________________________________
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
ANSWER 8.08 (2.00)
a. Yes. (0.5)
No. (0.5), due to ornoshkCt turnig kho b , C- o oo ogo y b. T.S. 3.0.5 (0.5)
c. Must take action of T. S. 3.0.5 (0.5)
REFERENCE Technical Specifications 3.0.5 L ANSWER 8.09 (1.00) '
MSS, STA, 2 other people recuired for the safe coeration or shutdown of the plant. (0.25 ea.)
REFERENCE OP-AP.ZZ-002 FJC 140 ANSWER 8.10 (2.25) ..
a. 3293 MWT ~
(0.25)
b. The average power over any B he period shall not e ::c e e o rated.
no time shall power xceed 102% rated. (1.00)
Check coolant c h e n. i s t r y any hermal lin.its c.(At % i d I b r-(1.00)
REFERENCE c.keD ~
OP-10.ZZ-006 - ~~ -
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ANSWER 3.11 (2.00)
a. TRUE b. FALSE c. TRUE -
d. FALSE Iruc s F eN vqenc i0.5 ea)
REFERENCE OP-AP.ZZ-109
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8. ADnINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38
ANSWERS -- HOPE CREEK -85/10/15-CRESCENZO, F.
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ANSWER 8.1Z (1.50)
T.S. 3.0.3. delineates the measures to be taken for those circumstances not directly provided for in the action statements and whose occurrence would violate the intent of the spectftcation. (1.50)
REFERENCE 1.5. basen 3.0.3.
ANSWER 8.13 (1.25)
a. 75 REh for life, 25 for equipment (0.50)
b. The site emergency plan must be activated, the exposure shall be requested by a department head and reviewed by the radiation protection manage *. (0.75)
REFERENCE SA-AP.ZZ-024 pg. 46
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TEST CROSS REFERENCE PAGE 1 QUESTI0tl VALUE REFERENCE
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05.01 2.25 FJC0000141 05.02 2.00 FJC0000142 05.03 2.25 FJC0000143 05.04 2 00 FJC0000144 05.05 2.00 FJC0000145 05.06 3.00 FJC0000146 05.07 1.00 FJC0000147 05.06 1 Qg FJC0000148 05.09 1.00 FJC0000149 05.10 1.00 FJC0000150 05.11 2.00 FJC0000151 05.12 1.50 FJC0000152 05.13 2.00 FJC0000153 05.14 2.00 FJCOOOO154
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25.00 06.01 'b.00 FJC0000155 06.02 3.00 FJC0000156 06.03 3.00 FJC0000157 06.04 2.00 FJC0000153 06.05 2.00 FJC0000159 06.06 2.50 FJC0000140 06.07 2.00 FJC0000161 06.00 3.00 FJC0000162 --
06.09 3.00 FJC0000163 06.10 ,
1.50 FJC0000180
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25.00 07.01 1.50 FJC0000164 07.02 2.00 FJC0000165 07.03 3.00 FJC0000160 07.04 1.50 FJC0000167 07.05 3.0v rJC0000168 07.06 2.00 FJC000016?
07.07 1.25 FJC0000170 07.03 2.00 FJC0000171 07.09 2.25 FJC0000172 07.10 1.50 FJC0000173 07.11 2.00 FJC0000175 07.12 3.00 FJC0000176
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25.00 05.01 1.50 FJC0000051 08.02 2.00 FJC0000121 08.03 2.00 FJC0000134 08.04 2.50 FJC0000135 08.05 1.50 FJC0000136
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TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE
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08.06 2.50 FJCOOOO137 08.07 3.00 FJC0000138 08.08 2.00 FJC0000139 06.07 1.00 FJC0000140 08.10 2.25 FJC0000174 08.11 2.00 FJC0000177 08.12 1.50 FJC0000178 06.13 1.25 FJC0000179
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25.00
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- PubhC Sefvice Dectric ano cas company Corbin A. McNeill, Jr. Putnc 5ervice E'ectnc and Gas Company P O. Box 236.Hancocks Bndge. NJ 08038 609339 4800 vice Prescem -
Nuciear October 23, 1985 Mr. David Lange Licensing Examiner U.S. Nuclear Regulatory Commission Region 1 631 Park Avenue King of Prussia, PA 19406 .
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Dear Mr. Lange:
Attached are question-by question comments on the Hope Creek Reactor Operator and Senior Reactor Operator written exaninations administered on October 15, 1985 at the Salem Nuclear Training Center. These comments include technical corrections and addi-t tional clarification that appear to haveor -information the most severepertaining impact on ho t ethost overallquestions clarity and correctness of the exam.
The questions are identified by category and all applicable
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references have been indicated. Aside from the specific comments, attached, both the SRO and RO examinations appeared to be a f air representation of -the- technical and operational knowledge required for the Hope Creek SRO and RO candidates.
Also worthy of note is the ' revised examination review process.
Providing the examinations and answer keys at the completion of the written examinations serves to improve the overall quality of the written examinations.
Sincerely,
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Attachment
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Question 2.06 c. The SLC pumps will be sequenced on (power restored) within 30 seconds following a LOCA (as per L.P. 302HC-000.00-023).
The pumps will not start unless an automatic or manual signal is present. Some students may indicate the SLC pumps have been deleted from the sequencer logic (J-105 Rev 6). In either case, power is restored to the SLC pumps following a LOCA.
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Question 5.01 b. Many operators are aware of the rod worth curve from the General Electric SNE manual. Several answers will agree with the key in that control rod worth increases due to e increased thermal diffusion length and 9th increases.
As the reactor continues at power, some operators may j say that worth decreases due to the increased pitch of control rod blades yielding better flux coupling between i cells in the core Rod Worth .LS0' .,
Question 5.06 ;
b. A turbine trip with a subsequent MSIV closure on a six minute trend recorder would look very much like this transient. i Feed pump availability cduld be due to moisture separater ,
steam supply. ~
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Question 5.12 -
b. As per the answer key, the Xenon concentration decreasing adds positive reactivity to the core and therefore less rod withdraw will be required for criticality. However, the negative reactivity due to the Xenon concentration 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> af ter a shutdown will require more rods to achieve criticality than if the core was Xenon free.
Question 6.05 and Question 2.01 c. The select error lamp will illuminate whenever the selected rod is not responsible for the current rod block; the ques-tion (as stated) should be TRUE. However a select error '
will also be generated if a rod not within the currently latched group is selected. Whenever implies "the only time" and the answer to that is false.
RWM L.P. 302HC-000.00-009-01 l
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Question 6.05 d. The question implies that both insert and withdrawal errors exist; therefore the question is true only if the latched group contains a withdrawal error, as withdrawal errors must be rectified before insert errors. If not, the withdrawal errors present (as stated) must be contained within another group, where no red motion would be allowed, because of the insert errors present in the currently latched group.
RWM L.P. 30211C-000.00-009-01 Question 8.04 Questions on 10CFR 50.72 should be treated as you are treating tech specs; provide the reference, then pose questions. General familiarity with one hour notifications is certainly necessary, however total memorization of a document readily available in the control room should not be required or encouraged. When a situation exists that could require one hour notification, the station personnel should consult the appropriate reference material for guidance, not take actions based upon memorization.
This question has parts that would require total memorization of this section of the Code of Federal Regulations (10CFR 50.72).
Specific Comments:
a. This question presents inaccurate information that could 3, have misled the operators while taking the exam. The HPCI mechanical overspeed device does not have to be reset locally, it automatically resets when speed is berbw the trip setpoint.
b'. The answer key has this as a " FALSE" statement however SA-SP.ZZ-006 Part 20.403 would lead one to interpret this as true. The confusTBn occurs over the interpretation of the last part of the last sentence which states "... including a byproduct, source, or special nuclear material." The feedwater turbine could certainly be contaminated from main steam and therefore be a byproduct.
This answer can therefore be either true or false depending upon whether you made the examiners interpretation or the interpretation as described above.
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NRC RESOLUTIONS TO FACILITY COMMENTS ON HOPE CREEK SRO EXAM GIVEN ON 10/15/85 Facility Comment Question 5.01 b. Some operators may say that rod worth decreases due to increased pitch of control rod blades yielding better flux coupling between
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cells.
NRC Resolution Correct discussions of pitch / coupling af fects on control rod worth will be accepted. -
Facility Comment
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Question 5.06 b. The given transient is similar to an MSIV closure and as such ,
discussion of MSIV closure should be correct.
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NRC Resolution Comment accepted. The closure of MSIVs would be nearly identical to the transient given.
Facility Comment Question 5.12 The negative reactivity effects due to xenon concentration 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> af ter shutdown will require more rod withdrawal during startup.
NRC Resolution i
The comment is correct however, the question asked how rod position would be af fected due to the change (trend) of xenon 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> af ter shutdown.
Comment not accepted.
- e Facility Comment Question 6.05 c. The correct' answer is false
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NRC Resolution See NRC resolution to question 2.01.
d. The answer could be true or false NRC Resolution Comment accepted, question deleted from exam.
Facility Comment
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Question 8.04 a, b The situations posed are confusing and ambiguous for the candidates to have correctly answered without more reference material.
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NRC Resolutions
The situations posed are indeed confusing and as such parts a, b will be
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, deleted from the exam. However, condidates should recognize one hour reportable events without consult'ing the procedures.
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ASeAr*reerMj NRC RESPONSES TO FACILITY COMMENTS ON HOPE CREEK R0 EXAM GIVEN 10/15/85
, Facility Comment:
Question 1.05 a. State whether control rod worth increases or decreases as the fuel depletes over core life.
Key: DECREASES Hope Creek lesson plans agree with the key, that is, if rod position is further out (providing less rod density) then rod worth will decrease.
However, if fuel depletes and rod position is maintained constant, rods are in better competition with fuel atoms for thermal neutrons which r.;akes their worth increase.
Neither th . lesson plan nor the Station Nuclear Engineering Manual provide
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a referenc answer for the stated question. We attribute the lack of reference to the complexity of rod worth as it is affected by moderator temp, voids, rod position, adjacent rod position, local and avgge flux, special relationship of fuels and other poisons, etc.
c. How is rod worth affected as rod density decreases (increase or decrease)?
Key: DECREASE 1. The reference referred to by the answer key (Hope Creek Rx Theory 4tudent Handout; pages 27-7,11,13) does not exist.
2. Hope Creek materials (same reference, pages 31-6 and 7) do support the answer key.
3. Both CRW equations [LOCALp),andSLd8 AVE) iP will support the answer key response.
4. Other reactor theory texts state in detail the opposite response (i.e., increase rod worth).
Ref: Limerick Reactor Theory, pages 12-25 and 26 Final Point - Either response should be acceptable if satisfactorily supported by the license candidate.
NRC Resolution Comment rejected; answer key not changed. The reference referred to by the answer key was a typographical error. It should have referred to Chapter 31 instead of Chapter 27.
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Facility Comment:
f Question 1.14 b. Hope Creek lesson plan does not identify a direct relationship between
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head loss and fluid temperature.
Hg = F bY D 2g
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but as temp increases then density decreases.
If m is constant (assumption), then
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m = pay then v increases with a corresponding increase in H '
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If V is constant (assumption), then .
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L NRC Resolution ..
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Comment accepted. Question 1.14b deleted from test. Question 1.14 will now be worth 1.50 points. 'Section 1 af.the exam will be worth 24.5 points. ;
Facility Comment:
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Question 2.01 c. The select error lamp will illuminate whenever the selected rod is not responsible for the current rod block; the question (as stated) should be TRUE. However, a select error will also be generated if a rod not within i the currently latched group is selected. Whenever implies "the only time" and the answer to that is false. RWM L.P. 302HC-000.00-009-01 NRC Resolution Comment accepted. Answer key changed. Answer 2.01c will now be " False".
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Comments 3
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Facility Comment:
Question 2.02 b.2 Only the A and B pumps (operating in shutdown cooling alignment) would trip. If C and/or D pumps are running (i.e., full flow test) they would not trip.
b.4 The outboard head spray isolation valve (HV-F023) will auto-close on high reactor pressure (>82 psig). Reference: Tech Specs Table 3.3.2-7,7 pg 3/4 3-15 (identifies valve group)
3.3.2-2,7 pg 3/4 3-21 (identifies setpoint)
3.3.2-4,7 pg 3/4 3-29 (identifies valves)
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NRC Resolution:
2.52b.2-Answerkeychangedtoaccept"willnot"ascorrectanswer.
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m 2.02. b.4 - Answer key changed to accept "will" as correct answer.
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Facility Comment:
- - * Question 2.03 '*
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3. In this question, the third parameter listed in the answer key is temperature, referencing the alterrex excitation system (L.P. #61-01, page
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10). However, this referer.ce discusses the air cooler associated with the exciter which has no interface with the hydrogen associated with cooling the main generator. If purity and pressure are mentioned, full credit should be given for this question. (Reference L.P. #61-01, page 10)
NRC Resolution:
Comment accepted, but only reference changed. Hydrogen temperature is still a valid answer. More indirect answers were also accepted as correct if they interfaced with main generator cooling. (e.g. TACS flow)
Facility Comment:
Question 2.05
b.3 The answer key says cooling water flow decreases. The purpose of the flow
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' control valve in the CRD System is to maintain cooling water flow at 63 gpm. As the drive water pressure control valve is being closed the flow j control valve will open to maintain 63 gpm. Therefore, the correct answer should be " remain the same" or "no change".
NRC Resolution:
Comment accepted; answer key changed.
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Facility Comment:
Question 2.06 b. The instrument air interface with the SBLCS no longer exists. The bubbler type level instrument was removed. P&ID M-47-1 Rev 7 c. The SLC pumps will be sequenced on (power restored) within 30 seconds following a LOCA (as per L.P. 302HC-000.00-023). The pumps will not start unless an automatic or manual signal is present. Some students may indi-cate the SLC pumps have been deleted from the sequencer logic (J-105 Rev 6). In either case, power is restored to the SLC pumps following a LOCA.
NRC Resolution:
b. Noinment accepted. Full credit will be given for explaining the service air interface alone.
c. Comment accepted; answer key changed to " False".
Facility Comment:
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Question 2.07 ~-
a.. RCIC minimum speed is 2150 rpm vice 2500 rpm as stated in this question.
. Some students may indicate *that there is no problem with turbine speeds below 2500 rpm as .long as the speed remains above 2150 rpm. (Reference E0P OP-EO.ZZ-100 Caution 12 and L.P. #132-00) _
b. RCIC will auto start at -38" subsequent to a high level trip at 54" as per Hope ' Creek Technical Specifications. (Reference Hope Creek Tech ' Specs, page 3/4, 7-12, Rev. 4)
Note: This is a recent change to the system, previously the trip throttle valve would trip, requiring operator action to restart RCIC.
NRC Resolution:
a. Comment accepted.
b. Comment accepted; key changed.
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Facility Comment:
Question 2.11 a. Answer k y states that with off site power available, C&D RHR pumps start
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imraediately and A&B RHR pumps start after 5 seconds. Actually, A&B RHR pumps -start immediately and C&D RHR pumps start after 5 seconds. (Re fe r-ence L.P. #28-02, page 79)
NRC Resolution:
Comment accepted; answer key changed.
s Facility Conrnent:
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Question 3.05
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g. Low lubl~ oil' pressure of 30 psig is a recirc drive motor trip only if the condition exists for six seconds. The absence of the time delay in the i
' question may have caused some examinees to answer differently than the answer key. '
NRC Resolution:
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Comment accepted; question deleted and points redistributed.
Facility Comment:
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Question 3.06 .- - ~ .
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b. To answer the "...and how is it achieved?", part of this question, saying that the RPT breakers trip was sufficient. This has been agreed to by the exam graders.
NRC Resolution:
Comment accepted.
Facility Comment:
Question 3.07 c. Part c of this question asks when the IRM scrams ar1 bypassed. Both the IRM Hi-Hi and IRM IN0P scrams are bypassed under the i;11owing conditions:
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Comments 6
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(1) That IRM channel is bypassed (2) The ruode switch is in "RUN" and the associated APRM is not downscale or bypassed.
(Reference: H.C. Lesson Plan #14, Figure 6)
NRC Resolution:
Comment accepted; answer key changed. Question 3.07 will be worth 3.5 points.
Section 3 of the exam will be worth 25.5 points. . . _
Facility Comment:
Question 3.10
- 3 APRM Downscale Answer "g" need not 56 included as a bypass for this trip. This was agreed to by the examiner.
Note: These comments alst pertain to question 6.08 on the SRO exam.
NRC Resolution:
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Answer "g" deleted from key.
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Question deleted from test. Point valves were redistributed.
Question 3.10-5 will now be 3.10-4.
Facility Comment:
Question 3.11 The answer mentions that the static inverter is " normal seeking" however, that wasn't asked for and should not be required in the answer. This comment was agreed to by the examiner. ~
NRC Resolution:
Comment accepted.
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Comments 7
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Facility Comment:
' Question 4.01a The HCGS training program requires an operaice, when given a procedural step,
, to explain the technical basis behind that step. Question 7.03a requires the operator to memorize and recite a step from the body of the procedure.
Ref: OP-E0.ZZ-207, Lesson Plan Objective #3 c. SBLC can be secured under two conditions:
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1. All rods are inserted to or beyond $2, or 2. 690# of Bcron have been injected into the core.
. If the Boron is injected via the SBLC pumps, then 690# injection weight cannot be verified, therefore we inject the entire contents of the tank.
NRC Resolution:
Comment accepted. Answers such as " Reactor subcritical" and " Low level alarm in SLC tank" will be given full credit.
Facility Comment:
Question 4.08b b. .The Answer key says that RCIC will trip when control is taken from the Remote Shutdown Panel. The reference listed in the key says that RCIC's
.125% overspeed protection is still in effect when remote control is taken.
This means that if the turbine overspeeds, a trip will occur, however, if an overspeed condition does not exist, no trip will occur. The answer key is correct if an overspeed condition occurred but, incorrect if no over-speed condition existed.
Ref: OP-IO.ZZ-008, Rev 8 and AB-130 MRC Resolution:
Comment accepted. Answer key changed to read "RCIC will continue to operate as is".
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Comments 8
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Facility Comment:
Question 4.11 Admin Limits
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17 yr - 10 mr/qtr w/o NRC-4 100 mr/qtr w/ NRC-4 18 yr - 300 mr/qtr w/o NRC-4 1000 mr/qtr w/ NRC-4 19 yr - 100 mr/qtr if exposed at another facility in that qtr and w/o NRC-4 1000 mr/qtr w/ statement of no dose or w/ NRC-4 36 yr - 1000 mr/qtr Above limits based on SA-AP-ZZ-24, Table 4, Rev 1 NRC Limits we -
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17 yr - 125 mr/qtr 18 yr - 1250 mr/qtr *
19 yr - 1250 mr/qtr 36 yr - 5(N-18)=2 rem LIMIT = 1250 mr/qtr
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Can be extended to 2 rem /qtr __
The limits for anci8 yr old, based on 10 CFR 20, paragraph 20.101a,'~ states the max dose to any individual shall not exceed 1.25 rem /qtr.
Par. 20.104, exposure to minors, states that any person less than 18 shall not receive a dose in excess of 10% of 1~.25 mrem or 125 mr/qtr. Does not apply to 18 yr old.
The limit for a 36 yr old is the same as an 18 yr old (per 20.101a) which does not apply a lifetime limit on quarterly doses.
Par 20.10lb does limit individual dose to 5(N-18) but only for the case
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of a dose extension from 1.25 to 3 rem /qtr.
Due to the complexities and unusual conditions of this situation, these limits would be verified with the above reference and memory of operator would not be relied upon.
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NRC Resolution:
Comment accepted. Question changed to delete the 17,18, and 19 year olds and full credit will be given for correct answers concerning the 36 year old.
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NRC Concern:
Question 4.04b asked for followup actions once the auto start function of the HpCI system had been defeated. As for any ECCS system, the correct answer is to:
1. - make frequent checks of initiating parameters, and 2. restore the system to auto / standby rode if possible.
NRC Resolution:
Some students misunderstood the question and gave a list of recovery measures, such as " runback recirc". These measures were given half credit if they were appropriate for recovery after a spurious HPCI initiation at 90% power.
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