ML20214L202
| ML20214L202 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/14/1986 |
| From: | Casto C, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214L200 | List: |
| References | |
| 50-338-OL-86-04, 50-338-OL-86-4, NUDOCS 8612020598 | |
| Download: ML20214L202 (249) | |
Text
{{#Wiki_filter:.- - - - - - - - _ - _ _ _ _ - - - - - 7 4- r6 ENCLOSURE 1 EXAMINATION REPORT 338/0L-86-04 Facility Licensee: Virginia Electric and Power Cc.npany Facility Name: North Anna
' Facility Docket No. : 50-338 Written and oral examinations were administered at North Anna near Mineral, VA. l Chief Examiner: d bbWJh/6 v //- /4'- P S Cpirles'E 9(sto Date Signed l Approved by: , Nd M //- /N- F 4 Date Signed l
dphn F.g snrg Acting Section Chief i Summary: 1 Examinations on June 23-27, 1986 r Oral examinations were administered to 10 candidates; 10 of whom passed. Ten 1 l candidates were administered written examinations 8 of whom passed, Simulator l examinations were administered to 10 candidates 6 of whom passed. -) Based on the results described above, 3 of 3 R0s passed and 2 of 7 SR0s passed. I 8612O20598 PDR 86112D ADOCK 05000338 PDR
T' REPORT DETAILS
- 1. Facility Employees Contacted:
- Donald C. Fellows Senior Inst. (Nuclear)
- Robert O. Enfinger Superintendent Operations
- Larry L. Edmonds Superintendent Nuclear Trng.
*L. Richard Buck Supervisor Power Sta. Ops. Trng.
- Walt Shura Senior Inst. (Nuclear)
- Michael D. Crist Senior Inst (Nuclear)
~* Brian P. O'Brien Coordinator-Nuclear Trng Audit
- Attended Exit Meeting
- 2. Examiners:
- Chuck Casto NRC Examiners Bill Dean NRC Examiners Dave Nelson NRC Examiners
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Richard Buck with a copy of the written examination and answer key for.
review. The comments made by the facility reviewers 'are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.
- a. R0 Exam (SRO exam question numbers in parentheses)
(1) Question 1.03 NRC Resolution: NRC Resolution: Concur. Due to typographical error. (2) Question 1.11 (5.03) NRC Resolution: Do not concur. Facility comment considered equivalent to answer. (3) Question 1.12 NRC Resolution: Do not concur. Facility reference material 86.2 page 7.14 lists two intrinsic sources. The question required candidates to state two intrinsic sources. Reference corrected to reflect 86.2.
. o 2
(4) Question 1.14 (5.01) r NRC Resolution: Concur. Answer key changed to reflect 22 deg. l F. (5) Question 1.18a (5.05a) NRC Resolution: Do not concur. Facility recommendations are l considered to be an amplification of the answer key (part 2); candidates answers will be evaluated on a case-by-case basis. (6) Question 2.01 (1.16) NRC Resolution: Concur. Question lacked specifics about plant conditions. (7) Question 2.02c i NRC Resolution: Concur. Facility needs to change reference material to reflect' actual plant conditions. (8) Question 2.03a&b l NRC Resolution: Concur. Examiner placed [CAF] check at p facility for complete answer. Answer key changed. The reference to " degraded" was used only as a generic description. (9) Question 2.06 NRC Resolution: Do not concur. Reference material provided basis for question. Question was extracted verbatim from facility material. (10) Question 2.09b (6.03b) NRC Resolution: Concur. An error was made in the development of the answer key. (11) Question 2.10b NRC Resolution: Concur. Answer key changed to reflect new setpoint of 185 F. (12) Question 3.10f NRC Resolution: Concur. The reference material used did not reflect this.
. n. 3 (13) ; Question 3.10d (6.11d) NRC Resolution: Concur. This' change in plant status is noted. (14) Question 3.11c (6.12c) NRC desolution: Concur. Answer key modified' to' reflect change. (15) - Question 3.13(6.14) NRC Resolution: Concur. pts were reversed in key. (16) Question 3.19c (6.17c) NRC Resolution: Concur. Facility should update reference material to reflect actual plant conditions. (17) Question 3.20a (6.18a) NRC Resolution: Concur. Examiner placed [CAF] to determine
-answer since reference material did not address this indication.
(18) Question 3.20b (6.18b) NRC Resolution: Do not concur. Facility comment considered an amplification of the answer key. Candidates response will be evaluated on a case-by-case basis. (19) Question 4.08 (7.07) NRC Resolution: Concur. Error in development of the arswer key.
- b. SR0 Exam (1) Question 5.18c NRC Resolution: Concur. The question did not contain enough information to support a correct answer.
Part C deleted. (2) Question 6.15 NRC Resolution: Concur. Facility reference material appears not to be precise in this fact; facility needs to revise said material. l 1 1 1
4 (3) Question 7.27 NRC Resolution: Concur. Answer key modified to reflect change. (4) Question 8.05 - NRC Resolution: Concur. Changed to reflect new limit. (5) Question 8.24a NRC Resolution: Do not concur. Proctor provided guidance to eliminate confusion on this question during examination
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
There were 3 generic weaknesses noted during the oral examination. The areas of below normal performance were:
- a. Control wiring diagrams (Instant SR0s).
- b. Wide range excore nuclear instruments.
- c. SI initiation on low pressurizer level.
The cooperation given to the examiners and the effort to ensure an atmosp-here in the control room conducive to oral examinations was also noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. i l N- - _ _ _ - _ _ _ _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ . - . _ -
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U. S. NUCLEAR REGULATORY COMMISSION l REACTOR OPERATOR LICENSE EXAMINATION l FACILITY: - NORTli ANNA 1&2 l REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 86/06/25 EXAMINER: CASTO, C APPLICANT: INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the ansuer sheets. Points for each question are indicated in parentheses after the question. The pausing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up sin (6) hours after j the examination starts. 2 % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY _ i ___ _ i ___________ ________
- 1. ?RIbCIPLES OF NUCLEAR POWER PLANT OPERATIONr TitERHODYN AFT CS , ,
HEAT TRAN9FER AND FLUID FLOW
'O.00
_'_______ _'S.00 [____ ___________ ________ 2 PLANT DESIGN INCLUDING SAFETY , AND EMERGENCY SYSTEMS _ 1 __ _ i ___________ ________ 3. INSTRUMENTS AND CONTROLS 30.00 0 _---..___ _ jl__0 ___________ ________
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL l CONTROL 120.00 100.00 TOTALS FINAL GRADF ______________ __%
All work done on this e" amination is my owni I have neither given nor received aid. 5PELIU55T~5~52G55IURE~~~~~~~~~~~~~~
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- 1. PAGE 2 '
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l QUESTION 1.01 (1.00) Reactivity is defined as which of the f o l l o w i rig ?
- a. The ratio of the number of neutrons at some point in this generation to the number of neutrons at the same noint in the previous generation.
- b. The fractional change in neutron population per generation.
- c. The factor by which neutron population chenges per geners-tion,
- d. The rate of change of reactor power in neutrons per second.
, OUESTION 1.02 (1.00) l Which of the following actions will INCREASE North Anna's thermodynamic cycle efficiency?
- a. DECREASING power from 100% to .
- b. LOWERING condenser vacuum from 29' to 25".
- c. REMOVING a high pressure FW heater from service,
- d. DECREASING the amount of condensate depression.
QUESTION 1,03 (1.00) Attached Fi3ure i 168 shows a power history and four possible samarion i traces (reactivity vs. time). Select (a, br er d) the correct curve l for displaying the expected saurium transient for the given power l history. l l l l t (***** CATEGORY 01 CONTINUE 0 ON NEXT PAGE *****) l
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fDUESTION l'iO4 (1.00) When coooling--~down;on natural:circulationr the+ procedure cautions the. foperator'not-to depressurize the' plant-.below 400 psi 3~before-the entire RCSfis.below 200 deg. F. .The-reason.for this is.to (choosecone)-
- a'. ... reduce [the combined thermal na'd pressure stressesion the reactor
-vessel..
~ L b'.. . . .' p c e v e n t i v o i d ' f o r m a t i o n ; i n t h e . r e a c't o r _v e s s e l .-
3c.;...be.within the design transients specified in section 5 Tech Specs.- Ed.:. . . . prevent. ther' eactori vessel f rom entering a . condition susceptible. to
-br'ittle - f r ac tur e .
~ 0UESTION 1.05 -(1.00)
.A; reactor _has beenloperatins at full power for three months-when'a manual 1 reactor-trip. occurs. All systems are operational and the steam dumps are
-immediately placed in.the. steam pressure control mode. Ten minutes after
=the;teactor' trips all-RCPs are tripped. Twenty minutes after the reactor
>triPr Loop-1 RCP.is~jossed momentarily. Which set of traces 1(a - d) on figure 41174 most~ closely represents the Previously described events?
' QUESTION li O6'- (1.'00)
Which of the following statements concerning Xenon-135 production and
. removal.is correct?
- a. L A't; full' power r equil-ibrium conditions, about half of the Menon is produced by iodine decay and the other half is produced as a direct fission product.
b.- -Followins a reactor trip from equilibrium conditions, xenon peaks because delayed neutron precursors continue to decay to xenon while neutron absorption (burnout) has ceased.
- c. Xenon production and removal increases linearly as power level increasesi i.e., the vtlue of 100% equilibrium xenon is twice that of 50% equilibrium xenon.
- d. At low power levels, xenon decay is the major removal method. At,high power levels, burnout is the' major removal method.
(xxxx* CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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~ ~~~~ TUER566YU555C5~~5E5T TR5O5EER'5UU~EEU5U~EL5E , QUESTION 1.07 -( 1. 5 0 )
Refer to figure # 173 and answer the following questions:
- a. The system flow when pumps 1 and 2 are running is GREATER THAN, LESS THAM, ESSENTIALLY THE SAME (choose one) as the system flow when only pump 2 is running. (0.75)
- b. Assume that pump 3 is running by itself at the required rpm for the given flow. What would the pump DISCHARGE PRESSURE be if the pump speed was increased 30%? (0.75)
QUESTION 1.08 (2.00) If steam-goes through a throttling process, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.
- a. Enthalpy (0.5) ,
- b. Pressure (0.5)
- c. Entropy (0.5)
- d. Temperature (0 5)
QUESTION 1.0? (2.00) TWO. major factors affect differential boron worth over core life.-. List these TWO factors AND irdicate how (MORE NEGATIVE or LESS NEGATIVE) they affect differential boron worth. QUESTION 1.10 ( .50) TRUE or FALSE? For similar heat exchangers operating under the same inlet temperatures and flow rates, a counterflow heat exchanger will transfer more heat than a parallel flow heat exchanger. (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) s
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'~~~~ TREER657sERICs? REET TRANSFER As5 FLU 50 FLUU
-QUESTION'si.11 (1.00)--
LHhichytwo safety limits necessitate operating within-the' control band on pr e s sur iz e r -- p r e s su r e ?~
'00ESTION- 1.12 (1.00):
E lYour. reactor.has'been shutdown for 36 hours. Which-intrinsic
. neutron ~ source is the'MAJORfCONTRIBUTOR-to the1 background-
. neutron-population.
LQUESTION 1.13 (1 00)- Arrange the following boiling phases' associated with nucleate boil'ing and departure from nucleate boiling in.the order in which they would
~ occur in.a channel _with normal flow and high heat: flux. .
- 1) Transition ~ Boiling
- 2) Bulk Boiling-
- 3) Film Boiling
- 4) Sub-cooled (local) Boiling
' QUESTION 1.14- (2.00)
'The' Core. Cooling Monitor has determined.the following readings result,in.
the most-conservative margin to_saturat. ion: N/R pressure PT-1444 = 2235-psis Incore Thermocouple (T/C) = 630 des.F
~a. Calculate the margin to saturation.
'b. Assume the T/C reference-junction box temperature indication has failed low.(zero)-and ac val box temperature is 170 des.F. Explain the effect of-this failure on the resultant nargin to saturation. Address both subcooled and superheated conditions.
Note: Mointor assumes 160 des.F reference temperature upon failure, answer without regard to part a. above. ! (***** CATEGORY 01 CONTINUED ON NEXT PAGE xxx**) i 4
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~~ITU5REbbEUd5EC5,~5E5T N 5RhU5fER 5 6 5LU56~ELUU 100ESTION 1 15 - ( 2. 50 )'
Fdrf_each of[the:following; sets:of conditions' EXPLAIN ~~which one woulo Febolt
~
-in theigreatestfrea,ctivit'y change.duelto~ control rod insert' ion.
Note: Assume 100% power, Bank O at 220.-steps,s BOL.
- 10. An1 area of high' relative. flux : vs. : low' relative flux.
- b. Edge.of-the' core 4vs.~ middle of the core..
- c. -
Rod 91 (inserted) vs.Lrod'52 inserted beside rod ti. QUESTION 1.16- (2.50)- For each:of the following STATE.whether containment partial pressure will.
; increase or decrease and EXPLAIN'the. factors.which cause this change.
a '. Containment. AIR volume increases. b, Draw a vacuum without air recirculation fans then-turn fans on.-
- c. A Mechanical 1 Chiller. trips.
- .00ESTION 1.17 (1.50)
List.~two dangers'AND Explain the consequences of Boric Acid precipitation post-accident.- Assume no. operator actions are taken'to mitigate boron c o n c e n t r a t'i o n . -
'0UESTION- 1 18 (1.50)
In response to ES-0.3 Natural Circulation Cooldown, it is directed that the RCS be borated to Cold Shutdown concentration prior to RCS cooldown. This
- action results in an over-boration of the ACTIVE (core +1oop) portions of lthe system, considering this answer the following questions:
a . _ - Why is this action necessary during Natural Circulation. Conditions?(.75)
'b. Why is-this action necessary prior to RCS depressurization? (.75)
(***** CATEGORY 01 C0t4TINUED ON NCXT P AGE ***** ) a
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7 :
- .-_______________________________________________ t THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ,
4 GijESTION 11.19 (1.00) ! 4 i Briefly EXPLAIN why increasing the baron concentration at low temperatures ' has little effect on a negative Moderator Temperature Coefficient ~(MTC) [ as compared to higher operating temperatures. l GUESTION 1.20 (1.00) 1 Consider'the equation below, answer the following questions:
- I n N= neutron count rate I S (1-K ) S= source count rate N= K= Keff i
- 1-K n= number of generations n
- a. How does 'N' respond to 'M '
approaching zero?
- b. Which term (s) determine (s) the total neutron production rate?
- . i 4
- QUESTION 1.21 (1.00) i l i The fission process in a commercial reactor requires the neutrons that are ' born' by fission to be "thermalized'. Which molecular ,
interaction in the reactor core is the most efficient in thermalizing neutrons? I OUESTION 1.22 (1.00) Explain how adding latent heat to liquid water at saturated conditions will ! affect its state. QUESTION 1.23 (1.00) Given that a battery capacity is 1650 ampere-hours, EXPLAIN the term
' ampere-hour".
t (***** END OF CATEGORY 01 *****) i w,-,.,.--
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 j- GUESTION 2.01 (1.00) e pg 3 i
t During RHR system operation discharging to the Haa-cr3s, MOVs 1720AaB Cold
- Les Injection valves are closed. State the purpose of having these valves j closed and de-energized during this mode of RHR operation.
- l. i QUESTION 2.02 (2.50)
I During Charging Pump transfer operations the following alignment e :< i s t h 1. Charging Pump 1-CH-P-1A is running in Automatic. ! 2. Charging Pump 1-CH-P-1B is in the Pull-TO-Lock position.
- 3. Charging Pump 1-Ch-P-1C is in the " connect' position from 15J7 l l Bus 1J and in the Automatic position.
i
- a. If Letdown is in service, what effect would this have (if any) on the Letdown flow path. (0,5)
- b. Should Charging Pump 1-C H-P-- 1 A trip what action (s) would occur (address j Letdown flow path and auto pump starts if any). (1.0)
- c. Is it possible, with the above alignments present, to rack-in 1-CH-P-1C l 15H7 (H-bus) Charging Pump feeder breaker? Why or Why not? (1.0)
! OUESTION 2.03 (2.50) l Answer the following with regard to the Emergency Electrical Distribution System. f a. What in the purpose of the " Stub Bus' on the Emergency Busces 1H and i 1J? I
- b. State the two signals which will open the " Stub Bus' breaker, s
- c. The "H" Train EDG is loaded onto the H-bus in response to an under-voltage condition. The operator attempts to transfer EDG control from the Control Room to the Diesel room.
EXPLAIN the result of this action upon the EDG controls. t P r (***** CATEGORY 02 CONTINUED ON NEXT PAGE *x***) I
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- 22. PLANTLDESIGN INCLUDING; SAFETY AND EMERGENCY' SYSTEMS PAGE- 9' GUESTION- 2.04- - (3.00 )-
.a. After a valid.SI. initiation signal has been generated which HHSI. pumps =
2
]
.should beLoperating? (assume all. pumps' power supply from normal I source) , (1.0) J, q
- b. State at least two' bases for " locking-out' HHSI pump (s)' during art SI' [
' initiation signal. 12.0) 'L i
00ESTION'.2.05- (3.'00)' l During a LOCA the following evolution has.taken. place with' regard to the ! LHSI-pumps. .
~
l T1 = A SI Recire Mode signal is present - T2 = The SI signal.is reset j i T3 = RHST low-low level. signal is received (4/4) T4 = The LHSI recire isolation MOVs shut ! T5 = (present time) f
'a. What.tuo signals'cause the action at T4 AND what is the. basis for this t action? . (2.0) j
~
- b. At T5 where are the~LHSI pumps discharging to AND from'where are they j receiving their suction? (i.0)
}
QUESTION! 2 06 (1.00) ; l Local surveys in the Sample Room (Aux.- Blds) have shown radiation levels j well in excess o# the setpoint for the Local Radiation Monitor, however, : no alarm has been received in the Main Control Room. What component would the operator expect as the cause for this malf ur;etion? j i I 00ESTION ?.07 (2.00) !
- a. Which component (s) of the Control Rod Drive Mechanism act as the pres- l sure boundary between the RCS and the Containment atmosphere? (0.5) !
- b. When paralleling the output of the two Rod Drive MG sets automaticallyr f the " speed matcher
- automatically changes the speed on which of the- i two MG sets? (incoming / running) (0.5)
{
- c. Explain why the stationary coils for the CRDM are supplied with TWO l DC voltages. (1.0)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) I l, r i I i l 1
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, f 2 i :- PLANTLDESIGN INCLUDING SAFETY?AND EMERGENCY SYST5MS' PAGE 10
'GUESTION 2.00 (
_.1.00)
~
Answer the following with respect to the? Containment System!- ai Post-LOCA-with t'he Hydrogen Recombiners in service the Hydrogen concentration is. increasing,fthe' Hydrogen Recombiner Catalyst'
;- temp'erature.should be _______(increasing / decreasing / remaining
'(O'.5) the-same).
. b..What^signalc will close the-containment vacuum pump. discharge valves-(TVs)? (0.5)
-00ESTION- 2.09: (2.50)
Refer-to figures #.200.1.a 200.2 attached, and answer the following!
~,
a For the full-range indication, describe-why its reading:is inval.id.AND i describe any other abnormal conditions which exist under the current conditions. (1.0)
- b. Figure 4 200.2 indicates several parameters, State _which RVILS range (s).
is/are affected by this' current display. (0.5)
- c. Explain.how the Reactor Vessel. Level Indication System is1 compensated
-to maintain required accuracy during LOCA conditions. (Address 3
-parameters and one physical design feature) E1.03
} 00ESTION 2 10 (1.00)
- a. How is the. reduced pressure operation modification in Unit 2 placed inte service? (0.5)
. b. At what Reactor Coolant System temperature will this system auto-matica11y. reduce its setpoint? (0.5) 1 } i + 1 (***** CATECORY 02 CONTINUED ON NEXT PAGE *****) I' l J ) i k l I i 4 rw qe-~* + - ~= +-4 nn a p w - n-s =n r we-
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5 I i 2 .- spLANT DESIGN: INCLUDING SAFETY AND-EMERGENCY ~ SYSTEMS ~ :PAGE 11 i _______________________________________________________ i a p 'GUESTION 2.11 (2.00) ,
. t Answer the followingLin regard to Emergency Diesel Generator operation: 1 1
.a. Explain the effect (if;any)'of an air VENT' solenoid valve remaining open .l 3 dutingtan.EDG start sequence. ;
- b. Explain the'effect (if any.) of_an air. START solenoid valve failing to- j open during an EDG start sequence. 1
- c. After.a normal stop signal is reset, before-the'60 see drop-out-time'has-
; expired, EXPLAIN.how the EDG would respond to EMERGENCY and NORMAL start i signals. ;
t t
. QUESTION 2.12 . ( 1. 00 )- [
a Can1the outside recirculation spray pump discharge be aligned to the-
'LHSI pump _ discharge?' Choose the ctrrect answer from below.
'a . Yes, for Unit 1 only.
b.- Yes, for Unit-2 only. .I
- c. Yes, for'both Units.
f
- d. No, both Units are aligned at the suction of the LHSI pump. !
t i.
.0UESTION -2.13 (1.00) [
'Which of-the following describer the Service Water System automatic actions on a single unit SI signal? j
- a. All SW pumps start, that unit's spray. header isolation MOVs receive 'f an open signal, b., Only the SW pumps supplied from that unit's emergency buses.et' art,.that unit's spray header isolation MOVs receive an 'open' signal. l
{- l c. . 'All.SW pumps start, all spray header isolation MOVs receive an 'open* ! signal'. y r d.- .The SW pumps supplied from that-unit's emergency buses start, -all spray ; i header isolation MOVs receive an "opon' signal. ' . I (***** CATEGORY 02 CONTINUED'ON NEXT PAGE *****) !
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E24-7 : PLANT DEGIGN-INCLUDING SAF.ETY AND EMERGENCY SYSTEMS PAG'E. 12-QUESTION 2.14- ~(1.'00)-
' ShichL of' the ffollowins . signals t ' actins independentlyr wil13 automatically LCLOSE the Main Feedwater;Resulatins Valves (MOV-FW-1154A, B'and C).?
~
ai H'i-Hi S/G' level in any-S/G (2'outlof-3 detactorn)- b.. .. Low Tavs.(2 out of 3)
- c. Reactor: Trip'
'd . - Phase B Isolation
'00ESTION 2.15 :(1.0'0)
/Which.ofz the.followins correctly describes the distribution of J-tubes around..the Feedwater Distribution Rins in the S/Gs?-
a.: They are evenly distributed.
- b. There'is no_ pre-determined distribution pattern.
.c. More J. tubes ar,e-located at the outlet chamber side of the downcomer thanLthe inlet chamber side,
.d. More J-tubesLare located et the inlet chamber. side of'the downcomer than the outlet chamber: side.,
QUESTIONL 2.16 . ( 1. 00 ) . Which of the followins is a load using Service Water as a BACKUP source oof heat' removal?
- a. Service. Air Compressors
- b. Instrument Air Compressors
- c. -Control Room Air Co'nditionins Units
- d. Containment Air Recire Coolers
(*****-CATEGORY 02 CONTINUED ON NEXT PAGE *****)
UNITED STATES ss ** G49 ,
' p,, NUCLEAR RE!ULATORY CCMMISSION E( o REGION 11 5 $ 101 MARIETT A STREET, N.W., SulTE 2000
- I o ATLANTA, GEORGIA 30323 s,...../
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,t 2.s . PLANT: DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13-
" QUESTION 2.17 (l'.00).. j Which one'of the'following describes the purpose of~the time delay l (195 see) in starting the Inside: Recirculation Spray pumps,on a'CDA-signal?' ,
i a.-Prevents-overloading of'the~ emergency ~'b u ses as loads are sequentially i
, ener3 zed durins. accident-conditions.
i : b.' Allows time for the containment sumps to; collect sufficient fluidEto :l prevent depleting sump level, avoiding a' loss of' makeup inventory. [ c. Enhances core cooling by increasing reflood rate after a LOCA-as the [- pressure drop-between core exit and the' break ~is" reduced with a-higher i . containment pressur~e. 1 I d.' Allows time for t'he fluid collecting in the. containment sumps to cool, ~ [ to avoid flashins.in-the RS. heat exchangers as the fluid is cooled by i Service Water. f GUESTION 2.18 (1.50)I Indicate whether the follow'ing statements regarding RCP seals are TRUE i or FALSE. a) .The floating. ting seair located between the pump radial bearing and the s .tl seal, will limit leakage to 50 apm if the ti seal fails. (.5) I i - b) t3 neal is designed to withstand full RCS pressure.. (.5) ; r j c) Seal water injection from CVCS enters the RCP between the seal package 1 and the pump radial. bearing. (.5) L QUESTION 2.19 (1.00) ' 1: Name the TWO. gas systems.that are part of the fire protection system. l I? ! I t I i (***rr END OF CATEGORY 02 *****) ! l i i l i I m_____. _ . . - . . - . . . . . . . . . . -
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- 3. . INSTRUMENTS [AND1 CONTROLS
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-QUES 1 ION 3.01 -(1.00) -
Which statement"belowEregarding the pressurizer pressure control"and
- protectiv~e. system is 'NOT' correct?- ,
C -
~
a'. :of'the
--The master . pressure : controller p"rovides' p0RVs.o
~
the cor,4ro1 signal fort.only one .
- c, s 1
p
.b. .There is s' lead /las compensation. circuit for; pressure inputs-toQhe low ~ ~
pressure-reactor trip that varies the. trip.setroint with the rate of
. pressure decrease.. + 1,
~
- c. The two pressurizer spray valves are controlled by separatei trar{nmitters id. To block SI actuation ~on a normal' plant depressurizatione;the'op'erator.
must operate :THO block switches to prevent inadvertant ECCS actuation.
,,, S.
; QUESTION 3.02 .'(1.00).
Which of.the following statements concerning the operation of: the letdown isolation valves (LCV-1460A and B) is correct?
- a. Neither of the letdown isola' tion valves can be opened if the it(
containment letdown isolation valve (TV 1204) is:shuto ,' ,'
. b '. - Shutting all orifice isolation valves will automatically, shut' ~
4 the' letdown isolation valves ",'
, 3 ,
- c. All orifice isolation valves must be shut in ur>Jer: to open the '
~ %,,,
letdown isolation valves
- d. High temperature on the letdown outlet of the regenerative!hept exchanger will aut omatically shut the letdown isolation "219et : <
s- . (***** C ATECORY 03 CONTINUED ON NEXT.
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UNITE 3 STATES
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. 3ol INSTRUMENTS AND : CONTROLS :PAGE 15'
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QUESTION R3 . 0 3 ' (1.00) Which oneauf the following malfunctio,ns',could cause one of'the.over
' temperature delta Titrip bistables to5 trip? ,
.? O ,
- a. Controlling turbine impulse- rf ressur e> ghannel f ailing . low.
- b. Power range N43 lower detector fdi'1Ehs lowE i
- c. Reactor coolant: flow detector ?dil'ina low.t e d '. Con.tio? ling pressurizer level channel failing low.
10UESTION- 3.04- L(1.00) With the reactor at 100%. power and the steam dump control system in the.
.Tavs moder'a115% step loss-of load occurs. Assuming no reactor trip. occurs the condenser is availabler and th,e, reactor operator manually OPERATES the control rods,~which of the following would occur if Bank 1 steam dump valves-failed to open?
- a. Bank 2fwould open.
- b. Atmospheric dumps would open.
- c. S/G safeties would open.
d..No other steam valves would open. QUESTION 3.05 (1.50) Indicate whether the-OP Delta T trip setpoint will increaser. decrease or. remain the same for-the.following-parameter' changes. Consider each
-seperately.and addressfonlygits affect on 0P Delta T.
- a. Increasing.Tavs.
- b. Tava < Rated-Thermal Power Tavs. ,
- c. Pressurizer pressure decreasing.
(x**** CATEGORY 03 CONTINUED ON NEXT PAGE *****) l
p sk""'009 UNITED STATES ' '
+j h, NUCLEAR REIOLATORY COMMISSION REo.oN ii
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3..- INSTRUMENTS AND CONTROLS PAGE 16 QUESTION 3.06L (3.00) Indicate what happens to the RodL Control System- (rods .in, rods out, no change) and DRIEFLY explain'why the change will or will not occur for the following conditions. Rods are in auto unless otherwise specified..
- a. Reactor power is 17%~when the controlling turbine first stage impulse pressure-transmitter fails high.
b.' Reactor' power is-100% abd loop 1 Thot fails high.
~
- c. Auctioneered high nuclearfpower is 50%, Rod Control'is in manual.
p ' Instrument testing is in-progress on the turbine-power-input to_ rod control which has turbine power at 100%. All indications have been stable for the last hour. The Bank Selector switch is then'placed in AUTO.
'0UESTION 3.07' (1 50) ;
' Indicate whether.the following statements.a'pply-to Fire Pump 1-FP-P-1, Fire Pump ~1-FP-P-2 or to DOTH. i a)- 2500-GPH, diesel driven pump:
b) Recirculates its^ discharge to,the Unit 1 A-well c) -The pump-can-only-be stopped locally-GUESTION 3.08 (2.00) For-the following protection / control circui.ts, indicate'whether the-list'ed
- permissive function occurs ~as a result of MANUAL or AUTOMATIC action.
; a. Bypassing-the Power Range
- Neutron Flux-Low" setpoint above P-10.
~b. Reenergizing the Source Range instrument below the P-6 setpoint, c._ Bypassing S/G 'lo-lo Water Level Trip
- when RCS Loop Stop Valves are closed.
- d. Blocking High-Steam Flow SI when Tavs < 543 des F (P-12).
4 (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) 1
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PAGE 171 3.~__ INSTRUMENTS AND CONTROLS:. QUESTION 3.09 c(1.50)
-Answer.the following. questions regarding the Excore.NIS TRUE or FALSC.
a) An-Intermediate-Range 1 detector.that is overcompensated could. result in the P-6.setpoint.never beingz reached on'a shutdown, b ). . Placing-the.So'urce Ran'ge instrument in ' Bypass at the instrument drawer.will prevent a-reactor _ trip signal if either the Instrument or _ the Control Power Fuses.are' pulled.
.c) .Neither the: Source Range or the Intermediate Range "Hi Flux" trips are
.taken' credit for:,in accident analysis.
QUESTION 3.10 (3.00)
. Indicate'whether the following valves receive a SAFETY INJECTION, PHASE t
A;or PHASE B input signal (s)!
-a'.-Main feedwater isolation valves.
b.-Chargin3 Pump suction from VCT.
- c. Borie: Acid tank pump to BIT valve.
- d. Charsing pump to-recire stop valve.
e.; Accumulator test line.
'f.' Steam Generator blowdown (TV8D 200C).
'00ESTION 3.11 (2.00)
.-For. each' condition EXPLAIN wh ~
i ch component ( s) would be senerating a red
' movement signal'and the response of Bank D rods to this signal.
Assume..no other. Rod Stop signals present, Reactor at power. , BANK SELECTOR SW. IN-0UT-HOLD LEVER -PLANT PARAMETER 'D' POSITION
- a. Manual- In RIL LO-LO ALARM 100 steps.
b.- ~ 'D' Hold Tavs-Tref +4 deg.- 180 steps
-c. 'Hanual Out '
.. Urgent Failure' 200 steps
- d. . Auto Hold Tave-Tref -4 des. 2221 steps g (***** C ATEGORY O'3 CONTINUED ON NEXT PAGE *****)
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n#3"209,0, UNITED STATES ' ' s , NUCLEAR RESULATORY COM:tSSION f,'
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PAGE: 18 e
'0UESTION~ ' 3 .' 12 '( 1. 00 ) -
Reference.figur.e #230 Core'Coolins Monitor. :After depressing the PSI
-MARGIN pushbutton.the soperatortreads.the. display and 1 finds that it reads 25' des F. In relation,to' saturation. conditions of the RCS what condition sdoes this' indicate?
I GUESTION 3,13 (1.00) 4 + . 4 I lThe f oPer6 tor' is'preparins. to place the.RHR system in service apon RCS l
- ~ cooldown.- .PT-402_has failed hish. DESCRIBE the',effect(s) of this j
. condition onsthe~ operation ~of MOV's 1700 and 1701 RHR suction valves. l l
-GUESTION .3.14 (1.00)
- J .
A pressurizer level transmitter has a Differential Pressure Cell which
'has ruptured.- -EXPLAIN how the level INDICATION in the Control Room would
. respond.
i' i
-QUESTION 3s15 (1.00) ,
. . 1 Followinsfa Guench Spray system actuation, what . two conditions - must' be - l met in order to shut the suction valve for a-Guench Sprev pump?
l
! QUESTION 3.16 (1.00)
-I
- -The.S/G 1evel bypass control valves will automatically operate through I their-full ranse.from zero to _____% reactor power. The controller is also
, capable of anticipating feed flow changes by a signal from ______. _(fill-in-the blanks). QUESTION 3.17 (1.00) What input signal is used by the Rod Insertion Limit calculator as a direct in'dication of reactor power? I I (***** CATEGORY 03 CONTINUED ON NEXi' PAGE *****) 4 i _ .. _ . . _ . . . . - _ _ _ ~ - , . . . , , . _ , _ . _ _ _ - . . . _ , _ . _ . . . . , . . . . - . _ , . _ . _ . . . . ~ _ . , _ , . _ - , . _ _ _ . . _ , - _. [
UNITEO STATES
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L3. INSTRUMENTS AND CONTROLS 'PAGE. 19
.0UESTION, 3 18 -(2.00).
La. .The'Excore.Honitoring' System Channel 1 has redundant power supplies.- From;where.does Channel'.1 receive its power supply 1 NORMAL and-Alternate?:
;b. How AND'where is' transfer to ALTERNATE (part a above). accomplished?.
--0UESTION 3.19 (1.50)
Answer the.following for the CVCS process radiation monitor RM-CH-120/129.
- a. How is the. monitor protected from high CVCS temperatures? (include-
. control setpoints if any)
~
- b. Explain how low fl'ow~is sensed lin the flow; path AND what-indications of, low flow are provided?
.c. What' signal enables the~high range monitor?
QUElSTION 13.20. (2 00)
- a. .AFW pumpL3A.has-three. indicating lights above the control switch.- What' does the AMBER light indicate? '
b.:For-properLAFW' system operation MOV 100Bjand D.and HCV.100C are'lef.t' opened. Explain why_this' alignment is necessary'for-proper, system
-operation.:
i (***** END OF CATEGORY 03 *****) r I I
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~#fy >* E8%g'g# NUCLEAR RESULATORY COMMISSION
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- PAGE 20'
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5 QUEST' ION' 4 .' 01 ' (1.00)
"When p1'ading the- RCS $in a solid watdr condition' per :0P-3.4, the.
. criterion'used to verify th'e'RCS_is_ solid is' -
- a. a high pressurezspike"occuringjon the pressurizer pressure v instrumentsi ,
- b. when spraying ttv.~p essurizer'no longer will.dect.easefRCS.
pressure.
- c. when' indication of' flow to the.PRT through the pressurizer,
'PORVs is verified.
- d.'when an increase inJ1etdown. flow has-been verified.'
cQUESTIONL 4.02: (1.00)- A. hydrogen' bubble formed in the reactor vessel is eliminated by-
'a..ine'reasing pressurizer t'emperature above core thermocouple readings. -
i
- b. injecting oxygen 1into_the reactor coolant system via the cherrical and volume contr01 system.
c'.' maximizing 1 coolant flow by running all rea'ctor coolant pumps, _ ; increasing letdown flow to 120 spm, and placing the c'ation ! Ebed demineralizer in service in parrallel'with the mixed bed dcmineralizer.
- d. venting the_ reactor vessel head.
- 00ESTION 4.03 (l'.00)
Which'of the'following indications require SI reinitiation
.following a spurious SI that has been secured?
- a. .PCS pressure at 1950 psig.
b.1RCSLsubcooling at 35 degrees:F.
- c. Pressurizer level at 17%.
- d. All-steam' generator levels at 15%.
(*****LCATEGORY 04 CONTINUED ON NEXT PAGE *****) 1: d 4 1
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NUCLEAR RE20LATORY COMMISS12N E^ o REGION il 5 a 101 MARIETTA STREET, N.W., SUITE 2000
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/4.-- PROCEDURES " NORMALr ABNORMAL, EMERGENCY AND PAGE 21
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R5656L66EUAL 66UTR6L
.GUESTION' 4.04s (1.00) 1Upon loss of_'a DC bus,:AP-10,-Restoration of'DC Buss se allows. .
scrosstying to another-DC bus if-the'deadibus cannot be reenersized from
. afbattery. charger. - Hhich of the-followins busses can-be-crosstied?-
ca..1-Ic with 2-I-
- b. 1-I with 1-III
-c.E2-I with 2-II
~d. 1-I with 2-II QUESTION 4 .~ 0 5 (1.'00)
Which one of the following procedures states as its entry conditions "This. Procedure is entered from 1-ES-3.4, SI termination following steam senerator tube rupture"? Ja. 1-EP-3, Steam Generator-Tube Rupture b.il-ES-3.1, SGTR Alternate Cooldown by Backfilling RCS
~
- c. 1-EP-1': Loss of1Rea'ctor Coolant
- d. 1-ES-3.'3r SGTR.with' Secondary Depressurination 4
+ e QUESTION- 4.06 ( .50) A rod control l logic' cabinet internal' failure can be verified in the-control room.-TRUE or FALSE? OUESTION- 4.07- ( .50) , :For a reactor startup, it i s required by the procedure to 30 critical while 1
-the power' indication is still in the source range. TRUE or FALSE?
i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) r I t o
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. 4. ' PROCEDURES - NORMAlt :. ADNORM AL,1 EMERGENCY AND' PAGE- 22 .
.i - RADIOLOGICAL CONTROL , 4 ____________________
- i
{ L - j ~ QUESTION. 4.00. (1.00) ; [~ 'Given~theEfollowing conditions state /whether or not it is allouable'per ~ -l l OP 5.2. Reactor Coolant System to open.the RCP'#1 Seal Bypass Valve.~- l l RCS pressure 1500 psis
.41Eteal leakoff valve open -
11' seal leakoff flowrate < 1-spm Seal 1. injection;flowrate to.each pump 8 spm ! i i-I 0UESTION- 4.09 (1.50') ; ! Match the descrip'tions below-with the appropriate Presrure/ Temperature ;
-operating-curves labelled A through G on the attached graph. ,
t
- a. Administrative RCP-NPSH limit. (seal delta.p >200 psid) l Lb. Point at which Over pressure protection must be inserted, l 1
- c. RHR-pressure 11 imitation. !
i , GUESTION 4.10 ( .50) , At what-mpe limit does an area become-an Airborne Radiation Area? . l' t, i: QUESTIG'N. 4.11 (1.00). i l: I i' What are/is the entry condition (s) for:AP 1.6 "RCCA Deviation from Tave l Control *? : l l t
- = QUESTION 4'.12 ( .50) t i i
- The containment instrument air receivers capacity is only-sufficient to 'f' f permit operation of all instrumentation for ______ minutes upon a loss of
.the containment instrument air compressor. (fill-in th'e blank) :
l ' i I I j; ,0UESTION- 4.13 ( .50) l t r l_ Will adverse containment. conditions cause affected control room indications t U to indicate higher or lower than actual conditions?- f (r**r* CATEGORY 04 CONTINUED ON NEXT PAGE *****) k c I 1 1.- 4
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QUESTION 4.14 (1.00) EEP-0 ' Reactor Tripcor SI' lists four.' paths" to check for a charging /SI pump:in the: event of an SI signal. What are these four " paths' AND how should.they.be aligned 1(opened / closed)?
- . QUESTION 4.15 (1.00)
With regard-to. personnel DOSIMETRY list THREE conditions under which' an employee should leave ~ a' work area and contact Health Physics.
- QUESTION 4.~16 (1.00)
;Following a-valid. reactor trip and safety injection,:what.are the' Reactor
. Coolant Pump Trip Criteria?- (give values' including adverse containment)
QUESTION 4.17 (1.00)-
~
Per AP-35' Loss of Containment Air Recirculation Cooling'rIF a complete. loss-of? chilled water. cooling. occurs the operator. is to ci'ose.TV-CC-liSA a B and
! align an alternate sou'rce of cooling medium to'the air.recirc coolers.
From where is this alternate source supplied AND how'is the alignment ac.complished?
.00ESTION 4.18 (1.00)'
Per:AP;48,1 ' Charging-pump cross-connect't. under what condition may charging' pumps be cross-connected?
' 00ESTION 4.19 (1.00)
For the EPs list all the conditions which establish adverse containment atmosphere. (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE'*****)
f H49 UNITED STETES h, NUCLEAR REQuLATORY COMMISSION
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~~~~ RUU5ULUUYEAL UUUTRUt i
'GUESTION' .4. 20. (1 00).
Durins; implementation of ECA-0.0 '. Loss ofrall A/C' Power'r~.a red path CSF
~
- on containment occurs.1 Which procedure should the operator perform.
EXPLAIN.' '
~ (1.00)-
LOUESTION .4. 21 - FRP.-P.1 " Response.to Imminent Pressurized. Thermal Silock",'has the oper.ator
- check for'SI termination' criteria-relatively' ear,1y in the procedure and
. wit's less rest ~r ictive conditions than in the EPs. Give'TWO bases for securins SI early into this procedure.
00ESTION- 4.22 (1.50) a.:While'performins a stepfin a Periodic Test the operator finds a step which'is invalid with the.existins operating statusLof a system. How
-should the operator' document non performance of the step?'
f
- b. Due,to plant conditiens'(e.g. mode inwhich system'not required)--the .
performance of a Periodic-Test can not be completed. .Hnw does the operator ' document -non performance cif the . procedure?
~
GUESTION' 4'.'23 (1.00) AP-4 ' Malfunction of NI' for NI-43 failure-has the operator remove computer points N0045A & N0046A from: scan if reactor power is >5%. EXPLAIN why
.this action.is necessary upon an NI-43 failure?
OUESTION' 4'.24 (1.00) A, loss of electrical power has occurred to an 4KV Emergency Busr the EDG
' has started and picked up the bus & loads. The Pressurizer backup heaters had lost power, however, power is now available. Explain how the-operation
'of.the heaters is affected by the sensed undervoltase AND Explain how to restore the heaters to normal operation.
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UNITED STATES d'~g@ Hi%'% NUCLEAR REEULATORY COZ"IISSION
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- RADIOLOGICAL: CONTROL ,
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-GUESTION 4.25 (1.00)
Per 0P-58 ' Full-Length Rod Control System", when INSERTING control rods caution .must: be t taken to prevent rod drive mechanism damage'uhile in Hanval'or: Bank Select. At what point (steps)_couldethis damage occur
'AND how'is it prevented? ,
'GUESTION 4.26 (1.50)
A-operator-is manually starting 1H. Emergency Diesel' Generator.- 'During the start an annunciator in the diesel' room alarms " Fire Trouble". Would the receipt of this alarm be abnormal under this condition? EXPLAIN. QUESTION 4.27' (2.00)
~
a..In accordance with ADM 14.0 " Tagging of Systems and/or Components", WHERE on the Danger-Hold (red) tag should the Tagging Record Number be recorded?
;b.1What'is-done.with part 2 of the Danger-Hold (red) tag once it is completed?
'0UESTION 4.28 (1,50)
In accordance with ADM 19.1 " Operations Records Administration", Station Los readings which are out of Frecification shall be annotated as such.
- HOW1are out of specification readings annotated?
0UESTION 4.29 (1.50) Give.the' location of the following support centers that are manned during a plant emergency. a.c0perat:ons-Support Center.
-b. Techniccl Support Center.
- c. Emergency Operations Facility.
I (****; END OF. CATEGORY 04 *****)
'(************* EN D - OF EX A MIN ATIO N * * * * ** * * * * * * * * * )
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P K8tu UNITED STATES
# ^q3 NUCLEAR REOULATORY CCMMISSl!N
[ f, 'o REGION il 5
* "$ 101 MARIETTA STREET, N.W., SUITE 2000 a
'f ATLANTA, GEORGIA 30323
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s I i I I L l l i i l o f I l i 4 __.,m,-...___,.. , _ . , . _ _ _ . , , _ _ . , _
. f o ca v o 5/t veio efficiency e (Net work out)/(Energy in) 2
,o cg, s a y,t + 1/2 at E = me- -
XE = 1/2 mv a = (Vf - 13 )/t A = AN A = Ag et PE = mgn vf = V, + at * = e/t a = In2/t1/2 = 0.693/t1/2 I/2 W = v :P A= nD l' 4 [(gl /2)
- IIb))
cE = 931 am -
. m = V,yAo -Ex Q.= m,ah I = I,e Q = mCpat 6 = UA4 T I= I,e'"*
Pwr = Wfah I = 1, 10'*/ M TVL = 1.3/u sur(t) HVL = -0.693/u P = P,10 P = Po e*/ SUR = 26.06/T SCR = 5/(1 - K,ff) CR, = S/(1 - K,ff,) SUR = 26s/1* + (a - o)T s CR)(1 - K,ffj) = CR2 (I ~ kdf2) T = ( t*/c ) + [(8 - o V Iol
- M " IIII - Edf) = CR /CR, j y = g/(o - s) M = (1 - Keffo)/(I - K,ff1)
T = (s - o)/(Io) SDM = ( -Kgf)/Kdf a = (Kgf-1)/Kdf ' 'Keff/Keff L" 10 sec0nos 1 = 0.1 seconds i o = [(t*/(T K,ff)3 + [I,ff /(1 + IT)] Idj=Id j P = (tov)/(3 x 1010) I jd) 2 ,2gd 2 22 l t = eN R/hr = (0.5 CE)/d2(,,g,73) f R/hr = 6 CE/d2 (f,,g) . Water Parameters Miscellaneous Conversions l 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010 ap, l 1 ga1 . = 3.78 liters 1 kg = 2.21 lbm 1 ft' = 7.48 gal. 1 np = 2.54 x 103 Stu/hr . Density = 62.4 1 /ft3 1 m = 3.41 x 100 5tu/hr l Density = 1 gm/cu lin = 2.54 cm Heat of vaporization = 970 Stu/lom 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm 'C = 5/9 ('F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in. e = 2.713 M- ~TNg
5 . volume, ft'/It EMhelpy.Otu/lb Entropy. Die /lb a F T p Water Evep Steem Water guep Steem Water Evap Steem e, e. 4 a, *. s 4, S 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 82 0.08859 0L01602 3305 3.00 1073A 1076A 0.0061 2.1706 2.1767 Z5 85 0.09991 0.01602 2948 2948 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 40 0.12163 0 01602 2446 2446 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 45 0.14744 0.01602 2037.7 2037.8 18.05 1065.3 1063.4 0.0361 2.0901 2.1262 50 50 0.17796 0.01602 1704.8 1704.8 28.06 1059.7 1087.7 0.0 M 5 2.0391 2.0946 60 GO 0.2561 0.01603 1207.6 1207.6 38.05 1054 0 1092.1 0.0745 1.9900 2.0645 70 70 0.3629 0.01605 868 3 868 4 48.04 1048.4 1096.4 0.0932 1.9426 2.0359 80 80 0.5068 0.01607 633.3 633.3 468.1 58.02 1042.7 1100.8 0 1115 1A970 2.0086 to 90 0.6981 0.01610 468.1 300 350.4 68.00 1031.1 1105.1 0.1295 1A530 1.9825 800 0.9492 0.01613 350.4 110 265.4 77.98 1031.4 1109.3 0.1472 1A105 1.9577 110 1.2750 0.01617 265.4 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 120 1A927 0.01620 203.25 203.26 97.96 1019A 1117A 0.1817 1.7295 1.9112 130 130 2.2230 0.01625 157.32 157.33 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 140 2A892 0.01629 122.98 123.00 97.07 117.95 1008.2 1126.1 0.215C 1.6536 1.8686 150 150 3.718 0.01634 97.05 160 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1A487 160 4.741 0.01640 77.27 62.06 137.97 996.2 1134.2 0.2473 1.5822 1A295 170 170 5.993 0.01645 62.04 50.22 148.00 990.2 1138.2 0.2631 1.5480 1A111 100 ISO 7.511 0.01651 S0.21 40.96 158.04 984.1 1142.1 0.2787 1.514S 1.7934 100 190 9.340 0.01657 40.94 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.77G4 200 200 11.526 0.01664 S 62 1.7600 210 27.82 178.15 971.6 1149.7 0.3091 1.4509 210 14.123 0.01671 27.80 26.80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 212 14.696 0.01672 26.78 220 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 17.186 0.01678 23.13 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 230 20.779 0.01685 19.364 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 240 24.968 0.01693 16.304 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 250 29A25 0.01701 13.802 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 240 35.427 0.01709 11.745 11.762 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 270 41.856 0.01718 10.042 10.060 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 ISO 49.200 0.01726 8.627 i 7.460 259.4 917.4 11762 0.4236 1.2258 1.6473 290 290 57.550 0.01736 7.443 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 300 67.005 0.01745 6.448 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 5.609 5.626 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 89.64 0.01766 4.896 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 340 117.99 0.01787 3.770 360 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 860 153.01 0.01811 2.939 380 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 340 195.73 0.01836 2.317 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.844s 1A630 < 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 420 305.78 0.01894 1.4808 1.4997 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 440 381.54 0.01926 1.1976 1.2169 441.5 763.2 1204.8 0.6e05 0.8299 1.4704 460 460 466.9 0.0196 0.9746 0.9942 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 450 566.2 0.0200 0.7972 0.8172 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 0.6545 0.6749 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 523 812.5 0.0209 0.5386 0.5596 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 540 962.8 0.0215 0.4437 0 4651 562.4 625.3 1187.7 0,7625 0.6132 1.3757 560 SEO 1133.4 0.0221 0.3651 0.3871 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 580 1326.2 0.0228 0.2994 0.3222 617.1 550.6 1167.7 0.8134 0.5196 1.3330 500 600 1543.2 0.0236 0.2438 0.2675 620 646.9 506.3 1153.2 0.8403 0.4689 1.3092 620 1786.9 0.0247 0.1962 0.2208 640 679.1 454.6 1133.7 0.8666 0.4134 1.2821 640 2059 3 0.0260 0.1543 0.1802 660 714.9 392.1 1107.0 0A995 0.3502 1.2498 660 2365.7 0.0277 0.1166 0.1443 680 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 640 2708.6 0.0304 0.0808 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 700 3094.3 0.0366 0.0386 0.0752 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 705.! . 705.5 3208.2 0.0508 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) A.3
9eheme. t:8/in Enthalpy. Str/lb Er.teopy. Sty /itaF tasegy Cw/in TMP Ccter Evep Steem Ccter Eep 8 teem hier see m Pro ** F c;ter Erep Seesm fn*e*- pe pole et vg v s As A t A s s, s, s, e, og 0.01602 3302.4 3302 4 0 00 1075.5 1075.5 0 2.1872 2.1872 0 1021.3 e.8886 e.0686 32.018 35 023 0.01602 2945.5 2945 5 3 03 10738 1076 8 0 0061 2.1705 2.1766 333 1022.3 0.10 0.10 13.50 1067.9 1081 4 0 0271 2.1140 2.1411 13.50 1025.7 0.15 0.15 45.453 0 01602 2004.7 2004 7 0 01603 1526.3 1526 3 21.22 10635 1084.7 0 0422 2 0728 2.1160 2122 1028 3 S.20 0.20 53.160 0 0641 2.0165 2.0809 32.54 1032 0 0.30 64 484 0 01604 1039.7 1039.7 32.54 1057.1 1089 7 0.30 40.92 10524 10933 0.0799 1.9762 2.0562 40.92 1034.7 e.go 0.40 72.869 0.01606 792.0 792.1 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 2.0370 4732 1036 9 0.5 D.5 79.586 0.1028 1.9186 2.0215 5324 1038.7 0.6 0.6 85.?!S 001609 540 0 540.1 53 2s 1045 5 1098.7
.466 94 58 JD 3042 7. 4400A < CJ .. 3.8966.2.0083 . 58,14,.1060L3 .02
.#- 2- *
- 8.7 - 90 09m n01610 466.93 0.1117 1.8775 1.9970 6239 1041.7 0.8 6219 1040 3 1102.6 08 94 38 0.01611 411.67 411.69 0.01612 368 41 368 43 46 24 1038.1 1104.3 0.1264 13606 1.9870 6624 1042.9 0.9 0.9 98 24 1.0 101.74 0.01614 333.59 333 60 69.73 1036.1 11058 0.1326 13455 1.9781 69.73 1044.1 1.0 2.0 126 07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 10512 2.0 118 73 109.42 1013.2 1122.6 0.2009 1.6854 1.8864 109.41 1056.7 8.0 3.0 141 47 0.01630 118 71 4.0 4.0 152.96 0.01636 90 63 90 64 120.92 1006 4 1127.3 0.2199 1.6428 1A626 120.90 1060.2 '
73.53 130 20 1000.9 1131.1 0.2349 1.6094 13443 130.18 1063.1 8.0 S.0 162 24 0.01641 73.515 6.0 170.05 0.01645 61.967 61.98 138 03 996.2 1134.2 0.2474 1.5820 1A294 138.01 1065I 5.0 7.0 176 84 0.01649 53 634 53.65 144.83 992.1 1136 9 0.2581 1.5587 13168 14431 1067.4 7.0 47.328 47.35 150 87 988.5 1139.3 0 2676 1.5384 13060 15034 1069.2 8.0 8.0 182 86 0 01653 9.0 9.0 18827 0 01656 42.385 42 40 156.30 985.1 1141.4 0.276C 1.5204 1.7964 15628 1070.8 10 193.21 0.01659 38.404 38 42 161.26 982.1 1143.3 0.2836 1.5043 1.7879 161.23 1072J 10 0 01672 26.782 26 80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 180.12 1077.6 14.696 14.696 212.00 15 213.03 0.01673 26 274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 070 20 087 196 27 960.1 1156.3 0.3358 1.3962 1.7320 196.21 1082.0 20 20 227.96 0.01683 30 250 34 0.01701 13.7266 13 744 218 9 945.2 1164.1 0.36S2 1.3313 1.6995 2182 1087.9 70 10 497 236.1 933 6 1169.8 0.3921 1.2844 1.6765 236 0 1092.1 40 to 267.25 0 01715 10 4794 bw 261.02 0.01727 8 4967 8514 250.2
- 923.9 1174.1 0 4112 1.2474 J.6585 250.1 1095J 50 60 292.71 0 01738 7.1562 7.174 262.2 915.4 1177.6 0.4273 1.2167 1.6440 262.0 1098.0 60 70 302.93 0.01748 6.1875 6 2C5 272.7 907.8 1180 6 0 4411 1.1905 1.6316 272.5 1100.2 70 80 312.04 0.01757 5 4536 5 471 232.1
- 900.9 1183 1 0.4534 1.1675 1.6208 281.9 1102.1 30 90 320 28 0 01766 4.8777 4 895 290 7 894 6 1185.3 04643 1.1470 1.6113 2904 1103.7 90 100 327.82 0 01774 4.4133 4.431 298.5 888.6 1187.2 0.4743 1.1284 1.6027 298 2 1105.2 100 120 34 .27 0.01789 3 7097 3 728 312 6 877.8 1190 4 0 4919 1.0960 1.5879 312.2 1107.6 120 140 353 04 0 01803 3 2010 3 219 325.0 868.0 1193 0 0.5071 1.0681 1.5752 324 5 1109.6 140 160 363 55 0 0;815 2.E155 2.834 3361 859.0 1195.1 0.5205 1.0435 1.5541 335.5 1111.2 160 2.531 346 2 850 7 1196.9 05328 1.0215 1.5543 345.6 1112.5 180 180 373 08 0 01827 2.5129 200 351 80 0 01839 2.2659 2.287 355.5 842.8 1198.3 0 5438 1 0016 1.5454 354A 2113.7 300 400 97 001865 1.8245 1.8432 3761 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115.8 250 250 392.9 1117.2 300 300 417 35 001859 1.5233 1.5427 394 0 808 9 1202 9 0.5682 09223 1.5105 350 411.73 0 01913 1.3064 1.3255 409 8 7942 1204 0 0 60i! 08909 1.4968 4086 1118 1 350 ,
1.1610 424.2 780 4 1204 6 06217 0 8630 1.4847 422.7 111E 7 400 400 44460 0.0193 1.14162 450 4t6 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 0.6360 0.8378 1.4738 435.7 1118.9 450 0 90787 0 9276 449.5 755.1 1204 7 0.6490 0 8148 1.4639 447.7 1118.8 900 500 AU 01 0 0193 456.9 1118 6 550
$50 476 94 00199 0 82183 08412 460.9 743.3 1204 3 06611 0.7936 1.4547 0.74962 0.7698 471.7 732.0 1203 7 0.6723 0 7738 1.4461 469.5 111E.2 600 400 48520 00201 488.9 1116.9 700 700 .503 08 0 0205 0.63505 0 6556 491.6 710.2 1201.8 0 692R 07377 1.4304 0.5690 509.8 689 6 1199 4 0.7111 0.7051 1.4163 506 7 1115.2 800 833 51821 0 0209 0.54809 900 ti! 93 0 0212 04796S 05009 526 7 669 7 11 % 4 0 7279 06753 1.4032 5232 1113.0 900 2000 5 *4.5B 0 0216 0 42435 0 4460 542.6 f 50 4 1192.9 0.7434 0.6476 1.3910 5336 1110.4 1000 557.5 631.5 11891 0 7573 0.6216 1.3794 553.1 1107.5 1100 1100 SLE 2d 0.0720 0.376f3 0 4006 3200 '367.19 00223 034013 0.3625 571.9 6130 1184 8 0.7714 0.5969 1.3653 566 9 1104.3 1200 585 6 544 6 1180 2 0.7843 05733 1.3577 580.1 1100 9 1300 1300 L77 42 00227 030722 0.3299 576 5 1175 3 0.7966 05507 1.3474 592.9 1097.1 1400 14C0 537 07 0 0231 0 278/1 0 3018 598 8 1500 596 20 0 0235 02h372 0.2712 611.7 558 4 11701 0.8035 0 5233 1.3373 605 2 1093.1 1500 6721 465 2 1135 3 0 8625 0 4256 1.?b81 662 6 10GS 6 2000 2000 635B0 0.02's? O16766 01833 731 7 361 6 1093 3 C 9139 03206 1.2345 118.5 1032.9 2500 2500 65d 11 0 02cf 01020'+ 0 1307 801 8 2184 1070 3 0 9728 01891 1.1619 782 2 973.1 3000 3000 695 33 0 0343 0 050/3 0 0850 906 0 0 906 0 1.0612 0 1.0612 875.9 875 9 320s.2 3298.2 701 47 0 0508 0 0 050d TABLE A3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)
A.4 _ . . . a
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~:1. - PR'INCIPLES '0F NUCLE AR.: POWER -: PLANT , OPER ATION, PAGE ~ ~ ~
'~~~~TU5Rb66 Udb5C5,~55dT 5RdU5FER d D ELU56'FL5E'
~
1 ANSWERS:--- NORTH' ANNA?la21 ' -86/06/23-CASTO, C [ ANSWER. 1.01 (1.00) i. V lb' 2 REFERENCE- l DPCr. Fundamentals of-Nuclear = Reactor En3 i neering,:p'. 96 0U1/000-K5'56 1(2.'8/3.1)- GGNS:'OP-NP-511
. North Anna NCRODP-86.1-Sec. 6, ' p .: 6.32 ANSWER l'.02 (1.00)
'd:
-REFERENCE BFNP RANKINE CYCLE LP,P.5,7-8
' North:-Anna Thermo'Sec. VI Obj.
H. p. 6.21 / ANSWER 1.03 (1.'00) d. REFERENCE'
-North: Annat:NCR00P 86.2 Sec. 4 IANSWER 1 ~. 0 4 (1.00)
'b.
REFERENCE NAPS 1-ES-0.50, P . :6 IANSWER' 1.05 (1.00) Ed. 7 -. i I l' e r
. - , - . . . . . . , . . - , - _ - ~ . - . - . .-_ -. , . - . . . . - - -. - . ;. . . - -
~
g># K8CO, UNITE 3 STATES
#p o NUCLEAR REEULATORY COMelSSION
- -t-y g REGION il
- g 101 MARIETTA STREET, N.W., SU?TE 2300
- 2 o, ATLANTA, GEORGIA 30323 s,...../
i l l l [ I l 1
r 7 (g- % , v , x-N 1'~:
.- PRINCIPLES OF NUCLEAR-POWER PLANT OPERATIDAr- . PAGE 27'
~ ~
~~~~TUERb667U55fd5~~U55T TR5U5fER 5 5~FL 15 FL6U
< ANSWERS % ~ NORTH-ANNA la2 -86/06/23-CAST 01 ~C-i &
~
.MSHER 1.06' - ( 1. 0 0 ) -
d
REFERENCE:
Westinghouse _ Reactor Physics, pp.-I-5.63'- 76 HORr Reactor =Theoryr'. Sessions 38 and 39 DPC~r Fundamentals of Nuclear- Reactor Engineeringr Section VI-
~001/0005 RS.33(3.2/3.5)
ANSWER ,1 07z._ (1.50)'. - a.Jessentially the same. (0.75). -
'b'.-76 psis +/-2:psis (0.75)
REFERENCE-
' North Anna: Centrifugal Pump Char._ L.P. p. O GP HT.&LFF.Sec. III-B_pp. 326 & 329 ^
' ANSWER 1.00. (2.00) .
~a .-RTS: (0.5ea)
-b.'Decreasa
-c.-Increase
- d. Decrease.
- REFERENCE Steam ~ Tables -
1010-000-K5 02.(2.6/3.0)_ a ANSWER' 1.09 (2.00)'
'1.1 Boron Concentration Decreases (0,5)EMORE NEGATIVE-(0.5)
- 12. Fission Product Buildup (0.5) LESS NEGATIVE (0.5)
' REFERENCE-West. NTO -
P. I-5.31 i s
-+- V+,-- ---r, v, i,, .v,e-- - vw.- - - , . y -mo,, , - -,,.& - ,ww.. % svv-, - ,2v,, w e-e.4 vw, - - + ..wm-wr%--=,-,-..w -
s-.e- - -
c p2rar0 UNITED STATES
, . 9,g
, NUCLEAR RE20LATORY COMMl8880N y g REGION 88
; E 101 MARIETTA STREET. N.W., SUITE 2000 o, [ ATLANTA. GEORGIA 30323 9 *
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- J>
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,_-..,n,,m,, .,n, _--.-,-,-r.,,,n,,
3
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 28
~ ~
~~~~T HE5566'~U555E5"_HE5Y~TR U5EER 5U6~FEUED FL6U ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 1.10 ( .50)
True REFERENCE i North Anna: GP HT & FF p. 176 002/000-KS.01 ee y-sS S & ANSWER 1.11 (1.00) 4 DNDR and RCS overpressure gq~ /lG5 //E95/'2 D(Mh 5 4n W T/ [A N A //( % gl k j 4b f Ec I'$ ") REFERENCE North Anna I & C Sec.I. Pr r Press. Cntr1. K/A 002-000-KS.8 (3.4/3.9) ANSWER 1.12 (1,.00) cl _
- c. - - ,
7-/5 'I ' et a - 4 ,6 vn,s,nn /; ._. tu ..u. Leon) REFERENCE DSEP: 02-0G-A, p 26 North Anna 86.1 Sec. VII p. 7.16 ANSWER 1.13 (1.00) 4, 2, 1, 3 (0.33 pts for each suitch necessary to get in correct order) REFERENCE North Annat NCRODP 83 References
s C Ctt UNITED 37ATES NUCLEAR RESULAyORY COMMs3SION [ w kg REGtON pg y E ' M A LANTA EOMotA
'+9 ,o 4
t e i
,-,,v,r - - , , ~ . . , - .,m ,e.-- . . - . ,-. , , , ,,.w_-en,. en,.,,e_,,,,,,-,_.,.m.,-,.,.-..,.n, ,, ,-,_,_,----,,,,__,.,,e,n-.,, . , .-.n.,, . ,,,,n,_.,,-, m,_,,,,,m.-
v
~1. PRINCIPLES.0F NUCLEAR POWER PLANT OPERATION, PAGE 29
~~~~iUERF557sAsics? REAi iREssFER Es5 FLUi5 FL6s ANSWERS -- NORTH ANNA'1&2 -86/06/23-CAGTO, C ANSWER 1.14 (2.00)
- a. Thot = 630 des Tsat G 2235.psisc 7- / f
= 653 (+/- 1) C1.03 subcooling = JMT deg.F (+/- 1)
At
- b. For subcooled conditions-the reading would indicate 10 deg. (0,25) more subcooling than actually exist (0.25).
For superheated conditions the reading would indicate 10 deg. (0.25)
'less superheat than-actually exist (0.25).
REFERENCE North Anna Core Cooling Monitor Obj. 2. K/A Comp T/C (3.0/3.1) 002-000-A1.04 (3.9/4 1) ANSWER 1.15 (2.50)
- a. High relative flux - causes a greater reactivity change due to CRW being proportional to. flux-tip/ flux avg. therefore, the higher the relative flux the greater the change. (0.333 for area /0.5 for Exp.)
- b. larger effect_for the middle - due to absorption of neutrons which have a high probability of causing fincion. Whereas control rods at the edge absorp neutrons which have a high probability of leakage.
- c. ti has higher worth. When inserted ti depresses the flux around itself, this increases the flux in other regions, when #2 is inserted the flux has been depoessed therefore its worth is lower (than .ts worth in an unrodded cor r:) .
REFERENCE North Anna RUP pp. 6.11 6.12,6.19 Obj. B K/ A 001-000 -K5. 02 (2.9/3.4)
. _ ~
4#**% . UNITED STATES fi ti kg NUCLEAR REOULATORY CO'2%lSSl3N REGION il [ 101 MARIETTA STREET, N.W., SulTE 2900 g a
,o ATLANTA, GEORGIA 30323 i
J l l l i n
)
il l
. _ - . . ~ . . . _ . . _ _ . . . _ . , _ . _ _ . _ _ . _ , , . . _ - . _ _ _ , _ _ . _ , . _ . . . , _ . . _ _ . _ , . _ _ . _ _ _ , . _ _ _ _ . , , _ _ _ _ . - . . _ . _ _ _ _ . _ _ , _ ,
. ~. - - . .. - - , , . ~,
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^ ?1..:PRINCIPLESL OF-NUCLEAR! POWER PLANTLOPERATION, PAGE 30
~~~~ - ~
THERs557sARiEs? SEAT.TRAssFER 5UU~ELUiU~ELUs
' ANSWERS ~-- NORTH AFNA182. -
-86/06/23-CAST 0, C
- AN9WER 1.16 (2.50)
- a. ' increases-due to increasing non-condensable content.
p
- b. . increases-temperatur.e-out~of the coolers. drop lowering Tsat, this de -
creases Psat.
.c . ^ decreases-increases. chilled-water temperature thisfineresses Tsat, increasing Psat.
(0.333 ine/dec - 0.5 reason)
. REFERENCE'
- North Anna-NCRODP-91.2
-K/A--022-000-A1 02 (3.6/3.8)
ANSWER 1.17. (1.50) 1.~ Potential.large reduction in cladding ability to transfer heat'(0,5) - result'- fuel pellet end clad heatup (0.25). 2.. Blockage offcoolant flow. passages between fuel: rods-(0.5) - result-an additional reduction in the core's-heat transfer ~ cap 3bilityr 1: causing fuel heatup (0.25). REFERENCE North Anna NCRODP-95.2-K/A 000-011-EK3.13 (3.8/4.2) . J. +
.(1.50)
, ANSWER 1.18
~
! -a.:Without RCP-dri'ven pressurizer spray no adequate.means of mixins loop- ~ +
' pressurizer exist. (0.75) b.:Provides' reasonable assurance that even"a fairly rapid temperature drop will not cause problemstwith afloss of core shutdown margin. . (0.75)
!- ANSWER 1.19 (1.00) Due to the non-linear density characteristics of the water - / des.F change ofJdensity is seester at. higher temperatures. .f._ i a I< J 4 l
- -- , , , . - . . , . - , - . - . - . . , . - . - , - , - . . ~ - - _ . - . , ..-...,. - . ..- - - ,, - ._ . - , - . -
p* Uluge UNITE 3 STATES d ',* NUCLEAR REIULATORY COMMISSION E' o REGION il 5 a 101 MARIETTA STREET, N.W., SUITE 2000
- 2 o ATLANTA, GEORGIA 30323 s,
...../
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7
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 31
~~~ isEEs557sERICs- REEi iEEssFEE Es5 FEUi5 FE6W ANSWERS -- NORTH ANNA 1&2 -86/06/23-CAST 07 C REFERENCE North Anna ROP MTC p. 2.13 ANSWER 1.20 (1.00)
- a. The m a:< i m u m neutron population (equilibrium) is reached when K^n approaches zero. (0.5)
- b. Keff (0.5)
REFERENCE North Anna ROP 7.21 ANSWER 1-21 (1.00) The interaction with Hydrogen atoms in the water molecules. b'/ '66d 5 A/ de'c 5 r 4 d * / Q ^ 7-W #4 REFERENCE EIH: L-RG-602, p9 BSEP: 02-0G-A, pp 10 -11 North Anna 86.1 sec. III, p. 4.6 ANSWER 1.22 (1.00) This will change the water to steam at the same temperature. REFERENCE GGNS: OP-HF-502, p 7 North Anna NCRODP 03.1 Se. III. p. 3.8 ANSWER 1.23 (1.00) The ability to deliver a certain number of amperes for a specific number of hours before the cell voltage drops to a specific minimum value. REFERENCE North Anna: NCRODP 90.3 Sec. I t
1 g>3E8CO UNITED STATES d # NUCLEAR RE2ULATORY CCMMISSION [ p o - REGION il 5 a 101 MARIETTA STREET N.W., SUITE 2000
- ATLANTA, GEORGIA 30323
%...../
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- NORTH ANNA 182 -86/06/23-CASTO, C ANSWER 2.01 (1.00)
To prevent accidental opening (0.5) which would result in Accumulator . A '/ a l,%
- c c * ** 'A*"
dumping into the RHR system (0.5)b.rW SfS /<w of's . !c y ,tc gc
/'#/ "' ^)
REFERENCE North Anna: NCRODP-80.2 p. I 12 ANSWER 2.02 (2.50)
- a. This would have no effect on Letdown system operation. (0 5)
- b. There would be NO auto pump start (0.5) Letdown would isolate {CAF}(0.5)
- c. #&-it is wJt possi le E0.51 interlock exist with 15J7 (0.5) /'' + f i REFER NCE
" 7^ ' #
North Anna 0.P. 8.1 & Attachments ANSWER 2.03 (2.50)
- a. Providee quick reduction in the aiount of load on the buc in a de-gradedkaltage sj tuation. cc 7-/f' (1.0)
Containment Depressurization Signal (0,5). b,fg'Ihee
- c. y dg gf'ef trans fromvo thegascontrol M CAF} room M to the Diesel Room is blocked while the EDG is loaded onto the M s.(0,5)
REFERENCE North Anna: NCRODP 90.1 Sec II & II EDG. ANSWER 2.04 (3.00)
- a. B,C (0.5 ea.)
- b. 1. LHSI can only deliver enough flow to provide sufficient suction pressure to two HHSI pumps. (1.0)
- 2. The pumps are 900hp each this action limits loading on the EDG to within specifications. (1.0)
REFERENCE North Anna: NCRODP 91.1
i 4 UNITED STATES ff p Ka4 '% , NUCLEAR RESULATO^.Y COMMISSIGN
-2 e o REGION il I g 101 MARIETTA STREET, N.W., SUITE 2000
* - t ATLANTA, GEORGIA 30323
\,...../
a
+
.---ae e----w- --..r-w-,--m- ..,7-w,,.ww- ,,,w,..,, , , , _ . , , , _ _ _ _
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENC'l SYSTEMS PAGE 33 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CAST 0r C l
j '>1 sq" '
- 1 ANSWER 2.05 (3.00) -
1, u vs L L 6< < <./f
- a. RWST LON_ LOW level (C.5) and HHSI discharge valves open (0.5)- Prevents highly radioactive sump water from getting into the RWST which is vented to atmosphere. (1.0)
- b. Discharge to the HHSI pumps (0.5) Svetion from the Containment Gump (0.5)
REFERENCE North Anna NCRODP 91.1 Sec. II ESF t ANSWER 2.06 (1.00) The local alarm module has failed to function prope.cly. (1.0) REFERENCE North Anna NCRODP 93.1
<. 7 -/ '-4 ANSWER 2.07 (2.00) -
Latch housing and rod travel housing (0.25 ea.)"a cf pf;pe' s h4 T f AtQ)A~f4Alj4lM 54 a.
- b. incoming (0.5) /e. na opf) S t, f 7
- c. The voltage on the coils is reduced to prevent overheating of the stationary coils which could cause damage to the insulation. (1.0)
REFERENCE North Anra NCRODP 93.3 Sec. I& II ANSWER 2.00 (1.00)
- 3. Increasing (0.5)
- b. High-High vent radiation monitor trip (0,5)
REFERENCE North Anna: NCRODP 91.2
r ' g>S uh UNITED STATES
& c. ' NUCLEAR RE:ULATORY COM::lSSION El ' y'o ntatoN u 3 $ 101 MARBETTA STREET, N.W., SUITE 2000
'* e o ATLANTA, GEORGIA 30323
%,,...../
l l 1 i I i l l l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- NORTH ANNA 1&2 -06/06/23-CAST 0r C ANSWER 2.09 (2.50)
- a. This range is invalid with any RCPs operating, in this condition there is one RCP running (0.5). The range also has :n Hydraulic Isolator which is in the alarm condition.(0.5).
/4 r et 7-/f-N w
- b. 1 Ran3e (DP1) (0,5) c'CIX)g 9 p u dvL/"5 Nr*4 2-/e %
- c. Temperature of Impulse lines E0.25 ea.]
RCS temperature Wide range pressure d/p cell located out side of containment REFERENCE North Anna: NCRODP 93.3 ANSWER 2.10 (1.00)
- a. Place key switches to AUTO (0.5)
- b. y /WC T (c 7" 0' M (0,5)
REFERENCE NAPS RCS Pressure Inst. Lesson Plan. ANSWER 2.11 (2.00)
- a. This failure would have no effect on the start sequence,
- b. This failure would have no effect on the start sequence due to redundent air start valves.
- c. NORMAL the EDG would not start. E0.5 ea.]
EMERGENCY the EDG dr op -out relay would reset and the EDG would start. REFERENCE North Anna NCRODP 90.1 EDG ANSWER 2.12 (1.00) a REFERENCE NAPS ESF RSS Lesson Plan.
e p e* Ul% UNITED STATES
%, ' NUCLEAR RE2ULATORY COMMDSSl:N g v, g. REG 60N il
; E 101 MARIETTA STREET. N.W., SulTE 2000 o I ATLANTA, GEORGIA 30323 s.,...../
+
1 J 4 e 1 i F l l i l I
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r
.t 6
- 12 . PLANT DESIGN; INCLUDING. SAFETY AND LHERGENCY SYSTEMS .PAGE _______________________________________________________'
. ANSWERS'-- NORTH' ANNA 1&2~ - - 86/06/23-CASTO, C ANSWER: -2.13 (1.00) a.'
REFERENCE:
North. Anna!-NCRODP 92.2 SHS-
~
' ANSWER 2.14- .(1.00) a
-Rf.FERENCE
'NA NCRODP 89.4,.'Feedwater Systems-Main Feed" ANSWER: 2.15 (1 00).
d (improves circulation ratio and velocity acrossitube sheet)
-REFERENCE.
NA NCRODP 08.1,RCS-S/G'
' ANSWER _2.16 l .(1.00)L d
REFERENCE NA NCRODP 92.2, " Service Water' System' ANSWER 2.17 (1.0U) c. REFERENCE North Annat NCRODP 91.1, ESF-RSS O
\
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+y jg NUCLEAR REOULATORY CCMMISSION - o REGION il 0 E 101 MARIETTA STREET, N.W.. SUITE 2800 o, ATLANTA, GEORGIA 10323 s,...../
I
++ IMAGE EVAi.UATION o
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-_.__._._.______._-..._.._._.-.____..________-_._-__._._.....mm-_.._--_ _ _ _ _ _ ,
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- 2. PLANT.' DESIGN INCLUDING' SAFETY AND EMERGENCY SYSTEMS PAGE 36 ,
. ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTOR C 1
jANSHER '2.10 -
.(1.50) i r
r a) True (+.5 ea) 6 _.b ) False c) False j REFERENCE-NA'NCRODP-88.1, 'RCS-RCP' ,
- ANSHER 2.19 (1.00)
Halon E0.5] , CO2,CO.53 l REFERENCE . I NAPS Fire Protection Advanced Sills Training, i h i t 5 i ' i i i i t I i; i i t 4 I 4 h V e , ewe - w
F
- i.
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!-' 13 . ' : INSTRUMENTS AND' CONTROLS PAGE 37- }; ---------------------------- I t ANSWERS --~ NORTH ANNA 1&2 ~-86/06/23-CASTO, C
-ANSWER 3.01. (1.00) c.
REFERENCE !, ' North Anna RCS Press Instr.-LP i- ANSWER- 3.02 (1.00)-
- p
} .- .c REFERENCE Farley SD, 'CVCS", pp 13 NA NCRODP 08.3, 'CVCS' 004/020; K4.03(3.0/3.5)- } [ j ANSWER. 3.03 (1.00) b. REFERENCE North Anna NCRODP 93.10 ANSWER 3 04 (1 00)
- -d.
' REFERENCE I
North Anna NCRODP 93.11 ANSWER 3.05 (1.50)
- a. decrease
- b. remain the same
- c. remain the same l
REFERENCE North Anna NCRODP RPS
).
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- 3. INSTRUMENTS AND CONTROLS .PAGEL 38 j
. ANSWERS - . NORTH ANNA 1&2 -86/06/23-CASTO,.C 1
i
- g. ANSWER = 3.06 '(3.00) l
.a.Jrods out CO.25] Tref will be mak so Tave/ Tref mismatch and NI/ Turbine 1 I powersmismatch will both give a. rods out signal CO.753 l
~b. rods in.CO.253 Loop:1 Tave~increuses and auctioneered high Tave also .{
increases.- Tave/TrefE-mismatch gives a rods in signal C0.75] ! c.' rods out CO.25] the. power misnatch circuit of.tlue reactor control unit-responds only to rate of change of deviation between turbine and nuclear power but rod motion will occur due to the Tave Tref difference. CO.75] REFERENCE-Topic 6 LessonJ2 Fig. RS-5 and pp 55,fi9,-20 North Anna.NCRODP 93.5
- ANSWER 3.07 (1.50) a) 1-FP-P-2 (+ 5 ea) b) 1-FP-P-1 c) Both REFERENCE-j' NCRODP 92.1, " Fire Protection'
'at'086/000; K6~.01 (2.1/2.3); b -006/000;'K4.02 (3.0/3.4) c: 086/000) A4.01 (3.3/3.3) i ANSWER 3.08 (2.00) ,
I a. Manual ! b. Auto ; { c.- Auto. l f d. Hanual REFERENCE North Anna 93.10 I I I i f I l b l' I l- i ( i l : " y w e %===mp pm. , , - . _ ___ _ _ - - - -- - - - - -
i l km se004 UNITED STATES
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- 3. INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 3.09 (1.50) a) False b) False c) True
- REFERENCE NA NCRODP 93.2, ' E >
- c o r e NIS' pp 14-22 a: 015/000; A2.06(3.1/3.5)i.b: 015/000; K4.06(3.9/4.2) c: 012/000; K4.02(3.9/4.3)
ANSWER 3.10 (3.00)
- a. SI e. ph A
- b. SI f. ph 8 i /h U 7#'
4
- c. SI d . #" a T -1f 4 0 REFERENCE North Anna NCRODP 77 ANSWER 3.11 (2.00)
- a. Manual signal calling for rod movement Rods move IN. CO.b ea.]
- b. Tave-Tref deviation calling for rod movementpowever with 'D' selected rods DO NOT move. e '/-/T
- c. u<fd b U Manual signal calling for rod movement Rods r~~ %. 4mT.v*/4 T' " r/,f n<s
- d. Tave-Tref deviation calling for rod movement hovever 228 steps blocks movemente r o d s O ':' ' " ' nove. Ecaf] % 2-2 y REFERENCE North Anna NCRODP 93.5 i ANSWER 3.12 (1.00)
This indicates that the RCS is 25 deg. superheated. REFERENCE i North Anna NCRODP 93.4 l l l t j
1 g4 R800g UNITED STATES
# #g NUCLEAR RE20LATORY CCMMISSION
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- 3. INSTRUMENTS AND CONTROLS PAGE 40 ANSWERS -- NORTH ANNA 182 -86/06/23-CASTO, C
$vh ANSWEgg,l 3.13 (1.00) 1700 can be positioned by the operator as needed.
MOU{1701 MOV will not be able to open when required. [0.5 ea.] REFERENCE h' Wh% M / // ~ V'/" C ) North Anna NCRODP 08.2 K/A 005-000-K4.07 (3.2/3.5) ANSWER 3.14 (1.00) The indication would read high. REFERENCE North Anna NCRODP 93 ANSWER 3.15 (1.00)
- 1. OS pump breaker open
- 2. CDA signal reset REFERENCE North Anna 91.1 ANSWER 3.16 (1.00)
- 1. 30% design or 25% actual
- 2. N44 REFERENCE North Anna NCRODP 93.12 ANSWER 3.17 (1.00)
Auctioneered high delta T REFERENCE North Anna NCRODP Rf S 77
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- UNITEO STATES
[f,gA ttrog#{o,, o f, g NUCLEAR RE*ULATORY COMMISSION REGION il 3 g 101 MARIETTA STREET. N.W., SUITE 2000
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- 3. INSTRUMENTS AND CONTROLS PAGE 41 ANSWERS -- NORTil ANNA 182 -86/06/23-CASTO, C ANSWER 3.10 (2.00)
- a. NORMAL - Associated units vital bus Ecaf for bus 4] EO.50]
ALTERNATE - Appendir R dintribution panel ~ opposite Unit Vital bus E0.50]
- b. A locked transfer switch E0.50] located on EOG isolation panel in the Emergency Switchgear Room CO.50].
REFERENCE North Anna NCRODP EX NIS [< 3 c e 7-/ W ANSW' 3.19 M g 3. -
- a. By CTCV-100] a temperature control valve C.25] which isolates at 136 deg F. E.25]
- b. A flow switch downstream of TCV-100 E.25] a low flow alarm light at the aux. bids. ground floor on SG chemical nanel. E
- c. Fri u i L 4 ; . -" '"*i
1t i .b - v..a.25].MlW 4 7 je f -h REFERENCE North Anna NCRODP PRMS ANSWER 3.20 (2.00) 7-f f bc" f , f . f,/ g j
- a. cc 3 n l?/ t *nkw bEs*.yrxca->y f pfgA *',", ' "qw W,'Y b w, g Lo
- b. To on wrn a f'l o w path in the event of a station blackout.
REFERENCE North Anna NCR00P 26
i
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4.' . PROCEDURES - NORMAL, ABNORMAlt-EMERGENCY'AND PAGE 42
~~~~ REUE6[UU5EEE~UUUTRUE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER: 4.01 (1.00) d.
-REFERENCE
{ ' NAPS, 1-OP-3.4r p. 13. ,, i- ! i ! ANSWER 4.02 (1 00) 1
-d !
REFERENCE. MNS EP/2/A/5000/16.3 j CNS EP/1//A/5000/2F3, p.7. , t NAPS 1-FRP-I.3Ar p.3. ; ANSWER 4.03 ( l '. 0 0 ) b.- REFERENCE NAPS 1-ES-0.2, Foldout pa3e* ANSWER !4.04 (1 00)
- c. [
t REFERENCE
-North Anna AP 10.8 p. 4 i
ANSWER 4.05 (1.00)
- b. j t REFERENCE .
North Anna 1-ES-3.1 p. 1 ! t 6 i
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f 4 PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 43
--------------~~-~~----- ~~~~RE515E55fEAC E5sTR5t ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTOR C ANSWER 4.06 ( .50)
FALSE REFERENCE North Anna AR--1 la-4 ANSWER 4.07 ( .50) False REFERENCE North Anna 1-OP-1.5 p.10 ANSWER 0 ,g g (1.00) peo << REFERENCE North Anna OP 5.2 p. 5 ANSWER 4.09 (1.50)
- a. F
- b. D
- c. B REFERENCE North Anna OP-3.0 attach, 1 ANSWER 4.10 ( .50)
>.25 MPC REFERENCE North Anna GCT handbook p. 31
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44
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~~~~R565ULUU565L UUUTRU[~~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- NORTH ANNA la2 -06/06/23-CASTOR C ANSWER 4.11 (1.00) Tavg-Tref mismatch of 1.5 deg F REFERENCE North Anna AP 1.6 ANSWER 4.12 ( .50) 10 minutes REFERENCE North Anna OP 46.3 p. 3 ANSWER 4.13 ( .50) Higher REFERENCE North Anna EP-0 foldout page ANGWER 4.14 (1.00) A
- 1. RWST suction open O*g2f u~
- 2. VCT noction closed Myjc s ca.]
- 3. N o r nia l char girig closed
- 4. Letdown isolation closed RE"ERENCE North Anna EP-0 p. 3 ANSWER 4.15 (1.00)
If dosi'ieter: 3/4 or scale off scale dpOpped any m al f unc tiori lost
>*E880 UNITED STATES
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- 4. PROCEDURES - NORMAL, ADNORMALr EMERGENCY AND PAGE 45
~~~~Rdb5UE655U5E~66hTRUE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C
' REFERENCE North Anna CET handbook ANSWER 4.16 (1.00)
Verify Charging /SI flow (0.125) and (0 125) RCS pressure < 1230 psig (0.125) [1600 psig] (0.125) or CCW lost (0.5) REFERENCE North Anna Foldout page.for 2-EP-0 ANSWER 4.17 (1.00) Frca the service water system (0.53 by the use or a changeover. switch3 on the vr.ntslation panel CO.S]. RCFERENCE i4 orth Anna AP 35 p. 6 -5.1.6 ANSWER 4.10 (1.00) upon a loss of all charging pumps due to a fire on the affected unit. REFERENCE North Anna AP 40 1 ANSWER 4.19 (1.00) 19.7 psia or 40000 R/hr in containment ANSWER 4.20 (1.00) The operator should remain ire ECA-0-0 since the FRPs are written on the premise that at leact one E -bus is e n e r g i::e d . REFERENCE West. background info. for ECA-0.0
)
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- 4. PROCEDURES'- NORMAL, ADNORMAL, EMERGENCY AND PAGE 46
~~~~ ~ ~~~~'"~~~~~~~~~~~~~~~~~~
RAbEbdbb5dAL UbUTR6L ANGWERS --- NORTil ANN A la2 -86/06/23-CASTO,'C ANSWER 4.21 (1.00)
- 1. SI flow is a significant contributor to any cold leg temperature decrease CO.53
- 2. It can also be a significant contributor to an overpressure condition if the RCS is intact. CO.53 REFERENCE West. background info..for FRP-P.1 ANGWER 4.22 (1.50)
- a. The step may be marked N/A and initialed. [0.5]
- b. Submit PT critique sheet E0.53 enter procedure into the Action Statement Status Log E0.53.
REFERENCE North Anna Admin 11.2 p 10 , ANSWER 4.23 (1.00) This prevents an erroneous flux penality from the delta riux program. REFERENCE North Anna AP-4 p. 11 ANGWER 4.24 (1 00) Heaters will not operate due to Undervoltage trip EO.52 to reset place the control switches to DFF and then to desired position E0 53 f REFERENCE l North Anna AP-10.1 p. 4 ANSWER 4.25 (1.00) AT the last 5 steps to the fully inserted position E0.53 Jos the control rods in E0.5] 3 i i
h 4 Hig UNITED STATES
,cf Io,, NUCLEAR REOULATORY COMMISSION 3 f, o REGION il 5 a 101 MARIETTA STREET, N W., SulTE 2900
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- 4. PROCEDURES - NORMAL, ACNORMAL, EMERGENCY AND PAGE' 47
~~~~R 5656EUU565[~6UUTRUE--~~-------------~~~~---
ANSWERS -- NORTH ANNA 18.2 -86/06/23-CASTO, C REFERENCE North Anna OP-50
- ANSk'E A 4.26 (1.50)
No 00.53 The EDG logic has locked out the fire protection system heat detector C1.03 REFERENCE North Anna OP 6.1 p. 5 ANSWER 4.27 (2.00)
- a. Upper right-hand corner of the tag. [1.03
- b. It is given to the person performing the work C1.0]
RFFERENCE North Anna ADM 14.0 ANSWER 4 28 (1 50) Red circled E0.53'and explanation of why it was and what corrective actions have been taken E0.53 ontered into the Remarkn Section E0.53 REFERENCE North Anna ADM 19.1 ANSWER 4.29 (1.50)
- a. 3rd floor conference room of Mntnce Olds
- b. Servico Oldo (adj. to CR)
- c. Training Bldg.
REFERENCE North Anna E-plan pp. 1.5-9
900 g t s' sA p- = f .
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- 4. PROCEDURES - NORMAL,'ADNORMAL, EMERGENCY AND PAGE 47
~~~~EED56E665C5E~CU5TE6E~~~~~~~~~~~~~~~~~~~~~~~~
. ANSHERS -- NORTH ANNA 1&2 -06/06/23-CASTO, C REFERENCE North Anna OP-50 ANSWER 4.26 (1.50)
No CO.53 The EOG logic hau locked out the fire. protection system heat detector ti.03 REFERENCE North Anna OP 6.1 p. 5 ANSWER 4.27 (2.00)
- a. Uppor right-hand corner of the tag. E1.03
- b. It is given to the person performing the work ti.03 REFERENCE North Anna ADM 14.0 ANSWER 4 28 (1 50)
Red circled CO.53 and explanation of why it was and what corrective actions have been taken E0.53 entered into the Remarks Section E0.53 REFERENCE North Anna ADM 19 1 ANSWER 4.29 (1 50) a. b. 3rd floor conference room of Mntnce Blds Service 81d3 (adj. to CR)
- c. Training Bldg.
REFERENCE North Anna E-Plan pp. 1.5-9 r
u UNIVED STQTES
+*j@ asc ,q'o, NUCLEAR REGULATORY COMMISSION
[ REGION il
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICEtl3E EXAMINATION FACILITY: NORTH ANNA 182 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 86/06/23 EXAMINER: CASTO, C APPLICANT: _.,___ ,__________ ________ INSTRUCTIONS TO APPLICANTt Une separate paper for the answers. Write answers on one side only. Staple question sheet on top of the ensuer sheets. Points for each question are indicated in parentheses after the question. The panning grade requires at leant 70% in each category and a final grade of at least 80%. E : a rn t n a t i o n papei will be picked up si: (6) h o i..i r s after the e :anina ti on starts. 0F CATECORY ,% Or ADPLICANT'S CATEGORY WLyE p TOT AL GCORE VALUL CATEGORY __ p x. . . _ _ __--___-_-_ _-__-_-_ _ - _ _ . _ _ - _ _ _ _ _ _ _ _ _ _ . _ - - _ _ _ . . _ _ _ _ _ _ . . _ . S. THEORY OF NUCLEAR POWEP PLANT OFERATION, FLUIOS, AMD l THEPMODYNAMlCS l 30 00 _1", 01_ 0 _ _ _ _ _ _ _ . _ __
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INGFRUMENTATION 30.00 _ _' _' i . 0 0
.,_______ __ _ _ _ . _ _ _ . _ _ . . __ ____ ~. PROCEDUREC - NORMAL, ADNORMAL, EMERGENCY AND PADIOLOGICAL CONtPOL 30.00 '"'On "
- 9. ADMINIUTPAlIUC r, P O C L D U P E G ,
< CONOITIONS, AND LIMITATI0HC 10 M%J) 0 100.00 TOTALS O
All work dono on thi e aminat.on in tri y oun. I have neither given nor r reivmi aid. i Al'I IC0NI S 5 [ C N A I U b'I i l l _ - - - - , . - - - - - - - - _ _ _ , - . . . . . . . . - . . . - - . - - . . - . . . . - -,--..______--__._m_- -
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'5, DTHEORY 0F-NUCLEAR POWER PLANT.0PERATION, FLUIDS, AND PACE. :2
.l QUESTION- S'.01 (2.00)
The1 Core Cooling Monitor has determined the following readings result in the most 1 conservative margin to saturation!- N/R pressure PT-1444 = 2235 psis l- :Incore Thermocouple.(T/C) =.630 des.F [ -a.. Calculate'the margin to saturation. l l l b. Assume the T/C reference junction box temperature. indication has failed l l low (zero) and actual box temperature is 170 des.F. Explain;the effect; of this failure;on the resultant _ margin to saturation. Address both l t subcooled and.superheated conditions. . [ l Note
- Mointor assumes'160 deg.F reference temperature-upon failurer answer without regard to part a. above.
i I GUESTION 5.02 (1.50) For.each o'f the foll.owing sets of conditions. EXPLAIN which onelwould result
~
in the. greatest reactivity change due.to control rod insertion.- l Note!-Assume 100%. power, Bank D at 220 steps', BOL.- ; r
~
a . An area of high relative; flux vs. low relative flux. !
- j. -b.: Edge of the core vs. middle of the core. <
l c.l Rod 41'.(inserted) vs. rod-J2 inserted-beside rod'ti. l i
' GUESTION 5.03 (1.00) -~
Which two' safety limits necessitate operating within.the control band on --
= pressurizer pressure? j
> f QUESTION 5.04 (1.50) L List two dangers AND Explain lthe consequences of Boric' Acid precipitation post-accident. Assume no1 operator actions'are,taken to mitigate boron [
concentration. l;
)
i b
'I"
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.05 (1.50)
In response to ES-0.3 Natural Circulation Cooldown, it is directed that the RCS be borated to Cold Shutdown concentr2tica prior te RCS cooldown. This action results in an over-boration of the ACTIVE (core +1oop) portions of the system, considering this answer the following questions! ,
- a. Why is this action necessary during Natural Circulation Conditions?(.75)
- b. Why is this action necessary prior to RCS depressurization? (.75)
GUESTION 5.06 (2.00) , I
- a. During naturai cireviation, Explain how it is possible to form a bubble
- in the reector vessel head when indications show that the RCS is subcooled? (1.0) l
- b. How will prec=orizer leve) respond. (INCREASE, DECREASE, or REMAIN THE SAME) if the cackup heaters are energized with a bubble in the reactor vessel head? Assume normal pressurizer level and briefly EXPLAIN your answer. (1.0)
GUESTION 5.07 (1.00) Consider the equation below, answer the following questions: n N= neutron count rate i S (1-K ) S= source count rate N = K= Keff 1-K n= number of generations o
- a. How does 'N' respond to 'M approaching zero?
- b. Which term (s) determine (s) the total neutron production rate?
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) t
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Si THEORY:0F NUCLEAR POWER PLANT OPERATIONP FLUIDS, AND PAGE 4 QUESTION ~ -5.08 (2.00) Referring to Figures S-27A and S-270, answer the following questions
-concerning a-loss of one RCP transient from 22% power without a reactor trip.
- a. Why does loop 2 RCS flow increase at point 3? (0.5) b.' Why does loop 1 RCS flow increase at point 2? (0.5)
.c. Why does loop 2 S/G 1evel decrease (shrink) at point 4? (0.5)
- d. Why does auctioneered high Tavs (operatins loop) increase at point 9?' (0.5)
'00ESTION 5.09 (1.00)
Given that.a battery capacity is 1650 ampere-hoursr EXPLAIN the term
' ampere-hour".
LOUESTION 5.10 (2.00) The reactor.has shutdown and there are no indications of voiding or core damase. Refer to figures 4182.1,2,3. For the Moveable Incore Detector System answer the following questions,
- a. The-depression in the curves are due to passage in the vicinity of the gridt -WHAT causes these " depressions" to be created?
b, There are two distinct differences between the samma and neutron traces *
-(sharp vs. rounded edges fig. 182.3 and a neutron peak at the core
- bottom fig. 182.1 & _182.2), EXPLAIN why-each of these differences exist.
1-QUESTION 5.11 (2.00)
-M(z), the height-dependent correction factor, is used to modify the Nuclear Heat-Flux Hot Channel Factor limit given in the Technical Specifications.
Refer to figure t 186 and EXPLAIN why the curve decreases'at point A. 4 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) i r t l
'g$2 Kaou UNITED STATES
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4 15 . HTHEORY$0F NUCLEAR POWER-PLANT'OPkRA ION, FLUIDGr AND PAGE 5
-QUESTION 5 12: (l'.00)
Which of the'following actions will INCREASE North Anna's_ thermodynamic cycle efficiency?
- a. DECREASING power fromf100%_to 25% .
- b. LOWERING condenser vacuum from 29' to 25".
- c. REMOVING a high pressure FW heater from service'.
d.-DECREASING the amount of condensate" depression. QUESTION 5.13 (1.00) Attached Figure 4 168 shows a power history and four possible samarium traces (reactivity vs. time). Select (a, b, c, d) the correct curve 5 for displayin3 the expected samarium transient for the given power history. GUESTION 5.14 (1.00) , Which of the following post accident containment hydrogen contributors is-dependent on the radiation field intensity'inside containment for the amount'of hydrogen released?
- a. Ze + 2H 0--> Zr0 + 2H .
'2 2 2-
- b. 2Al + 3H 0-->'Al O + 3H 2 ~2 3 2
- c. Zn + 2H 0--> Zn(OH) +H 2 2 2
- d. 2H 0--> 2H + 0 1
2 2 2 f h I
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y 55 . - THEORY.-OF NUCLEAR POWER PLANT-OPERATION, FLUIDS, AND-
. PAGE 6 GUESTION. 5.15 (1.00) ,
l Which of the following statements concerning Xenon-135 production and removal'is correct?
- a. At full power, equilibrium conditions, about half of the xenon is Produced by iodine decay and the other half is produced'as a direct fission product.
b.- Following a reactor trip from equilibrium conditions, xenon peaks because-delayed neutron precursors continue-to decay to xenon while neutron' absorption (burnout) has ceased.
- c. Xenon production and removal increases lirearly as power level increases; i.e., the value of 100%-equilibrium xenon is twice,that of 50% equilibrium xenon.
c .- At low power levels, xenonzdecay is the major removal
. method. At high power levels, burnout-is the ma-jor removal mhthod.
.GUESTION 5.16 (1.50);
For each of the following STATE whether containment' partial pressure will increase or decrease and EXPLAIN-the factors which cause this change.
- a. Containment AIR volume increases.
- b. Draw a. vacuum without air recirculation fans then turn fans on.
c.1A Mechanical Chiller trips. QUESTION 5.17 (1.50)
-Indicate whether the following will INCREASE, DECREASE, or REMAIN THE SAME.
a .- Available RCP NPSH as volumetric flow rate increases. (0.5)
- b. Minimum required RCP NPSH as volumetric flow rate increases. (0,5) c.- Available NPSH to condensate (hotwell) pumps as condenser subcooling increases. (0.5)-
(***** CATEGORY 05 CONTINUED ON NEXT'PAGE *****)
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U 1; - j: F ~ <. I . l- 5.~ zTHEORY OF NUCLEAR POWER PLANT OPERATION,' FLUIDS,'AND PAGE 7-( ________________________________________________________ !; THERMODYNAMICS t ______________- b
-QUESTION
~
.5.18 . ( 1. 50 )-
l Indicate whether the following changes'cause the differential boron ! worth'to become:MORE NEGATIVE, LESS NEGATIVE, or REMAIN THE SAME. Consider ~each' separately. a.- Baron' concentration increases (0.5) Moderator temperature. increases [ b. (0.5) i; c. LCore age increases (at a constant boron concentration) (0.5) 1 f 0UESTION '5.19 '(1'.50) - Assuming a symmetrical'(ideal) . axial flu > shape, match the CONDITION-in Column A to the LOCATION.that it would occur in Column B, COLUMN.A COLUMN B. j a. MINIMUM Critical Heat Flov 1. BOTTOM 2.-Between BOTTOM & MIDDLE. 4
- b. MAXIMUM Actual Heat Flov 3. MIDD'LE
- 4. Between MIDDLE & TOP
- c. MINIMUM DNBR 5. TOP OUESTION 5.20 (1.50)
' Match the heat transfer location in Column A with its MAJOR heat transfer process in Column B, COLUMN A COLUMN B a '. From Center-Line to surface of a fuel pellet 1. CONDUCTION
- b. From outside of clad to coolant 2. CONVECTION
- c. Across-S/G tubes (primary to secondary) 3. RADIATION GUESTION 5.21 (1.00)
The fission process in a commercial ceactor requires the neutrons that are ' born" by fission to be "thermalized'. Which molecular interaction in.the reactor core is the most efficient in thermalizing neutrons? (***** END OF CATEGORY 05 *****) l'
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- 6. PLANT SYSTEMS DESIGNr CONTROLr:AND" INSTRUMENTATION PAGE 8
'00ESTION' 6.01- (2.50)
During Charging Pump transfer operations the following alignment exist!
-1.-Charging Pump 1-CH"P-1A is running in Automatic.-
2.-Charging Pump 1-CH-P-1B is'in the-Pull-TO-Lock position.
- 3. Charging Pump 1-Ch-P-1C is in the " connect *' position from 15J7~
Bus 1J and in the Automatic position.
.a. If Letdown-is in servicer what effect would this have (if any) on the-Letdown flow path. (0,5)'
- b. Should Charging Pump 1-CH-P-1A trip what action (s) woulo occur (address Letdown flow path and auto pump starts if any). (1.0) c.'Is it possible, with the above alignments present, to rack-in 1-CH-P-1C 15H7 (H-bus) Charging Pump feeder breaker? Why or-Why not? (1.0)
' QUESTION 6.02. (2.50)
Answer the following with regard to the Emergency Electrical Distribution System.
- a. What is.the. purpose of the " Stub Bus' on the Emergency Busses ill and 1J?
- b. State the two signals which will open the " Stub Bus' breaker.
- c. The 'H' Train EDG is loaded onto the H-bus in response to an under-voltage condition. The operator attempts to transfer EDG control fron- the Control Room to the Diesel room.
EXPLAIN the result of this action upon the EDG controls. GUESTION' 6.03 (2.50) Refer to figures 4 200.1 8 200.2 attached, and answer the following!
- a. .For the full range indication, describe why its reading is invalid AND describe any other abnormal conditions which exist under the current conditions. (1.0) b._ Figure 4 200.2 indicates several parameters, State which RVILS range (s) is/are affected by this current display. -(0.5)
- c. Explain how the Reactor Vessel Level Indication System is. compensated to maintain required accuracy during LOCA conditions. (Address 3 parameters and one physical design feature) E1.02
(***** CATEGORY 06 CONTINUED ON NEXT PAGE-*****)
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- a. Explain the effect (if any) of an air. VENT solenoid valve remaining open {
during an EDG start sequence.
- 4~ b. Exnlain~the effect (if any) of an air START solenoid valve - failing to -i op=n during an EDG start sequence.
{
- c. After a normal stop signal-is reset,~before the 60 sec drop-out time.has h expired, EXPLAIN how the EDG.would respond to EMERGENCY and NORMAL start j signals.' i i !
! )~
- . . QUESTION 6.05 (1.00)
!. Can the outside recirculation spray pump discharge be aligned to the
- j. LHSI pump discharge? Choose the correct answer from below. '
! a. Yes, for Unit 1 only.
.b. Yes, for Unit 2 only.
c.- Yes, for both Units.
- d. No, both Units are aligned at the suction of the LHSI pump. j
-QUESTION 6.06 (1.00) !
LWhich of?the'following describes the Service Water System automatic actions t on a single unit SI signal?
- a. All SW pumps start, that unit's spray header. isolation MOVs receive ,
an open signal. b.- Only the SW pumps supplied from that. units emergency buses start, that unit's spray header isolation MOVs receive an "open' signal.
- c. All SW pumps start, all spray header.' isolation-MOVs' receive an' "open' !
signal. d.- The SW pumps supplied from that unit's emergency buses start, all spray header isolation-MOUs receive an "open" signal. _l L (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i I L h i w.~. .
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i: , , f . i 15 . - PLANTISYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION 'PAGE 10 [ [ L LOUESTION 6.07 (1.00) l Which~one of the following describes the purpose of the time. delay (195 sec) in starting the Inside Recirculation Spray pumps'on a CDA signal? g a.. Prevents overloading of the emergency buses as loads are sequentially energized during accident conditions. .
- b. Allows time for the containment sumps to collect sufficient fluid to !
prevent depleting sump level, avoiding a loss of makeup inventory, j
- c. -Enhances core cooling by increasing reflood rate af ter a 10CA as the [
pressure drop between core exit and the break is reduced with a higher
~
l containment pressure. l
- d. Allows time for the fluid collecting in'the containment sumps to cool, to avoid flashing in the RS heat exchangers as the fluid is cooled by Service Water. t GUESTION .6.00 (1.00) !
( L
- j. Whieb one of the followingLmalfunctions could cause one of the over ;
- temperature delta T trip bistables to trip? .;
.a. Controlling turbine impulse pressure channel failing low.
l' ~ L b. Power range N43 lower detector failing~ low. , c.~ Reactor coolant flow detector failing, low. . [ t
.d. Controlling pressurizer level channel failing law. ;
OUESTION. 6.09 (1.00) : 1 With the-reactor at 100% power and the steam dump control system in the - Tavs mode, a 15% step loss of load occurs. Assuming no reactor trip occurs } the condenser is available, and the reactor operator manually OPERATES the ! control rods, which of.the following would occur if Bank 1 steam dump {
. valves failed.to open?
- a. Bank 2 would open. .l
- b. Atmospheric dumps would open. f
- c. S/G safeties uould open. l
- d. No other steam valves wou!d open. ;
i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) I l . ! . . . . . - I
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ll ;6. PLANT; SYSTEMS DESIGNr.CONTROLr AND INSTRUMENTATION PAGE 11- -l 4 4 I i.- l 0'UESTION 6.10' (3.00) k Indicate-what.happens to.the Rod Cor.' trol 1 System (rods in, rods'out, no change) and BRIEFLY explain why-the' change will oc will~not occur for the followins conditions. Rods:are in auto unless otherwiseEspecified, h a, Reactor power is.17% when the controllins" turbine'first stage impulse ! F pressure transmitter fails hish. ! !; 'b. Reactor power is 100% and loop 1 That fails hish. [ I c. Auctioneered high nuclear. Power is 50%, Ro'd Control is in manual. '
. Instrument testing is in progress on the turbine-power. input to rod ;
control which has turbine power at 100%. j All indications have been stable for the last hour. [ The Bank Selector switch is then placed in AUTO. i l GUESTION 6.11 (3.00) } ! -Indicate whether the followins valves receive a SAFETY INJECTION, PHASE : A or PHASE B input. signal (s)* . j
- a. Main feedwater isolation valves.
-b. Charging pump suction from VCT. -l ~
~
- c. Boric Acid tank pump to BIT valve.
- d. Charsins pump to recire stop valve. ;
- e. Accumulator test line. i sf. Steam Generator blowdown (TVDD 200C). l
-QUESTION 6.12 (2.00)
For'each condition EXPLAIN which component (s) would be generatins a rod movement signal and the response of Dank D rods to this signal. i Assume no other Rod Stop signals presentr Reactor at power. ! BANK SELECTOR SW. IN-0UT-HOLD LEVER PLANT PARAMETER 'D' POSITION '!
-( F
- a. Manual- In RIL LO-LO ALARM 180 steps I
- b. 'O' Hold Tavs-Tref +4 deg. 100 steps :
- c. Manual Out ".. Urgent Failure' 200 steps [
_d. Auto Hold Tavs-Tref -4 des. 222 steps ; L i i p' (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) f i i f I i e h t
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-6. PLANT.SYSTEMSrDESIGNr. CONTROL,.AND. INSTRUMENTATION - P A G E- 12 l
OVESTION 6.13 '( 1. 0 0 )J Reference figure:1230 Core Cooling Monitor. Afteridepressing.the' PSI
- f. 'HARGIN pushbutton the operator reads the display and finds that it, reads j
~25 des F. :In relation to saturation conditions of the RCS what condition j
,. does this indicate? j 00ESTION 6.14 (1.00) 'l-The operator is preparing to.P l ace the RHR system in service upon RCS , cooldown. PT-402 has failed high. DESCRIBE the effect(s) of this condition on the operation of MOV's 1700 and 1701 RHR suction valves. i i h QUESTION 6.15 (1.00) ; I I
- List three conditions / interlocks which are required to allow automatic j
[ swapover of the LHSI pumps to the recire mode following and SI. l l i. e !! QUESTION 6.16 (1.00) , ! 'During RHR system operation discharging to the Hot Legs, MOVs 1720AaB Cold- ! ! -Les Injection valves are closed. State the purpose of having these valves ( closed and de-energined during this made of RHR operation, l
, 'l QUESTION 6.17 (1.50) :
l t Answer the following for the CVCS process radiation monitor RM-CH-128/129. ;
- a. How is the monitor protected from hich CVCS temperatures? (include i c o n t r o l . s e t p o i n'+.s if any)
- b. Explain how low flow is sensed in the flow path-AND what indications;of l low flow are provided? i i
- c. What signal enables the high range monitor? l l
1 I l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i L
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'6. .FLANT SYSTEMS DESIGN,. CONTROL, AND INSTRUMENTATION- PAGE 13 l
! -i t
~. F i 0UESTION- 6.18- (2 00)' l 1 . .
r
- - .a. 'AFH. pump 3A has'three.indicatins lights above-the control switch. What i does..the AMBER light indicate?
.b.-'For proper AFW system operation MOV 1008 and D and HCV 100C are left i opened. Explain why this alignment is necessary for proper system !
[. operation. n a. r t-i
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- 7. . : PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 14
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656L665C5L 66NTR6L GUESTION 7.01 (1.00)'
A' hydrogen bobble formed in the. reactor vessel'is. eliminated by
~
I
- a. increasing pressuriner temperature above core thermocouple readings.
- b. injecting oxyaen into the reactor coolant system via the chemical and volume control system.
- c. maximi =ing coolant flow by running all reactor 1 coolant pumps, increasing letdown-_ flow to 120 spm, and placingz the, cation bed demineralizer in service in parrallel with the mixed bed demineralizer.
- d. venting the. reactor vessel head..
GUESTION 7.02 (1.00) Which one of the following procedures states as its entry conditions
- 'This procedure is entered from 1-ES-3.4, SI termination following steam
. generator tube' rupture *?
- a. 1-EP-3, Steam Generator Tube Rupture
- b. 1-EG-3.1, SGTR Alternate Cooldown by Backfilling RCS
- c. 1-EP-1 Loss of Reactor Coolant
- d. 1-ES-3.3, SGTR with Secondary Depressurization ,
.00ESTION 7.03 (1 00)
Which of the-following actions constitutes entry.into Mode 6?
- a. Increasing the baron concentration to 2000 ppm.
'b. Decreasing reactor coolant temperature to less than 140 F.
- c. .Detensioning the first reactor vessel heed stud,
- d. Decreasing keff to less than 0.95.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) k__ . _ _ . _ _ _ _ _ _ _ _ _ _
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QUESTION. 7.04. (1.00) During a natural circulation cooldown, what is the preferred method of reactor coolant system depressurization? QUESTION 7.05 (1.00) Which of'the following statements regarding Axial FluxLDifference (AFD) requirements for Unit 2 is correct? a.. Reactor power CANNOT-be increased above 50% ratedJthermal power unless the indicated AFD is within the target band.
- b. If the indicated AFD is outside the target band for more than 1 hour cumulative over a 24 hour period, with reactor power between 50% and 90% of rated, reduce thermal power to less'than 30% and reduce the
- power range Neutron Flux High setpoint to less than 35% within 30 mins.
- c. If indicated AFD is outside the target band and thermal power is 3reater than 90% rated thermal power, within 1 hour AFD must be restored within the band or power reduced to < 90%.
- d. Below 15% rated thermal power, penalties are given for being outside the target band due to the fact that uneven xenon buildup'in the core has an adverse impact at lower power levels.
QUESTION- 7.06 ( .50) A rod control logic cabinet internal failure can be verified in the control room. TRUE or FALSE? OUESTION 7.07 (1.00)
.Given the following conditions state whether or not it is allowable per ./
OP 5.2 Reactor Coolant System to open the RCP 11 Seal Bypass Valve. RCS' pressure 1500 psig 1 41 seal leakoff valve open ) 41 seal leakoff flowrate # 1 gpm Seal injection flowrate to each pump 9 gpm (***** CATEGORY 07 CONTINUED ON NEXT PAGE rwxxx) t
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t Qu'STION E 7 00: (1~00) cWhat are/is'the entry condition (s) for AP 1.6.'"RCCA Deviation from Tavs Control'T g
- 00ESTION 7iO9 (1.00)
What is the PREFERRED lineup for providing a means of decay heat removal , durins a refueling outese with the reactor vesser . head removed if all RHR pumps'are inoperable? e .;;= t ik QUESTION 7.10' (1.00)- i' During recovery from a refueling outager what tobl 6 utilized to relatch' the.RCC elements to their. drive shafts? , 00ESTION 57.11 (1.00) T s EP-0.*.leactor Trip or SI" lists four " paths" to check for a charsing/SI -s pump in.the event of an SI signal. What'are these f our -
- paths * ' AND ':hnu ,;
should they be. aligned (opened / closed)? ws , s - - QUESTION 7 12 (1.00) -> - With. regard to personnel DOSIMETRY 1ist THREE conditi'ons-under,whi'ch an employee.should leave a work-area and contact Health" Physics. QUESTION 7.13 (i.00) Following a valid reactor trip and safety injectionk'0 hat are the Reactor
' Coolant Pump Trip Criteria? (sive values includins adverseH-conpainment)
GUESTION 7.14 (1.00) *
-List the whole body administrative limits per-calender quarter that can be achieved (a) without any. additional approvalland (b) with the highest level of approval.
.t
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fDUESTION 7.15 (l'.00) Per AP-35' Loss of Containment' Air. Recirculation Cooling'rIF a complete' loss
-of.. chilled water-cooling occurs the' operator is.to close TV-CC-115A & B and align 1an' alternate source of cooling medium to the air recirc coolers.
From where is this alternate source' supplied AND how is the alignment eccomplished? QUESTION 7.16 (1.00) Per.AP 48.1 " Charging Pump cross-connect *, under what condition may charging pumps be cross-connected? 00ESTION 7.17 (1.00) During' implementation of ECA-0.0 " Loss of all A/C Power *r a' red path CSF on containment occurs. Which procedure should the operator perform. EXPLAIN. QUESTION 7.18 (1.00) FRP-P.1 ' Response to Imminent Pressurized Thermal Shock", has th'e operator-check for SI termination criteria relatively early in the' procedure and with less restrictive conditions than in the'EPs. Give TH0 bases for
. securing SI early into this procedure.
QUESTION 7.19- (1.50) a.1Hhile performing a step in a Periodic Testithe operator finds a step which is invalid with the existing operating status of a system. How should the operator. document non performance of the step? b..Due to plant conditions (e.g. mode inwhich system not required) the performance of a Periodic Test can not be. completed. How does the operator document non performance of the procedure? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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l b ~~~~ R5656LU55 EEL CUUTR6L'~~~~~~~~~~~~~~~~~~~~~~~ , L : 1- l i: , l: i -QdESTION 7.20 (1.00) .j AP-4 ' Malfunction of NI" for NI-43 failure has the operator ~ remove computer j_ points N0045Al& N0046A from~ scan if reactor power is >5%. ~EXPLAINLwhy i ! this action is necessary upon an NI-43 failure? , l QUESTION' 7 21 (1.00). (- l- A loss of electrical power has' occurred to ~ an.4KV Emergency Bus, the EDG j has started and picked up the bus-a loads. The! Pressurizer: backup heaters 3 had-lost powerr.however, p'ower is now available.- Explain how the' operation i
.of.the heaters is affected by.theLsensed undervoltage AND Explain-how to
~
! restore-the heaters to normal operation.: I QUESTION 7.22 (1.00) Per OP-50 ' Full Length. Rod Control System", when INSERTING control rods caution must be taken to prevent rod drive-mechanism damage while in Manual or Bank Select. At'.what point (steps)' could this damese. occur AND how is it prevented? l 00ESTION 7.23 (1.50) l
- 'A. operator is manually starting'1H Emergency Diesel Generator. During the
? - ' start an annunci.ator in the diesel room alarms " Fire Trouble" ' Would l[ the receipt of-this alarm be abnormal under this condition? EXPLAIN. 1
-GUESTION 7.24 (2.00)
- a. In accordance with ADM 14.0 ' Tagging of Systems and/or Components',
k!HERE on the Danger-Hold (red) tag should the Tassing Record Number i be recorded? ! b. What is done with part 2 of the Danger-Hold (red) tag once it is j completed? 1 1 5 l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) t 1 4
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'7. PROCEDURES' . NORMALS 1 ABNORMAL, EMERGENCY-AND PAGE 19
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~~~ E5656LUGIEEL EUUIRUL QUESTION -7.25 1( 1. 50 ) .
In accordance;_with ADR 19.11' Operations Recor_ds Administration *r -Station.
- Los readings wh:ch are out~of'speci.fication shallobe annotated as such.
HOW areJout'of' specification readings ~ annotated?. i QUESTION 7.26 (1.00) What is'the criterion used to verify the RCS is. solid when placing the-RCS in a solid water condition per OP-3.4? ] j. QUESTION 7.27 (1.00)
- ~When is control rod coolin3 cir required to be operating?
OUESTION 7.28 (1.00) ^ If you are i ri a 100 mrad / hour samma field for 45 minutes, what;is your dose in MREM.after 45 minutes? T i i i i i i 9 i (***** END OF CATEGORY 07 xxxxx) l' f l 4
UNITED STATES
+fstP 884,'pp NUCLEAR RE2ULATORY COMMISSION g REGION il g #
.- E 101 MARIETTA STREET, N.W., SulTE 2000
- 4 o, ATLANTA, GEORGIA 30323 s,
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8'..: ADMINISTRATIVE PROCEDUFES, CONDITIONSr ~~ 7 AND' LIMITATIONS PAGE 20 QUESTION; 8.01 (1.00) Refer to figures 4209.1 &-4289.2. Which. Table 3.4.1 applies to Unit 2?
- GUESTION 8.02' (1.00)
Diesel Generator 2H which supplies E-bus 2H is inoperative. LPSIP 2B
' supplied by E-bus 2d is inoperative. The Tech-Specs for ECCS and AC power 1 sources are attached. Which statement below is most correct concerning Mode 4 operations?
a.1 Mode 4 must be maintained (entry into mode 5 acceptable).
-b. Restore D/G to operable, status within 72 hours or be in at least Hot S/B within the next 6 hours and in cold S/D within the_followins 30 hours.
c.-Place the unit in Hot S/B within 1 hour, in at least Hot S/D within the next 6 hours and in at least Cold S/D within the following 30 hours.
- d. Startup-activities may continue; Mode 3'may be entered.
QUESTION :8.03 (1.00) With one Unit 2-. rod position indicator inoperable ~in mode ir the non-
~
indicatins rodEis moved from its last known-position more than 24 steps.- As an. alternate to reducing powerr the operator may-determino that rods' position using the movable incore detectors if done within.a certain time interval after the last' rod movement. What is that time-interval?
- a. immediately
- b. 15 minutes
- c. 30 minutes
- d. I hour
- e. 8 hours s
-(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
/
.. - __. _ - . _ . . _ _ _ _ . _ ..-m _. ., .
p3"4g UNITED STATES
# 'g'e . NUCLEAR RESULATORY COMMISSION
?' ReGeoN n 5 g -101 MARIETTA STREET N.W., SLITE 2900
# e o, ^ ATLANTA, GEORGIA 30323 s,...../
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~ ' 8 +J ADMINISTRATIVE PROCEDURES, CONDITIONS,-AND LIMITATIONS PAGE- 21' -
00ESTION 8.04. (1.00); ' { Thetunit 1: baro.n injection tank will tue INOPERABLE under which of the
- following conditions?
- a. If'it'contains 900 gallons of water.
4 l b. If it contains 950 gallons of borated water,
- c. ' If it contains114000 ppm.of borated water'and is at.120 deg. F. ..
- d. While in Mode 4 if it contains 800 gallons of. borated water.
. QUESTION ~ 8.05 (1.00)
What are the technical specification maximum heatup rates innany one hour. - ; period'for -the reactor' coolant system-(including the pressurizer)- for Unit- ' 1 AND Unit 2?
-QUESTION 0.06. ' ( 1. 0 0 ) :
If the lowest operating loop.Tave drops below in mode ir Unit 1 'a 2 tech specs allow ______ to restore Tave within______ the limit before proceeding tosplace the unit in hot standby.
- a. 541 degrees F, 30 minutes
- b. '541 degrees F, 15 minutes
- c. 547 degrees.Fr 30 minutes
-d. 547 degrees Fr 15 minutes
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) i I' i + ,,, , - - - - - - - . .
.pSOh UNITED STATES
#- g'k-NUCLEAR RE20LATORY COMMISSION O y, o REGION il 3 g 101 MARIETTA STREET, N.W., SulTE 2000
* * - ATLANTA, GEORGIA 30323
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- 8. 1 ADMINISTRATIVE PROCEDURESk CONDITIONS,'AND: LIMITATIONS -PAGE 22
- 0UESTION. 8.07 .(l'.00)
ADM 5.7' entitled " Correcting Data on. Completed Procedures
- makes.which of-the followins' provisions?
- a. It'is permissible to make corrections to' data which.has been recorded. This must-be performed in the following manner: draw a
'line through the erroneous data'and record correct data. Initial
-and date ~the correction, b.-'The. cognizant supervisor shall list any available substantiating information.-
- c. The individual ~who conducted the test:and the cognizant supervisor shall attest to the corrected data'by affixing their signature and the date.
- d. The station nucle'ar. safety and operating committee ~shall review the circumstances'and the chairman ~shall sign and date the form 1 when their review is completed.
- 0UESTION - 0.08 (1.00)
Which one of the following statements is.true of jumpers as controlled by ADM 14.1r Jumper.s (temporary modificati'ons)"?
- a. Temporary hose connections necessary to perform a test are defined as constituting a jumper if their use is described in the text or drawings of the FSAR.
l -b. - In order to control their user ADH 14.1 only permits the use of jumpers as described in either an MOP or an MMP.
- c. Jumpers not controlled by any-approved procedure shall only be used with the shift supervisor's prior knowledge and approval,
- d. For those jumpers that are installed by an approved procedurer a jumper los form shall be initiated and all pertinent data associated with the installation recorded.
l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) i I I t v--, --m- mm-.----,,...r.-.ew, ew...m,-r-.-um-- r w er ..,+.e. .,.ww--,mmy.,-,--.w%_mm.e,.re.m %. - way-e w ,m e %.--re v +-..- . e erv..--.r, s-w,=w.
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#g,, NUCLEAR REOULATORY CCMMISSION
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_ . ~ . _ _ _ _ . _ , l . 4- 4 4 8.' ' ADMINISTRATIVE PROCEDURES, CONDITIONS,'AND' LIMITATIONS PAGE' 23 -l 1, - t I
.i OUESTION- 8.09 (1.00)' t i L t According to Unit 2 tech specs,fvhich of the following statements regarding 12 the site fire brigade is correct?
j a. Shall be composed of'at least 10 members. .l l~ b. Shall.not include any of the minimum shift crew required by table i
- j. 6.2-1 (minimum shift manning). j F c. Shall be responsible for the control room command function until i
[: the fire emergency is secured. [ t
- d. May NOT be - less than the minimum requirement even to accomodate- 1
- j. unexpected absences. ;
t-t QUESTION 8.10 (1.00) ; Which one of the'following is a condition requiring stoppage of all work [ and immediate evacuation of containment according to the precautions.and ; j limitations in 1-OP-4.1, ' Controlling Procedure for Refueling *? t !. a. 'hi flux at shutdown
- alarm only if actuated during fuel movement. i j
b.- Loss of audible neutron count rate (< two tones per minute) after offloading 3/4 of the reactor core. I r
- c. The station evacuation alarm sounds. j
- d. Declaration of an " alert". [
GUESTION 0.11 (1.00) Which of the following is the principal' candidate for relief of the Interim .; Station Emergency Manager in the event of,an emergency at the station. l
- a. Assistant station manager (ops)
- b. Superintendent of operations ;
- c. Station manager f d.. Manager nuclear-operations and maintenance. '
L (***** CATEGORY OG CONTINUED ON NEXT PAGE *****) L i t i
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#{p2 reg % NUCLEAR REOULATORY CO MISSION
- g. *\1 g f, g REGION il
.* g 101 MARIETTA STREET N.W., SUITE 2000
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i.; B. -ADMINISTRATIVE PROCEDUREGi CONDITIONS, AND LIMITATIONS PAGE 24 .. i t f l j
.GUESTION 8 .' 12 . '(1.00) !
I If.an operator ~.is returning to shift after a three week vacation, he is ~ required to read and initial the logs.for the previous ______. (choose one) !
- a. 11 day' l
- b. 7 days ;
- c. 14 days ;
- d. 21' days i' 00ESTION B.13 (1.00) t Controlled leakage.as defined by Unit 1 Technical' Specifications refers to l
, s. Letdown flow. i i >
- b. Liquid radwaste release flow.
[
- c. Reactor coolant pump seal water flow. l d.- ' Excess letdown flow.
J , i e.- .S/G tube ' leakage. !
. OUESiION .8.14 (1.00)
Which one of the following reports must be completed in the event i of a gasoline spill (in the fuel depot) which caused a fire that lasted S minutes? I
- a. Licensee Event Report.
- b. Near Miss Report. l Plant Safety-Suggestion Report.
- c. ,
- d. Notification of Unusual Event. f l
5 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) ; I t
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$ $ 101 MA91ETTA STREET, N.W., SulTE 2000
- 8 ATLANTA GEORGIA 30323
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s 8.1 ADMINISTRATIVE PROCEDURESr CONDITIONS,'AND LIMI1ATIONS PAGE 25 GUESTION 0.15 (l'.00)
-During Unit 2:startup with the reactor about 2% power, you~ find'that~the-
~
PORV block valve is stuck ~open and incapableLof closing. Which of.the following is a1 CORRECT action (see attached-LCO)?'
- a. Continued operation is-allowed provided.the PORV is operable and power '
is' removed from'the block valve. b.-If action.b.-is satisfied you are' allowed to increase-power
'into Mode 1..(Block 1 valve cannot be restored operable).
- c. The PORV must be-closed, power removed from the solenoid valve ~and the block valve'must.be repaired prior to soins into Mode 1.
d'. Since the block valve is. incapable of closins, you must pr'oceed-to Hot Ltandby'within-the next 6 hours and ColdLShutdown'within.the'followins
- 30. hours.
GUESTION 8.16 (1.00) Which of the followins conditions' requires action accordins.to. Tech Specs in_less than 1 hour if in Mode ~2-on Unit-2?.
- a. The shutdown margin is 1.8.
b.'One train of. heat tracins on the BAST is inoperable.
- c. One Shutdown: rod fully. inserted.with the reactor critical.
- d. .Two of the three chargins pumps are inoperable.
QUESTION 0.17 ( .50) According to emergency plan implementins procedure, EPIP 5.03, durins a site emersency, personnel inside the protected area and unaccounted for
-shall be determined within ______ minutes of declaration of the emersency.
i (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) i a k j (
F p24% _ UNITED STATL' f,g ,o,, NUCLEAR RECULATCRY CCMZISSION y -c, g REGION il
- t 101 MARIETTA STREET, N.W., SUITE 2000 o, [ ATLANTA, GEORGIA 30323 3
( 4 3 i i f s 5
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8.- ADMINISTRATIVE PROCEDURESr. CONDITIONS, LAND LIMITATIONS PAGE 26-
. QUESTION 1 8.18 : ( 1. 00 ) -
Unit 2 Sis in' Mode 3'during o' Reactor: star' tup withLthe - fellowing deficiencies! One Main-SteamiTriP Valve - islinoperable and closed -
'One' Motor Driven 1 Aux. Feedwater Pump is inoperable See attached LCO's.
Which.one of the following actions most accurately details the allowances and/or limitations imposed by the Tech Specs in this instance?' a,1 Mode'3 must be maintained (Entry into Mode 4 acceptable)
- t. Startup-activities may continuel Mode 2 may be entered.but not exceeded,
- c. Startup and power-oper'ation'into Mode 1.may be accomplished provided Mode 1 action statement for MSTV met.
d.-Startup activities may continue into Mode 2-provided subsequent restoration of the MDAFW pump to operable status within 72 hours. QUESTION 8.'19 (1.00)
- During a Unit 2 startup with the reactor at 2% powerr one power range
- neutron flux monitoring channel is found to be inoperable. Which of the i following statements is correct? Refer to attached Tech Specs.
1 j a. Operation above 5% rated thermal pouer is not allowed until the q inoperable channel is repaired and declared operable, j
- b. If the inoperable channel is placed in a trip condition and the other i l
i three channels are operable, you must veri #y compliance with the l j shutdown margin requirements of Tech Specs within is minutes. l
- c. If the inoperable channel in placed in a tripped condition and the other three channels are operable, operation to 100% rated thermal j power may proceed only if all functioning units receiving an input from it are tripped (may not be bypassed).
j d. The only restriction on proceeding to 100% rated thermal power are , i that the inoperable channel be placed in a bypasced or tripped , condition (within time restrictions);in addition monitor OpTR. 9 e i l l (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****) i l l i L
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. NUCLEAR RE2ULATORY COZMISSION REGION il 5 g 101 MARIETTA STREET, N.W.,5UITE 2900
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v 0.- ADMINISTRATIVE PROCEDURES, CONDITIONS,_AND LIMITATIONS PAGE 27
- 00ESTION 8.20- (1.00)
- TheLunit 11 reactor coolant system pressure exceeds 2735 psig'when in mode 3. . -Accordina'to technical specifications, pressure must be restored within acceptable limits within what time frame:siven.below?
- a. 5 minutes.
- b. -15 minutes.
- c. 30 minutes.
- d. one hour.
QUESTION 8.21 (1.00) Which of'th'e following require activation-of both the TSC and OSC?
- a. Either an unusual event, alert, site area emergency or general emergency.
- b. .Only an alert, site' area emergency, or general. emergency.-
- c. Only a site area emergency or general emergency. -
d..'Only a-general emergency. QUESTION' 8.22 (2.00) Complete the.following table for Unit 2: MINIMUM SHIFT CREW COMPOSTION Hith Unit 1 in_ Mode 1,2,3 or.4 Number of Individuals Required to Fill' Position Position Mode'1,2,3 or 4 Mode 5 or 6 SS (SRO) ------a------ '
'------f------
SRO ------b------ ----- g- ---- RO ------c------ ------h------ i AD ------d------ ------i------ H STA ------e------ ------J------ (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
\
l 61 f 87 UNITED STATES l Af {o,, NUCLEAR RE!ULATORY COMMISSION l . , ,, o REGION il 3 g 101 MARIETTA STREET. N.W., SUITE 2000
~$ ATLANTA. GEORGIA 30323 '+4 . . . . . $
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d 8..' ADMINISTRATIVE PROCEDURESr CONDITIONS, AND LIMITATIONS PAGE-' 20 iOUESTION O '. 2 3 ' ( .50) Fill-in the BLANKi A' monthly surveillance requirement of Tech Specs may be extended up to _____ days without declarin3 the component inoperable due to the surveillancoPtesting not being performed.
.0UESTION 0.24 (1.00)
Hbat are the maximum allowable non-emergency whole body dose equivalent for an employee.with a completed NRC Form 4 for the following time periods?
- a. Iri any calendar quarter.
'b. In any calendar year, without corporate approval.
QUESTION 8.25 (1.00) Fill-in.the. BLANK with the appropriate Tech Spec definition. A _______shall be'the: injection of a simulated signal into-the channel as close to the primary sensor.as practicable to verify operability including alarm and or trip functions'.
- a. Channel Check
- b. Channel Calibration
- c. Channel' Functional Test
- d. Channel Instrument Check 00ESTION'.8.26 (1.00)
Tech Specs- defines Shutdown Marcin as ...
- Shutdown Marcin;shall be the instantaneous amount of reactivity by which the Reactor is suberitical or would be soberitical from its present condition assuming...."
STATE _the assumptions made for the plant conditions which complete the definition of Shutdown Margin. QUESTION 0.27 (1.00) To. prevent entering a technical specification action statement on Unit 1, the quadrant power tilt ratio shall not exceed ______ when reactor power is above 50%. If this limit is exceeded, then an extended temporary GPTR limit of.______ is allowed during efforts to restore OPTR to normal levels. (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
g>2reto UNITED STATES 4 ,o,, NUCLEAR REIULATORY COMMISSION E o REGION il
'S * \
101 MARIETTA STREET, N.W., SUITE 2000 ATLANTA, GEORGIA 30323
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8 .- ADMINISTRATIVE-PROCEDURES, CONDITIONS,.AND LIMITATIONS PAGE- 29 GUESTION' O.28 (1.00) If .an emergency condition- exist and security 'is controlling access to a. the control room HHERE do the. oncoming operators report-to?
- b. If-an emergency condition arises during shift turnover what ACTION do the oncoming shift crew take?
- GUESTION 'iB.29 ( .50) ,
Which individual (by position) determines VALVE Post-Maintenance Test requirements?
. 0UESTION 8.30 ( .50)
Fill-in the BLANK for the following! In accordance with 10 CFR 55 "if'a' licensee has not been actively perform- l ing the function of an Operator or Senior Operator for a period of _______ months,. or longer, he shall, prior to resuming activities licensed put svant to this part, demonstrate to the Commission that his knowledge and understanding of facility operations and administration are satisfactory.' QUESTION 0.31 (1.00) During Mode 1-operation of unit 2 it is found.that 2 of 4 channels for l Pressurizer Pressure high Reactor trip are inoperable due . r to a generic material deficiency (repair time 14 days). Using Tech Spec LCO's provided, determine what actions must be taken as a result of this failure? State specific LCO/ action steps which apply. t k (***** END OF CATECORY 08 *****) (************* END OF EXAMINATION ***************) t [ t
g UNITED STATES
/p >* E8I0f NUCLEAR RE!ULATORY COMMISSION o,^ $ f, o REG 40N il 5 a 101 MARIETT A STREET. N.W., SulTE 2000
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+
out)/(Energy in) 2 o o mg s e Vo t
- 1/2 at g = mc 2- ,
A=Ae'I
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KE = 1/2 mv a = (Vf - /3 )/t A = AN g PE = 89h Vf = V, + a t * = e/t a = an2/t1/2 = 0.693/t1/2 y , , .p A* nD 2
*1/2'II " EI*1") S )3 4
[(t1/2) * (*b)3 AE = 931 am
. m = V,yAo -D Q.= m,ah I*Iec Q = mCpat 6 = UAa T I = 1,e'"*
Pwr = wfah I = I, 10-*/D L TVL = 1.3/v P = P010 sur(t) HVL = -0.693/u t P = P,e / SUR = 26.06/T SCR = S/(1 - Kdf) CR, = S/(1 - Kdfx) SUR = 26o/1* + (s - o)T s CR)(1 - K df1) = CR 2 II - "#f2) T = (1*/s) + [(4 - o)fIol - M = 1/(1 - Kdf) = CR j/CR, T = s/(o - s) M = (1 - Kg f,)/(1 - Kdf1) T = (s - o)/(Io) SDM = (1 - Kdf)/Edf a = (Kgf-1)/Kgf = cKdf
/K df t= = 10- seconos I = 0.1 seconds
=((t*/(TKgf)] + [s df /(1 + T)]
I jd) =2 ,2 I d2 P = (reV)/(3 x 1010) Id i gd 22 2 I = eN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (f,,g) , Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010 aps 1 ga; . = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. I hp = 2.54 x 103 Stu/nr . Density = 62,4 lbT/ft3 1 rme = 3.41 x 100 5tu/hr Density = 1 gm/cW lin = 2.54 cm Heat of vaporization = 970 Stu/lem 'F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm .
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 773 ft-lbf I ft. H O2
= 0.4335 lbf/in.
e = 2.718
- Wolv.s.W/Ib Int u lpy,t e m tat,epy. Dis /lb a F 7e eg. , c, .. w., i, , , Te
., 4 a, a. s ., s -_
0 02 1075.5 1075.5 0.0000 2.1873 2.1873 32 gg gagggg g01402 3305 3205 3.00 1073.8 10762 0.0061 2.1706 2.1767 35 0.09991 OA1602 2948 2948 35 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 0.12165 0.01602 2446 2446 40 13A4 1068.1 1081.2 0 0262 2.1164 2.1426 45 0.14744 0.01402 2037.7 2037.8 45 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 30 0.17796 0.01602 1704A 17043 60 1207.6 28.06 1059.7 1067.7 0.0535 2.0391 2A946 80 0.2561 OA1603 1207.6 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 0.3629 0.01605 868.3 868.4 TO 633.3 48.04 1048.4 1096.4 0.0932 1.9426 2.0359 30 90 0.5068 0.01607 633.3 0 1115 2.0086 468.1 58.02 1042.7 1100.8 1A970 to 90 0.6981 0A1610 468.1 100 350.4 68.00 1037.1 1105.1 0.1295 1A530 1.9825 100 0.9492 0.01613 350.4 ISO 265.4 77.98 1031.4 1109.3 0.1472 13105 1.9577 110 1.2750 DA1617 265.4 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 1A927 OA1620 203.25 130 157.33 97.96 10193 1117A 0.1817 1.7295 1.9112 130 22230 0.01625 157.32 140 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 2A892 0.01629 122.98 150 97.07 117.95 1006.2 1126.1 0.2150 1.6536 1.8686 150 3.718 0.01634 97.05 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 13487 160 360 4.741 OA1640 62.06 137.97 996.2 1134.2 0.2473 1.5822 1A295 170 170 5.993 0.01645 62.04 50.22 148.00 990.2 1138.2 0.2631 1.5480 1A111 130 180 7.511 041651 50.21 ISO 40.96 158.04 984.1 1142.1 0.2787 1.514S 1.7934 190 9.340 0.01657 40.94 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 300 300 11.526 OA1664 33.62 210 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 14.123 0.01671 27.80 26.80 180.17 970.3 1150.5 0.3121 1A447 1.7568 212 212 14.696 0.01672 26.78 220 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 17.186 OA1678 23.13 230 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 20.779 0 21685 19.364 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 240 24.968 0A1693 16.304 250 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 29A25 0.01701 13302 228.76 938.6 1167.4 0.3819 1.3043 1.6862 360 260 35.427 0.01709 11.745 11.762 438.95 931.7 1170.6 0.3960 1.2769 1.6729 270 270 41.856 0.01718 10.042 10.060 8.644 249.17 924.6 1173A 0.4098 1.2501 1.6599 280 280 49.200 CA1726 8.627 290 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 230 57.550 OA1736 7.443 1.6351 300 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 300 67.005 OA1745 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 5.609 5.626 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 320 39.64 0A1766 4296 4.914 1.5892 340 3.770 3.788 311.3 378.8 1190.1 0.4902 1.0990 340 117.99 0.01787 1.0517 .1.5678 360 2.939 2.957 332.3 862.1 1194.4 0.5161 360 153Al 021811 0.5416 1A057 1.5473 380 2.317 2.335 353.6 844.5 1198.0 380 195.73 0.01836 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.8444 12630 420 396.9 306.2 1203.1 0.5915 0.9165 1.5080 420 305.78 0.01894 1.4808 1.4997 1.2169 419.0 785.4 1204.4 0.6161 02729 1.4890 440 440 381.54 0.01926 1.1976 1.4704 460 0.9942 441.5 763.2 1204.8 0.6405 0.8299 460 466.9 0.0196 0.9746 1.4516 480 0.7972 0A172 464.5 739.6 1204.1 0.6648 0.7871 480 566.2 0.0200 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 500 680.9 0.0204 0.6545 1.4146 520 0.5596 512.0 687.0 1199.0 0.7133 0.7013 523 812.5 0.0209 0.5386 1.3954 540 04651 536.8 657.5 1194.3 0.7378 0.6577 540 962.8 0.0215 0.4437 1.3757 560 0.3871 562.4 625.3 1187.7 0.7625 0.6132 SEO 1133.4 0.0221 0.3651 1.3550 580 0.3222 589.1 589.9 1179.0 0.7876 0.5673 580 1326.2 0.0228 0.2994 617.1 550.6 1167.7 0.8134 0.5196 1.3333 8100
" 600 1543.2 0.0236 0.2438 0.2675 0.2208 646.9 506.3 1153.2 0.8403 0.4689 1.3092 620 620 1786.9 C.0247 0.1962 1.2821 640 679.1 454.6 1133.7 0A666 0.4134 640 2059 9 04260 0.1543 0.1802 0.1443 714.9 392.1 1107.0 0A995 0.3502 1.2498 660 660 2365.7 0.0277 0.1166 1.2086 680 0.1112 758.5 310.1 1068.5 0.9365 0.2720 640 2708.6 0.0304 0.0808 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 700 3094.3 0 0366 0.0386 1.0612 705.o 0.0508 906.0 0 906.0 1.0612 0 705.5 3208.2 0.0508 0 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
A.3 -. - _ _ . __ _ A
9stume, it'/it Iathelry. Dis /lb tal'opy. Dt:/* a F gase8y.Dev/in h*P trop M ter toep Steem toter Loop Seeam tener tesem Pe*** F toler $seem Pn*e* Do 9* he he er % % *s ** % % *> *e t 0.01602 3302.4 3302.4 0 00 1075.5 1075.5 0 3.1872 2.1872 e 8021.3 eA886 e.0886 32.Gl8 0.30 35.023 0.01602 2945.5 2945.5 3 03 1073 8 1076 8 0 0061 2.1705 3.1786 323 1922.3 e.10 0.15 45.453 0 01602 2004.7 2004 7 13.50 1067.9 1081 4 0 0271 2.1140 2.1411 13.50 1025.7 0.15 0.20 $3.160 001603 1526.3 1526 3 21.22 1063 5 3084.7 0 0422 2 0776 2.1160 2122 10283 e.20 9.30 64 484 001604 1039 7 1039.7 32.54 1057.1 1089 7 0 0641 2.0165 2.0809 32.M 1032 0 e.30 9.40 72.869 0.01606 792.0 792.1 40.92 1052.4 1093.3 02799 3.9762 2A662 40.9;t 10341 9.40 0.5 79.586 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 24370 4742 1036 9 S.5 c.6 85.716 0 01609 540.0 540 1 53 25 1045 5 1098.7 01028 1.9186 22215 53.24 1038 7 9.6
+.
.* e 3.7 ' 90094 RO1610 466.93 .. 466 94 5830 3042 7. 410044 0J .. 1.4966.,3.0083 . 98,10,,1064 3 . e2.
0A 94.38 0.01611 411.67 411.69 62.39 1040 3 1102.6 0.1117 1A775 1.9970 6229 1041.7 c.B g.9 98.24 0.01612 368 41 368.43 66 24 1038.1 1104.3 0.1264 12606 1.9870 6624 1042.9 c.9 1.0 101.74 0.01614 333 59 333.60 69.73 1036.1 11058 0.1326 12455 1.9781 49.73 1044.1 1.0 2.0 126 07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 10513 2A 3.0 141 47 0.01630 118 71 118 73 109.42 1013.2 1122.6 0.2009 14854 12864 109.41 1056.7 3A 4.0 152.96 0.01636 90 63 90 64 120.92 1006 4 1127.3 0.2199 1A428 13626 120.90 1060.2 4.0 ' 6.0 162 24 0.01641 73.515 73.53 13020 1300.9 1131.1 0.2 M 9 1.6 W 4 13443 130.18 1063.1 5.0 0.01645 61.967 61.98 138 03 996.2 1134.2 0.2474 1.5820 12294 138A1 1065.4 6.0 6.0 170.05 7A 176 84 0.01649 53 634 53.65 144 83 992.1 1136 9 0.2581 1.5587 13168 144A1 1067.4 2.0 8.0 182 86 0.01653 47.328 47.35 150.87 988.5 1139.3 02676 1.6384 13060 15034 1069.2 a.0 , 9.0 188.27 0 01656 42.385 42.40 156.30 985.1 1141.4 0.2760 1.5204 1.7964 15628 1070.8 9.0 3 10 193.21 0.01659 38.404 38 42 161.26 982.1 1143.3 0.2836 1.5043 1.7879 161.23 1072A to 14.696 212.00 0.01672 26.782 26.80 180.17 970.3 1150.5 0.3121 1A447 1.7568 180.12 1077.6 14.896 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20 070 20 0ts7 196 27 960.1 1156.3 0.3358 1.3962 1.7320 196.21 1082A to 30 250.34 0.01701 13.7266 13 744 218 9 945.2 1164.1 0.3682 1.3313 1.0995 218 3 1087.9 30 40 267.25 0 01715 10 4794 10 497 236.1 . 933.6 1169.8 0.3921 1.2844 1.6765 2360 1092.1 40 0.01727 8 4967 8514 250.2
- 923.9 1174.1 0.4112 1.2474 J4585 250.1 1095.3 50 80 261.02 40 292.73 0.01738 7.1562 7.174 262.2 915.4 1177.6 0.4273 1.2167 1A440 262A 1098.0 60 70 302.93 0.01748 6 1875 6205 272.7 907A 1180.6 04411 1.1905 1A316 272.5 1100.2 70 80 312.04 0.01757 54536 5 471 232.1
- 900.9 1183 1 0.4534 1.1675 14208 281.9 1102.1 30 0 01766 4.8777 4.895 293 7 894.6 1185.3 0 4643 1.1470 14113 290.4 1103.7 90 90 320 26 100 327.82 0.01774 4.4133 4.431 258.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 120 341.27 0.01789 3 7097 3 728 312.6 877A 11904 0 4919 1A960 1.5879 312.2 1107.6 120 140 353 04 001803 3.2010 3 219 325 0 864.0 1193 0 0.5071 1.0681 1.5752 324 5 1109.6 140 160 363 55 0 01815 2.8155 2.834 336.1 859.0 1195.1 0.5205 1.0435 1.5641 335.5 1111.2 360 380 373 08 0 01827 2.5129 2.531 346.2 2 50.7 1196.9 0.5328 1.0215 1.5543 345.6 1112.5 180 200 35180- 0 01879 2.J689 2.287 355.5 842A 1198.3 0.5438 1.0016 1.5454 3543 2113.7 300 250 400 97 0 01865 1A245 13432 376.1 825 0 1201.1 0.5679 0 9585 1.5264 375.3 1115.8 250 300 417 3b 0 018E9 1.5233 1.5427 394 0 806 9 1202.9 0.5682 fl.9223 1.5105 392.9 1117.2 300 350 431.73 ,001913 1.3064 1.3255 409.8 794 2 1204 0 0 60M C8909 3.4968 400 6 1118 1 350 1.14162 1.1610 424.2 760 4 1204.6 0 6217 0 8630 1.4847 422.7 111E 7 400 400 444 60 0.0193 450 ab6 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 0.6360 04378 1.4738 4351 1118.9 450 0 90787 0 9276 449 5 755.1 1204 7 0.6490 0.8148 1.4639 447.7 11184 500 500 467 01 0 0199 553 476 92 00199 0 82183 0 9418 460.9 743.3 1204 3 06611 07936 1.4547 456.9 1118 6 550 0.74962 0.7698 471.7 732.0 1203 7 0.6723 0 7738 1.4461 469.5 111E. 2 800 600 48520 0 0201 700 703 .503 08 0 0205 0.63505 0 6556 491.6 710.2 1201.8 06928 07377 1.4304 488.9 1116 9 0 54809 0.5690 509.8 689 6 1199 4 0 7111 0.7051 1.4163 506.7 1115.2 000 830 514 21 0 0209 900 531 93 0 02i2 0 47965 0 5009 526 7 669 7 1196 4 0 7279 0.6753 1.4032 5232 1113.0 900 1000 5
- 4.5B 0.0216 0 42435 04460 542.6 650 4 1192.9 0.7434 0 6476 1.3910 53:16 1110.4 1000 0 4006 557.5 631.5 11891 07573 0.6216 1.3794 553.1 1107.5 1100 1100 555 2e 0.0720 0 37803 3200 J M7.19 0 0223 034013 0.362b 571.9 613.0 1184 8 0.7714 0.5969 1.36S3 566 9 1104.3 1200 0.3299 585.6 594.6 1180 2 01843 05733 1.3577 580.1 1100 9 1300 1300 57742 0 0227 030722 1400 557 07 0 0731 0 278/1 0 3018 598 8 576 5 1175 3 0.7966 05507 1.3474 592.9 1097.1 1400 1500 5 % 20 0 0235 02h372 0.2772 611.7 550 4 1170 1 0.8035 0 5253 1.3373 605 2 1093.1 1500 2000 635 80 0 02*,7 016760 0 1883 672.1 4662 1138.3 0 8621 0 4256 1.7881 662 6 10GS 6 2000 7313 361 6 1093 3 C 9139 0 3206 1.2345 718.5 1032.9 2500 2500 65d l2 0 02c.f 010209 01307 801 8 218 4 1070 3 0 9728 0.1891 1.1619 782 3 973.1 3000 3000 6'e5 33 0 0343 0 050/3 0 0850 906 0 0 9060 1.0612 0 1.0612 875.9 875.9 3700.2 '
32982 70147 0 0508 0 0 050d TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) A.4
S-27A 0683 pasasusiasm ocessun tosco samt o-c;os Gtsps tucLean sowsz equ stl:pupyr ulgy,g g 4 i j i 3 l o w t= E* 3 i i 1 l
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003TW8777 9 Tr-i l l . TYPICAL MOVABLE DETECTOR TRACE
- IN LOW LEVEL SETUP (Gamma Response)
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4-20-81 TA8LE 3.4.1 '
,g REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Maxiaum (a)(b)
Valve No. Allowable Leakace System Low Head Safety Injection to Cold Legs 83 5 5 gpm Loop 1 5 5 gpa
-SI-195
.-51-86 5 5 gpm Loop 2 5 5 gpa 197 89 5 5 gpa Loop 3 5 5 gpm
.-51-199 l 22 Footnotes:
(._ (a) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable. 4..
- 2. Leakage rates greater than 1.0 gpa but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpa by 50% or greater.
- 3. Leakage rates greater than 1.0 gpa but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the I rate determined by the previous test by an amount that reduces the margin between measured leakage rata and the maximum permissible l
' rate of 5.0 gpa by 50% or greater.
- 4. Leakage rates greater than 5.0 gpa are considered unacceptable.
(b) Minimum differential test pressure shall not be less than 150 psid. Z s t. / t s NORTH AhMA - UNIT 3/4 4-18a
~
8-21-80 REACTOR COOLANT SYSTEM
\
TABLE 3.4-1 REACTOR COOLANT SYSTEN PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 5I-85 High head safety injection to cold legs and SI-93 hot legs 5I-107 5I-119 MOV-2836 High head safety injection off charging MOV-2269A, B hesser
~.
MOV-2867C, 0 - Baron injection tank outlet valves
-SI-91 Low head safety injection to cold legs 99
-5I-105 5I-126 Low head safety injection to het legs
'3I-123 151 -5I-170 Accumulater discharge check valves 153 SI-185 168 .-51-187 MOV-2700 RHR system isolation valves MOV*2701 MOV-2720A, 8 MOV-2390 A, B, C & 0 Low head safety injection to cold legs and hot legs Ejq 24 9. 2-NORTH ANNA - UNIT 3/4 4-lSa 1
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, . . . g-.. . . . . . . .,4 ... . g. ... . . .. . . gg . . .__ m _t-g . . . . .. . .. .........._..g.. .... . . . . . . ..
. . ...m.. . . . . . . . . . . ~ . . . . . . . ,.....t...... . . . . . . . . . . .
.... .f... .. .
. = . . .. _ . . . . - -- . ..
. . . . . . . . : . .... ... . . .. *..**;;. . : . . . TE
! ='? .==* * . .=~.; . * .
.. . 3.==..=;
?.".....*..
;.r."..t. .
. . ..: . =r=.::. ... . .. ... .. . .-
. . .1.=.: .
. i g
...m.,.
....... t....
. . .. .3
. . t
... . . . . . .... . . . . ....t... -
.....w ...4 ... .
.......: . _ . ...8... .. . .. . . . . . . . . . . _ .. s.. . .
E._.._...... . _ . .... . . ... . . . ..... . . .. ... .i .. .... ... 0 2 4 . 6 8 10 12 . C033 uff2T (7T) Figurs 3.2-2 E(Z) - Notuslized Yq(Z) as a Function of Con Height NORTE ANNA - UNIT 2 3/4 2-8 - Amendment Ns. 28 .
. /
e
- Y Q/r .hb -
Y".' - *=. g u. . . ., . . . . . , .
2-1-85 3/4 LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS e 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 4
- 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervels, completion of the ACTION requirements is not required..
3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION . shall be initiated to place the unit in a MODE in which the Specification
,does not apply by placing it, as applicable, in: -
- 1. At least NOT STANDBY within 6 hours,
. 2. At least HOT SHUTDOWN within the next 6 hours, and
- 3. At least COLD SHUTDOWN within the following 24 hours. l Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requiremen'ts are stated in the individual specifications. This specification is not applicable in MODES 5 or 6.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the condit4ons of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. F,xceptions to these requirements are stated in the individual specifications. 3.0.5 When.a system, subsystem, train, enaponent or device is determined to be inoperable solely because its emergency power source is inoperable. or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s), and device (s) are OPERA 8LE, or likewise satisfy the ' requirements of this specification. Unless both conditions (1) and (2) are
. satisfied, the unit shall be placed in at least HOT STAND 8Y within 1 hour, in at least HOT SHUTDOWN within the next 6 hours, and in at least COLD SHUTDOWN within the following 30 hours. This specification is not applicable in MODES K 5 or 6.
NORTH ANNA - UNIT 2 3/4 0-1 Amendment No. 46 ,
8-21-80 f N. . EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,y, GREATER THAN 350*F LIMITING CONDITION FOR OPERATION
. 3.5.2 Two independent ECCS subsystems shall be OPERA 8LE with each subsystem comprised of:
- a. One OPERA 8LE centrifugal charging pump,
- b. One OPERA 8LE low head safety injection pump, .
- c. An OPERA 8LE flow path capable of transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on a safety injection signal or from the containment sump when suction is transferred during the recirculation phase of cperation.
APPLICA8ILITY: MODES 1, 2 and 3. ACTION:
- a. With one ECC5 subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT SHUTOOWN within the
, next 12 hours.
! b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to Trthe Commission pursuant to Specification 6.9.2 within 90 days describ-
,ing the circumstances of the actuation and the total accuelated
/ actuation cycles to date. The current value of the usage factor i for each affected safety injection nozzle shall be provided in this 7_,Special Report whenever its value exceeds 0.70.
- c. The provisions of Specification 3.0.4 are not applicable to Specifi-cations 3.5.2.a and 3.5.2.b for one hour following heatup above 340*F or prior to cooldown below 340*F.
l l SURVEILLANCE REQUIREMENTS l l 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: f a. At least once per 12 hours by verifying that the following valves i
, are in the indicated positions with power to the valve operators
- removed
l r f NORTH ANNA - UNIT 2 3/4 5-3 l l l _ . - - - . . _ _ . . _ - - . _--,,,_.__--,.m.._
l l 1
- 3/4.8 ELECTRIf.ALPOWERSYSTEMS 3/4.8.1 f..C. SOURCES I i
0PERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERA 8LE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system, and
- b. Two separate and independent diesel generators:
- 1. Each with a separate day tank containing a minimum of 750 gallons of fuel, and
(
- 2. A fuel storage system containing a minimum of 45,000 gallons of fuel, and
- 3. A separate fuel transfer pump.
APPLICA8ILITY: MODES 1, 2, 3 and 4. , ACTION:
- a. With one offsite circuit of 3.8.1.1.a inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. If either EDG has not been successfully tested within I the past 24 hours, demonstrate its OPERASILITY by performing Surveil-lance Requirspent 4.8.1.1.2.a.4 separately for each such EDG within
- 24 hours. Restore the offsite circuit to OPERABLE status within
- 72 hours or be in at least HOT STAND 8Y within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- b. With one diesel generator of 3.8.1.1.b inoperable, demonstrate the OPERA 81LITY of the A.C. offsite sources by perfoming Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter; and if the EDG became inoperable due to any cause other than preplanned preventative maintenance or testing, demonstrate the OPERASILITY of the remaining OPERABLE EDG by perfoming Surveil-lance Requirement 4.8.1.1.2.a.4 within 24 hours *; restore the diesel generator to OPERABLE status within 72 hours or be in at least HOT following STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the 30 hours.
' *This test is required to be completed regardless of when the inoperable EDG is restored to OPERABILITY.
NORTH ANNA - UNIT 2 3/4 8-1 Amendment No. 48
4-25-85 j i l l ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- c. With one offsite circuit and one diesel generator inoperable, demon-strate the OPERA 81LITY of the remaining A.C. sources by performing i
Surveillance Requirement 4.8.1.1.a within one hour and at least once per 8 hours thereafter; and if the EDG became inoperable due
- to any cause other than preplanned preventative maintenance or testing, demonstrate the OPERA 81LITY of the rJanining 0PERABLE EDG by performing Surveillance Requirement 4.8.1.1.2.a.4 within 8 hours *;
restore one of the inoperable sources to OPERA 8LE status within 12 hours or be in at least HOT STAN08Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore the other A.C. powersource(offsitecircuitordieselgenerator)toOPERA8LEstatus l in accordance with the provisions of Section 3.8.1.1 Action Statement a or b, as appropriate with the time requirement of that Action State-ment based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel.0PERA8ILITY per i Surveillance Requirement 4.8.1.1.2.a.4 performed under this Action Statement for an OPERABLE diesel or a restored to OPERABLE diesel satisfies the EDG test requirement of Action Statement a or b.
- d. With two of the required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by sequentially performing Surveillance Requirement 4.8.1.1.2.a.4 on both diesels within 8 hours, unless the diesel generators are already operating; restore one of the inoperable offsite sources to OPERABLE status within 24 hours or be
- in at least HOT STAND 8Y within the next 6 hours. Following restora-tion of one offsite source, follow Action Statement a with the time requirement of that Action Statement based on the time of initial loss of the comaining inoperable offsite A.C. circuit. A successful-test (s) of diesel 0PERA8ILITY per Surveillance Requirement 4.8.1.1.2.a.4 performed under this Action Statement for the OPERA 8LE diesels satisfies the EDG test requirement of Action Statement a.
e. With two of the above required diesel generators inoperable, demonstrate i the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.a within one hour and at least once per 8 hours thereafter; restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Following restoration of one diesel generator unit, follow Action Statement b
- with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable diesel generator. A successful test of diesel OPERABILITY per Surveillance Requirement 4.8.1.1.2.a.4 performed under this Action Statement for a restored to OPERABLE diesel satisfies the EDG test requirement of Action State-ment b.
J
*This test is required to be compl~eted regardless of when the inoperable EDG is restored to OPERA 8ILITY.
_ . N0fHANNA dNI{2__ - 3MR Amen h M L
l l 8-21-80 l
~
REACTOR COOLANT SYSTEM
. RELIEF VALVES j
LIMITING CONDITION FOR OPERATI0'N
. l 3.4.3.2 Two power relief valves (PORVs) and their associated block valves shall be OPERA 8LE. i APPt.ICABILITY: MODES 1, 2, and 3.
ACTION:
- a. With one or more PORV(s) inoperable, within 1 hour either restore the PORV(s) to OPERA 8LE status or close the associated block valve (s) and remove power from the block valve (s); othenvise, be in at least HOT STAN08Y within the next 6 hours and in COLD SHUT 00WN within the following 30 hours.
- b. With one or more block valve (s) inoperacle, within 1 hour either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STAN0BY within the next 6 hours and in COLD SHUT 00'4N within the following 30 hours.
6 SURVEILLANCE REOUIREMENTS 4.4.3.2.1 Each PORY shall be demonstrated OPERA 8LE:
- a. At least once per 31 days by performance of a CHANNEL FUNCTIONAL
- TEST, exclud+ng valve operation, and .
! b. At least once per 18 months by performance of a CHANNEL CALIBRATION. I 4.4.3.2.2 Each block valve shalli be demonstrated CPERABLE at least once per l , 92 days by operating the valve t>cugh one c:aplete cycle of full travel. l Y { l l - i
\
[ NORTH ANNA - UNIT 2 3/4 4-7a
8-21-80
? ~
/
PLANT SYSTEMS AUXILIARY FEEDWATEW $YSTEM LIMITING CONDITION 7UR OPERATION 3.7.1.2 At least three independent steam generator auxilfary feedwater pumps and associated flow paths shall be OPERA 8LE with:
- a. Two motor driven auxiliarf feedwater pumps, each capable of being powered from separate emergency busses, and
- b. One steam turbine driven auxiliary feedwater pump capable of being powered from an OPERA 8LE steam supply systas.
APPLICA8ILITY: M00ES 1, 2 and 3. ACTION:
- a. With one auxiliary feedwater pump inoperante, restore the recuired auxiliary feedwater pumps to a OPERA 8LE status within 72 hours or be in at least HOT SHUTDOWN within the following 6 hours.
. b. With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY N. within*6 hours and in HOT SHUTDOWN within the following 6' hours.
- c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to CPERA8LE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2 In addition to the requirements of Specification 4.0.5, each auxiliar/ feedwater pump shall be demonstrated OPERA 8LE:
- a. At least once per 31 days by:
- 1. Verifying that each motor driven pump develops a discharge pressure of greater than or equal to 1260 psig at a flow of greater than or equal to 53 gpa.
- 2. Verifying that the steam turbine driven pump develops a dis-charge pressure of greater than or equal to 1380 psig at a flow of greater than or equal to 35 gpa on recirculation flow. The
. provisions of Specification 4.0.4 are not applicable.
NORTH ANNA - UNIT 2 3/4 7-5
,--,,-,,,,,,-,.--17.,.,-r.,.ww-- -,,w. .---..----v-, , , ,4--, <.-,.. . .,%..- --- . _
5-21-80
\
PLANT SYSTEMS MAIN STEAM TRIP VALVES ; LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam trip valve shall be OPERABLE. APPLICA8ILITY: MODES 1, 2 and 3. i ACTIO_N: MODES 1 - With one main steam trip valve inoperable, POWER OPERATION may continua provided the inoperable valve is either restored to OPERA 8LE status or closed within 4 hours; otherwise, be in HOT SHUTDOWN within the next 12 hours. MODES 2 - With one sain steam trip valve inoperable, subsequent and 3 operation in MODES 1, 2 or 3 may proceed and the provisions of
~
specification 3.0.4 are not applicable provided the main steam trip valve is maintained closed; otherwise, be in HOT SHUTDOWN within the next 12 hours. _ SURVEILLANCE REQUIREMENTS , 4.7.1.5 Each main steam trip valve shall be demonstrated OPERABLE by verifying l full closure within 5 seconds when tasted pursuant to Specification 4.0.5. l l e NORTH ANNA UNIT 2 3/4 7-10
8-21-80 r i 3/4.3 INSTRUMENTATION
'3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION i LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERA 8LE with RESPONSE TIMES as shown in Table 3.3-2 APPLICA8ILITY: As shown in Table 3.3-1.
ACTION: As shown in Table 3.3-1. SURVFILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the N00ES and at the frequencies shown in Table 4.3-1. ( 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the proceeding 92 days. The total interlock function shall be demonstrated OPERABLE at i[f-least once per 18 montes during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.1.1.3 TheREACTORTRIPSYSTEMRESPONSEf1MEofeachreactortrip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that ?- both logic trains are tested at least once per 36 months and one channel Tw per function such that all channels are tasted at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1. I - 1 - ( . l . l f NORTH ANNA - UNIT 2 3/4 3-1 l l
. . _ _ ~ . .. _._...____..._i. _ . . _ _ _ . . .
h h TA8t[ 3.3-) 4
.l
(
- REACTOR TRIP SYSTIN,,1NSTRUMENTATION
~~'
MININIM e TOTAi.NO. CIWINELS CilANNELS APPLICABLE g IINICTI0llAL llDili 0F CilANNEtS 10 TRIP OPERAbg N00E5 ACTION
, ~
[ 1. Manual Reactor Trip 2 1 2 1, 2. ar.d * , 12
- 2. #
Power IIan08 Neutron Flux 4 2 3 1, 2 2 .
- 3. #
Power Ran08. Neutron Flux 8 4 2 3 1. 2 2 liigh Positive Rate
- 4. Power Range Neutron Flum, #
4 2 3 1, 2 2 High Negative Rate
- 5. Intermediate Range, Neutron Flux 2 1 2 1, 2, and
- 3 A 6. Source Range, Neutron Flum gg l A. Startup 2 1 2 2 . and
- 4
- 8. Shutdown 2 0 1 3, 4 and 5 5
- 7. Overtemperature AT Three loop Operation Two Loop Operation 3 2 2 1, 2 [
3 1** 2 1, 2 9 7 0 l' l 4 ., ", . e
^
% ,.e j
i
! fARLE 3.3-1 (Continueel) .
I 8
= , RfACIOR IRIP SYSTEM INSTRISTNTAll0N iil 3; MINilRSI
, g TOTAL N0. CHA100ELS CilANNELS APPLICABLE ] , itNICTIONAL initT Of CilANNf t S TO IRIP OPERASIE MDOES ACTION
'$ 8. Overpower Ai l -i Three Loop Operation 3 2 2 1, 2 /
N iwo Loop Operation 3 1** 2 1, 2 9
- 9. Pressurizer Pressure-Low e 3 2 2 1, 2 ./
l 10. Pressurizer Pressure--High 3 2 2 1, 2 / j 11. Pressurizer Water Level--High 3 2 2 1, 2 /
; 12. Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in I /.
l y (Above P-8) any oper- each oper-
- ating loop ating loop l 13. Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop 1 /
(Above P-7 and below P-8) . two oper- each oper-l ating loops ating loop i 14. Steam Generator Water 3/ loop 2/ loop in 2/ loop in 1, 2 / Level--Low-low any oper- each eter-i ating loops ating loop y
' n
- 15. Simun/Feedwater Flow 2/ loop-level 1/ loop-level 1/ loop 1, 2 / j, Mismatch and Low Steam asul coincident level asul c Generator Water level 2/ loop-flow with 2/ loop-flow mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop asul 1/ loop-fIow mismatch '
i
i:
"i II y TAntE 3.3-1 (Continued) g REACTOR 1 RIP SYSTEN T Sitt0ENTATION y -- m 4
E MINil658 TOTAL NO. CHANNELS CHANNELS APPLICA8tE -
! s__ . FUNCTIO 10AL INIIT OF CliANNEt $ TO 1 RIP OPERA 8tE N00ES ACTICII ti
" 16. Undervoltage-Reactor Coolant .
i Psamp Busses 3-l/ bus 2 2 1 / ; 2 # 17. Underfrequency-ReactorCoojant 3-1/ bus 2 2 1 7 ]l Pamp Busses
- 18. Turbine Trip
! A. Low Auto Stop 011 Pressure 3 2 2 1 [g j B. Turbine Stop Valve Closure 4 4 4 1 7
- 19. Safety injection Input y 1, 2 from ESF 2 1 2 1 8-i
- 20. Reactor Coolant Pamp Breaker i ! Position Trip A. Above P-8 1/ breaker 1 1/ breaker 1 10 j 8. Above P-7 1/hreaker 2 1/ breaker 1 11 per oper-l ating loop j 21. Reactor Trip Breakers 2 1 2 1, 2, and
- 1 l
1 22. Automatic Trip Logic 2 1 2 1, 2, and
- 1 O e I
i 1
/ *
- l w- .
1-7-82 f.
.', TABLE 3.3-1 (Continued)
..J' l,, ,. U . TABLE NOTATION
*/ ,l
Nith the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal. The channel (s) associated with the protective functions derived from t,he out of service Reactor Coolant Loop shall be placed in the tripped condition. i The provisions of Specification 3.0.4 are not applicable.
##Bigh voltage to detector any be de-energized above the P-6, (Block of Source *
- Range Reactor Trip), setpoint.
ACTION STATDENTS ACTION 1 - With'the number of chan==1s OPERABLE one less than re-quired by the Minimus Channels OPERABLE requirement, be in EDT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE. ACTION 2 - With the number of OPERABLE channels one less than the ' Total Number of Channels, STARTUP and POWER OPERATION'may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped' -
condition within 1 hour. i '
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel any be bypassed for up to 2 hours for surveillance testing of the redundant channel (s) per Specification 4.3.1.1.1.
i
- c. Either THERMAL POWER is restricted to n 75% of RATED THERMAL and the Power Range, Neutron Flux trip setpoint is reduced to 4 85% of RATED THERMAL POWER within 4 hours; or, the-QUADRANT POWER TILT RATIO is monitored at least once per 12 hours.
- d. The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using 22 the moveable incere detectors to confirm that the normal-1 sed symmetric power distribution, obtained from 2 sacs of 4 synastric thimble locations or a full-core flux amp, is consistent with the indicated QUADRANT POWER
/~ TILT RATIO at least once per 12 hours.
NORTH ANNA - UNIT 2 3/4 3-5 Amendment No. 13
l 8-21-80 l ( TA8LE 3.3-1 (Continued) ; j ACTION 3 - With the number of channels OPERA 8LE one less than re-quired by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. . Below the P-6, (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to CPERA8LE status prior to increasing THERMAL POWER above the P-6 Setpoint.
- b. Above the P-6, (Block of Source Range Reactor Trip) setpoint, but below 55 of RATED THERMAL POWER, restore the inoperasle channel to OPERA 8LE status prior to increasing THERMAL POWER above 55 of RATED 114ERMAL POWER. .
- c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.
ACTION 4 - With the number of channels OPERA 8LE one less than re-quired by the Minf aum Channels OPERA 8LE requirement and with the THERMAL PCWER level:
- a. Below the P-6, (81ock of Source Range Reactor Trip) setpoint,
. restore the inoperable channel to OPERA 8LE status prior to increasing THERMAL POWER aeove the P-6 Setpoint.
e.-
- b. Above the P-6, (Block of Source Range ' Reactor Trip) setpoint, operation any continue.
ACTION 5 - With the nummer of channels 08 ERA 8LE one less than re-quired by the Minimum Channels OPERA 8LE requirement, verify compliance with the 5HilTD0hN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicanle, within 1 hour and at least ones per 12 hours thereaftar. ACTION 5 - Not applicable. , f ACTION 7 - With the number of OPERA 8LE channels one less than the Total Neber of Channels, STARTUP and POWER OPERATION any proceed until performance of the next required CHAMMEL FUNCTIONAL TEST provided the inoperable channel l 1s placed in the tripped condition within 1 hour. ACTION 8 - , Not applicable O NORTH ANNA - UNIT 2 3/4 3-6 .
~ - - - - - - - - - - - - - . . - _ _. , _ . ___ _ _ _ .
7 3, . t
- 5. THEORY OF NUCLEAR POWER' PLANT' OPERATION, FLUIDS, AND PnGE 30 ANSWERS -- NORTH ANNA 182 -86/06/23-CASTO, C ANSWER 5.01 (2.00) s
- a. Thot = 630 deg T s a t 9 2235 p s i g << 7' / #'3v = 653 (+/- 1) E1.03 subcooling =q{deg.F (+/- 1)
- b. For subcooled conditions the reading would indicata 10 deg. (0.25) more subcooling than actually exist (0.25).
For superheated conditions the reading would indicate 10 deg. (0 25) less superheat than actually exist (0.25). REFERENCE North Anna Core Cooling Monitor Obj. 2. K/A Comp T/C (3.0/3.1) 002-000-A1.04 (3.9/4.1) fc 7,f k -S
- 7.5 ANSWER 5.02 ( tT7?s )
- a. High relative fluy tauses a greater reactivity change due to CRW being p'oportional to fluy tip/ flux avg. therefore, the higher the relative flux the greater the change. (0.333 for area /0.5 for Exp.)
- b. larger ,ffect for the miodle - due to a b <,o r p t .4. o n of neutrons which have a high probabilitv of cauning fission. Whereas control rods at the edge absorp neutrons which'have a high probability of leakage.
- c. 41 has higher worth. When inserted ti depresses.the flux around itself, this increases the flux in other regions, when 11 2 is inserted the flux has been depressed therefore its worth is lower (than~its worth in an unrodded core).
REFERENCE North Anna ROP pp. 6.11,6.12,6.19 Obj. B M/A 001-000-KS.02 (2.9/3.4) ANSWEP 5.03 (1.00) SQ = f.A'5 /' f t? TI DNBR and RCS overprecrure >- g4 p,1, fr , 5 < ; r g
- 1 l*J q. 7aj. v REFERENCE 7- / >' 'f 4 c t North Anna I & C Sec.I. Prze Press. Cntrl.
K/A 002-000-K5.8 (3.4/3.9) a
>1 H 2% UNITED STATES d'
. g< h-o NUCLEAR RECULATORY COMMISSION f, REGION il 5 $ 101 MARIETTA STREET. N.W., SulTE 2900
*
- ATLANTA, GEORGIA 30323
%,...../
i I l l l l l t 4 ( i v ,, . . . , - ._,,..,~,.v.. -y....y ,,---._m._-,_m.--__.,y.-_ . _ - - _ _ _ _ . . . .. , -_ -- __.. _ - . . _ __- __ _ _ -_____.. __ - __ _a
V - ,
~
~
J 5 .1 _ THEORY OF; NUCLEAR POWER PLANTf0PERATION, FLUIDS, AND ' PAGE 31' ANSWERS'-- NORTH ANNA ~182 -86'/06/23-CASTOR C
-ANSWER 5~04
. (1.50)
- l. Potential- large- reduction :in cladding ability. to .transf erf heat (0.5) -
result - fuel pellet and clad heatup,(0.25).
- 2. . Blockage of1 coolant flow passages between fuel rods (0.5) - result-an additional reduction in the core's. heat transfer capability, causing fuel'heatup (0.25).
REFERENCE-LNorth Anna NCRODP-95.2
~K/A 000-011-EK3.13 (3.8/4.2)
ANSWER 5.05 (1.50) a.:Without-RCP-driven. pressurizer spray no adequate means of mixing loop-pressurizer exist. (0.75)
- b. Provides reasonable assurance that even a fairly rapid temperature drop will.not cause problems with a loss of core shutdown margin. (0.75)-
ANSWER 5.06 (2.00)
- a. Subcooling is based on core exit T/Cs or hot les RTD readings. During natural circulation the mass of metal in-the head can' retain heat and keep local temperatures above saturation. The temperature indicators would not reflect this local saturated condition. -(1.0)
- b. Pressuriner level decreases because the pressurizer pressure increase will compress the vesse'l void and force water out of the pressurizer.
-(1-.0) )
REFERENCE G.P. Heat Transfer and-FF Pp 355-358
-{caf}
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'5. 1 THEORY OF~ NUCLEAR ~ POWER. PLANT OPERATION, FLUIDS, AND PAGE--.32
'~~~7TUER566YUdsiC5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- NORTH ANNA 1&2- -86/06/23-CAGTO, C.
= ANSWER' 5.07- (1.00)
-a. .The maximum neutron population (equilibrium) is reached when K^n approaches'=ero. (0.5) lb, Keff (0,5)
REFERENE,E-
' North Anna: ROP 7.21 ANSWER 5.08' (2.00)
- a. Establishment of rtverse flow (0.5 ea.)
;b. Less pressure drop across core (less total' core; flow). . _
c..No lanser transferring-heat into S/G causing cooldown/ contraction of water / steam in S/G.
Ed . Increased delta T.in operatins loop-(with approximately-constant _Teold). REFERENCE Porth Anna::
'003-000-K5.02-(2.8/3.2)
-K5.03 (3.1/3.5)
-K5.04-(3.2/3.5) 000-015-EK 1.02'(-3.7/4.1)
ANSWER. 5.09l (1.00) TheLability to deliver _a certain number of. amperes for a specific number of
- -hours before'the' cell voltage drops to a specific minimum value.
l REFERENCE . North' Anna NCRODP 90.3 Sec. I e t t g l-i L l-s l l-
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(-- 5.. THEORY OF NUCLEAR POWER PLANT'OPERATIONr.FLUIDSr LAND PAGE 33 ANSWERS -- NORTH ANNA l'A2 -86/06/23-CAST 0r:C
' ANSWER 5.10 (2.00) a._The depressions are seen because the local burnup in the vicinity of the Grid (0.5) and thus the amount of gamma producing fission fragments is smaller (0.5). ,y7
- b. The samma mean free' path is larger than the neutron mean free path (er5).
thus the detector sees moreareggrder g the Samma flux and the detail c?e?s k $ r Y Ye Y Y s S $ ,337 REFERENCE North Anna: MCD Sec. III Obj. GrC
~K/A EPE-007-EK1.01 (2.4/2.9)
ANSWER 5.11 (2.00) During a LOCA the core may be blown dry and reflooded by the CLAs. Since the upper half will be reflooded last (1.0) mere restrictive limits are placed._on'the upper half of the core during normal' operations (1.0). REFERENCE G. P. HT & FF p. 249 ANSWER 5.12 (1.00)- d REFERENCE LDFNP RANKINE CYCLE LP,P.5,7-8 North Anna Thermo Sec. VI Obj. H. p. 6.21 ANSWER 5.13 (1.00) d. REFERENCE North Anna! NCRODP 06.2 Sec. 4
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- 5. THEORY OF NUCLEAR. POWER PLANT' OPERATION, FLUIDSr AND P' AGE 34
____ _________e____________________________ q ANSWERS -- NORTH ANNA 1&2 -86/06/23'CASTO, C ANSWER 5.14 (1.00) d
-REFERENCE FNP FSAR CH 15 Westinghouse Accident and Transient Analysis, pp 4.21 EPE-009; EK3.11(4.4/4.5)
ANSWER 5.15 (1.00) d REFERENCE Westinghouse Reactor Physics, pp. I-5.63 - 76 HDR, Reactor Theory, Sessions 38 and 39 DPC,. Fundamentals of Nuclear Reactor Engineeriner Section VI 001/000; K5.33(3.2/3.5)
/
23 ANSWER 5.16 ( M )) a.-increases-due to increasing non-condensable content.
- b. increases-temperature out of the coolers drop lowering Tsat, this de-creases Psat.
- c. decreases-increases chilled water tenperature this increases Tsat, increasing Psat.
(0.333 ine/dec - 0.5 reason) REFERENCE North Anna NCRODP-91.2 K/A 022-000-A1.02 (3.6/3.8) i i (
1 t4H4 UNITED STATES 9(g, NUCLEAR RE20LATORY COMMISSION '[, , g REGION il 5 g 101 MARIETTA STREET, N.W., SulTE 29CO
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 35 ANSWERS -- NORTH ANNA 182 -86/06/23-CASTOR C ANSWER 5.17 (1.50)
- a. Decreases (0.5)
- b. Increases (0.5)
- c. Increases (0.5)
REFERENCE Turkey Point, Thermal-Hydraulic Principles and Applications, pp. 10 61 North Anna: NCRODP-83 Sec. VIII App. A: NPSH(3.4/3.6) 19 # ANSWER 5.18 p
- a. Less negative (0.5)
- b. Less negative -fr (0,5) e.
meantive a d v (0.5) REFERENCE Turkey Point, Reactor Core Control, Chapter 5, Fig. SNP-RF-9 Nrth Anna NCRODP 06.2 Sec. V. 004/000; K5.06(3.0/3.3) ANSWER 5.19 (1.50)
- a. 5 (0.5 ea.)
- b. 3 (4 with explanation)
- c. 4 REFERENCE G.P. HT & FF pp. 220 - 230 ANSWER 5.20 (1.50)
- a. 1 (0.5 ea)
- b. 2
- c. 1 REFERENCE G. P. HT & FF pp. 99 & 218
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 36 ANSWERS -- iiORTH ANN A 1&2 -96/06/23-CASTO- C ANSWER 5.21 (1.00)
The interaction with Hydrogen rtoms in the water molecules. REFERENCE pp4 ho/ / 7 (- ec 7*/4 "O EIH L-RG-602, p9 4 , BSEP: 02-0G-A, pr. 10 -11 5 North Anna
- 86.1 sec. III, p. 4.6
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 37 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 6,01 (2.50)
.a. This would have no effect on Letdown system operation. (0,5)
- b. T ere would be NO auto. pump start (0.5) Letdown would isolate {CAF}(0.5)
- c. -it is Jurt possible CO.5] interlock exist with 15J7 (0.5) f f sa t-% / ,
REF RENCE 7 #" b North Anna: 0.P. 8.1 & Attachments ANSWER 6.02 (2.50)
- a. Provides a quick reduction in the amount of load on the bus in a de-graded voltage si uation. (1.0)
- b. Degraded voltageII+CCAF}(0.5), Containment Depressurization Signal (0.5).
- c. The transfer from the control room to the Diesel Room is blocked while.the EDG is loaded onto the bus.(0.5)
REFERENCE North Anna! NCRODP 90.1 See II & II EDG. ANSWER 6.03 (2.00)
- a. This range is invalid with any RCPs operating, in this condition there is one RCP running (0.5). The range also has an Hydraulic Isolator wtich is i the alarm condition.(0.5).
L r' 0' %/ '/f ~ E
- b. Range (DP1)Q(0.5)
- c. Temperature of Impulse lines [0.25 ea.]
RCS temperature Ilide range pressure d/p cell located out side of containment REFERENCE North Anna: NCRODP 93.3
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 38 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 6.04 (2.00)
- a. This failure would have no effect on the start sequence.
- b. This failure would have no effect on the start sequence due to redundent air start valves,
- c. NORMAL the EDG would not start. [0.5 ea.]
EMERGENCY the EDG drop-out relay would reset and the EDG would start. REFERENCE North Anna NCRODP 90.1 EDG l ANSWER 6.05 (1.00) l REFERENCE NAPS ESF RSS Lescon Plen. I l ANSWER 6.06 (1.00) a+ REFERENCE North Anna! NCRODP 92.2 SWS ANSWER 6.07 (1.00) c. REFERENCE ! North Anna: NCRODP 91.1, ESF-RSS j ANS4ER 6.00 (1.00) b. REFERENCE Notth Anna NCRODP 93.10 l l t l
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTA1 ION PAGE 39 '
ANSWEPS -- NORTH ANNA 1&2 -86/06/23-CAST 0r C ANSWER 6.09 (1.00) i
- d. l REFERENCE ,
North Anna NCRODP 93.11 ! r ANSWER 6.10 (3.00) l
- a. rods out E0.25] Tref will be max so Tave/ Tref mismatch and N1/ Turbine power mismatch will both give a rods out signal E0.75]
- b. rods in [0.25] Loop 1 Tave increases and auctioneered high Tave also :
increases. Tave/ Tref mismatch gives a rods in signal [0.75] l
- c. rods out [0.25] the power mismatch circuit of the reactor control unit responds oniy to rate of change of deviation between turbine and nuclear Power but rod motion will occur due to the Tave - Tref difference. CO.75]
REFERENCE ; Topic 6 Lesson 2 Fig. RS-5 and pp 55, 19, 20 l North Anna NCRODP 93.5 ANSWER 6.11 (3.00) I
- a. SI e. ph A .
- b. SI f. ph B !d gq i
- c. SI r'
- d. S-I,% At_ cc 7
( I REFERENCE North Anna NCRODP 77 ANSWER 6.12 (2.00)
- a. Manuel signal calling for rod movement Rods move IN. E0.5 ea.]
- b. Tave-Tref deviation calling for rod novement however with 'D' selected rods DO NOT move. / /h,1.
- c. Manual signal calling for rod movement Rods =n",
g _a q ,-[FN/" nT. d' /
- d. Tave-Tref deviation calling for rod movement htWever 228 steps blocks movement, rods 00 e move. Ecaf] /2 c/5, a., g / e 4 l '1T $ 475
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 40 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C REFERENCE North Anna NCRODP 93.5 ANSWER 6.13 (1.00)
This indicates that the RCS is 25 deg. superheated. REFERENCE North Anna NCRODP 93.4 ANSWEg glo (1.00) MOV 1700 can be positioned by the operator as needed. MOL 1701 will not be able to open when te uired. [0.5 ea.] REFERENCE J/
# # *) b #I 2 North Anna NCRODP 88.2 M/A 005-000-K4.07 (3.2/3.5)
ANSWER 6.15 (1.00) RWST lo-level -
.I recirc signal present (SI signal) << 7' / i 'd ' d.i' One recire isolation valve has (4 02G ea.)
close} REFERENCE North Anna NCRODP 91.1 ANSWER 6.16 (1.00) To prevent accidental opening (0.5) which would result in Accumulator dumping into the RHR system (0.5). REFERENCE North Anna
- NCR ODP--8 0. 2 p . 1.12
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- 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 41 l
ANSWERS -- NORTH ANNA la2 -86/06/23-CAST 07 C
/.D ,
ANSWER 6.17 (W) f y3 " f y t, ' ~
- a. By ETCV-100] a temperature control valve E.ZSJ which isolates at 136 deg F. E.25]
- b. A flow switch downstream of TCV-100 E.25] a low flow alarn light at the aux. bldg. ground floor on SG chemical panel. E.25]. -
- c. H 6 "'H- M actiVIT7 u;i L. e- l u.- .
.f IWJ3 eM 2^/I' REFERENCE North Anna NCRODP PRHS 7 - / ' < . Y4-ANSWER 6.18 vuu . .v ,ew, (2.00) a -can ~
g c5 f e.,{ [ hD2
- a. Ecaf2 g) ),7 w te qw
- b. To ensure a flow path in the event of a station blackou*,.
REFERENCE North Anna NCRODP 26 l 1
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- 7. PROCEDURES NORMALr ADNORMAL, EMERGENCY AND PAGE 42 0 -----RABIdb6818Ab-80HTReb-----------------------
ANSWERS -- NORTH ANNA 1&2 -96/06/23-CASTO, C ANSWER 7.01 (1.00) d REFERENCE
-MNS EP/2/A/5000/16 3 CNS EP/1//A/5000/2F3r P.7.
NAPS.1-FRP-I.3A, p.3. ANSWER 7.02 (1.00)
~
b. REFERENCE North Anna 1-ES-3.1 P. 1
- ANSWER- 7.03 (1.00)
C REFERENCE NAPS 1-OP-4.1r p.22. 004/020; PWG-8 (3.6/4.4)
- ANSWER 7.04 (1.00)
Dy opening the auxiliary spray valve. REFERENCE MNS EP/2/A/5000/1.2, p.5. LCNS EP/1/A/5000/1A1, p.4. NAPS 1-ES-0.3, p.5. 002/020; PWG-12 (3.7/3.7) ANSWER 7.05 (1.00) A.
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- 7. PROCEDURES - NORMAtr ADNORMALr EMERGENCY AND PAGE 43
- ---~~~~~~~~~~~~~~~~~~~~~
~~~~R E5i5tBEiEEL 55NTR5t ,
ANSWERS -- NORTH ANNA 182 -86/06/23-CASTO, C REFERENCE North Anna TS 3.2.1 ANSWER 7.06 ( .50) FALSE REFERENCE . North Anna AR--1 1a-4 ANSWER 7.07 (1.00) Y,pr fvd 7 " / 5' ~ REFERENCE North Anna OP 5.2 p. 5 ANSWER 7.08 (1.00) 4 Tavg-Tref mismatch of 1.5 deg F REFERENCE North Anna AP 1.6 ANSWER 7.09 (1.00) By lining up the refueling purification system to recirculate water from the reactor cavity to -Lhe spent fuel pit using the spent fuel coolers for cooldown. REFERENCE NAPS, 1-AP-11, p.9. 005/000; A2.03 (2.0/3.1)
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-7. : PROCEDURES:- NORMAL,.ABNORMAli EMERGENCY'AND ~ PAGE" 44
~~~~R5656LU55E5['"EUUTR6[~~~~~~~~~~~~~~~~~~~~~~~~
iANSWERS --'.' NORTH ANNA-1&2 -
-86/06/23-CASTO, C j'.
- h. :
l ANSWER .7.10 (1.00) .! t:' t l-- The same' tool which is used to-unlatch the drive shafts - ! ~ t I l. REFERENCE E s ' I r \ -* f NAPS 11-OP-4.1, p.41. 1
/
034/000; K6.01 (2.1/3.0) i AMSHER 7.11 (1.00) ' l~. ;1. RHSTl suction open t l' 2..VCT suction closed ..' CO.125 ea.3 3 '. Norma 1' charging closed ,, ;
- 4. Letdown isolation closed
REFERENCE:
North-Anna EP-0 p.3
-L
' ANSWER L7.12 (1.00) a 7
-If-dosimeter 3/4 of scale I L
off scale. f dropped-' i any malfunction- v, ! f lost-3
= REFERENCE. *
' North' Anna GET handbook
% b
= ANSWER -7 13 (1 00)' 4 j
. Verify Charsing/SI flow (0.125) l
- and-(0.125) RCS pressure < 1230 psis'(0.125i'C1680 psigJ (0.125) .i
-or CCW, lost.(0.5) i i
REFERENCE. North Anna Foldout page for 2-EP-0 t L L
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45
~
~~~~R5656LUU5E5L UUUTR5L-----~~~~~~~~~--~~~~~~~~ ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 7.14 (?.00) For 0.5 points each. (a) 750 mr/qtr (b) 2750 mr/qtr REFERENCE NAPS Radiation Protection Manual, p.2.3-7. PWG-15' Radeon Knowledge (3,4/3.9) ANSWER 7.15 (1.00) From the service water system E0.53 by the use of a changeover suitch on the ventilation panel CO.5]. REFERENCE North Anna AP 35 p. 6 5.1.6 ANSWER 7.16 (1.00) upon a loss of all charging pumps due to a fire on the affected tait. REFERENCE North Anna AP 40.1 ANSWER 7.17 (1.00) The operator should remain in ECA-0.0 since the FRPs are written on the premise that at least one E-bus is energized. REFERENCE West. background info. for ECA-0.0 l
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'7.. PROCEDURES - NORMAL,'ADNORMALi ' EMERGENCY AND PAGE 46
-~~~------------~~~~~~--
~~~~R E5i5E65fEEE E6sTR5L ANSWERS -- NORTH ANNA 1&2 -86/06/23-CAST 01- C ANSWER ~ 7.18 (1.00) 1.:SI flow is a significant contributor to any cold les' temperature decrease , EO . 53 -
- 2. It can also.be a significant contributor to an overpressure condition-if the RCS is. intact. . E'O . 53 REFERENCE West. background info. for FRP-P'.1 ANSWER' 7.19 (1.50) a'. The step may be marked N/A and initialed. CO.5]
b'.' Submit PT critique sheet E0.53 enter procedure'into the Action Statement Status Los E0.53. REFERENCE
-North Anna Admin 11.2 p 10 ANSWER 7 20 -(1.00)
-This Prevents an erroneous flux penality from the delta flux program.
~
REFERENCE North.-Anna AP-4 p.. 11 ANSWER 7.21 (1.00) Heaters will not operate due'to Undervoltase trip CO.S] to reset place the control' switches to 0FF-and then to desired position E O . 5]; REFERENCE North AnnaEAP-10^.1 p. 4
. ANSWER- '7.22 (1.00)
LAT the last 5 steps to the: fully inserted position E0.5] jos the control ~ rods in E0.53
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- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 47
~~~~RAUi6E56iEAL E5NTRUE~~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C REFERENCE North Anna OP-58 ANSWER 7.23 (1.50) No E0.53 The EDG logic has locked out the fire protection system heat detector [1.03 REFERENCE North Anna OP 6.1 p. 5 ANSWER 7.24 (2.00)
- a. Upper right-hand corner of the tag. [1.03
- b. It is given to the person performing the work E1.03 ws L(. sce r f t c.svs%.</o C-e/ r e < /s t, f l M [$-
REFERENCE North Anna ADM 14.0 ANSWER 7.25 (1.50) Red circled [0.53 and explanation of why it was and what corrective actions have been taken EO.53 entered into the Remarks Section E0.53 REFERENCE North Anna ADM 19.1 ANSWER 7.26 (1.00) An increase in letdown flow has been verified. REFERENCE NAPS, 1-OP-3.4, p. 13. 004/010; K5.05 (3.0/4.2)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48
~~~~RA515E55isAE 55 sir 5E-~~~~~~~~~~~~~~--~~~~~~~
ANSWERS -- NORTH ANNA 1&2 -86/06/23-CAST 0, C ANSWER 7.27 (1.00) Whenever the P l ant temperature is above 350 deg. F tregardless of the rod mechanism status. /0c T_ - 3 $~d T J,{ ( 4 p /h ' 5 e^ ett c ~ $ oaf REFERENCE NAPS 1-OP-1.1. 001/000; A2.01 (3.1/3.7) ANSWER 7.28 (1.10) QF=1 for samma 100(45/60)(1)=75 REFERENCE 10 CFR 20. PWG-15: Radeon Knowledge (3.4/3.9)-
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS -- NORTH ANNA 1&2 -86/06/23-CASTO, C ANSWER 0.01 (1.00) 239.2 REFERENCE North Anna U-2 TS Table 3.4.1 ANSWER 8.02 (1.00) c.
REFERENCE North Anna TS 3.0.5 ANSWER 0.03 (1.00) (a) REFERENCE MA U2 TS 3.1.3.2 014/0003 PWG-5 (2.9/3.9) ANSWER 8.04 (1.00) a. REFERENCE NA U1 TS 3.5.4.1 006/050; PWG-5 (3.2/4.3) ANSWER 8.05 (1.00) 1QO den. F Hni* 1 and 2 jr. c 5 (, oIh/2 REFERENCE A 7 / pfc f .* [/gA NA U1&2 TS 3.4.9.1 / fig 002/020; PWG-5 (2.9/4.1)
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8.: ' ADMINISTRATIVE PROCEDURES, C0NDITIONS, AIOD LIMITATIONS: PAGE 50 ANSWERS -- NORTH ANNA.182- -86/06/23-CASTO, C l
- ANSWER' r 8.06- (1.00)
(b)- REFERENCE NA U1 & 2 TS 3.1.1.5 002/0205-PWG-5 (2.9/4.1)
, ANSWER -8.07 (1.00)
-(a)
REFERENCE' ADM 5.7,;p 1. PWG-26: Loss / Records:(3.3/3.6) ANSWER 8.08 (1.00) (c) REFERENCE ADM 14.1, p 3. PNG-14: Tassing (3.6/4.0) ANSWER 8.09 (1.00)' (b)
-REFERENCE NA U2 TS 6.2.2 PWG-19: Fire Brigade (3.4/4.2)
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- 8. ADMINISTRATIVE PROCEDURESrLCONDITIONSr AND LIMITATIONS PAGE' 51-ANSWERS'-- NORTH ANNA'1&2 -86/06/23-CASTO,'.C AHSWER 8.10 -(1.00) _
~(c)-
~l REFERENCE.
1-OP-4.ir p 15. EPE-036;.EK3.01'(3.1/3.7) - ANSWER 8.11. (1.00)
.(c)
REFERENCE-
~NA EP 5.2.1.1 PNG-36: E-Plan'(2.9/4.7) ' ANSWER 8.12 (1.00)
(b)- REFERENCE' ADM 19.3, p 1. PWG-26: Loss _(3.3/3.6) ANSWER 8.13. (1.00)- (c) REFERENCE NA U1 TS'1.7
'004/020;.PNG-5 (2.9/4.1)
ANSWER 8.14 (1.00) b. [.
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r D. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 APSWERS -- NORTH ANNA 182 ~B6/06/23-CASTO, C REFERENCE North Anna ADM 20.16 ANSWER 8.15 (1.00) d.- REFERENCE North Anna TS 3.4.2. ANSWER 8.16 (1.00) C. REFERENCE North Anna TS 3/4 1-1,-9,-12,-22 ANSWER 8.17 ( .50) 30 REFERENCE North Anna EPIP5.03 p. 1 ANSWER 8.18 (1.00) a. REFERENCE St Lucie Tech Spec North Anna TS Sec. 5
'061/000; PWG-5(3.3/4.1) & 039/000; PWG-5(3.1/3.7)
ANSWER 8.17 (1.00) (d ) REFERENCE SL, TS, pp 3-2,4,5. North Anna TS 3.3.1.1 015/020; PWG-5(2.8/3.9)
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~8o- " ADMINISTRATIVE PROCEDURES, CONDITIONSr AND' LIMITATIONS PAGE 15 3 ANSWERS -- NORTli ANN A1182- -86/06/23-CASTO, C ANSWER- 8.20 ,(1.00)-
(a)' REFERENCE NA U1 TS 2.1.2 010/000; PWG-5c(2.9/4.1) ANSHER 8.21: (1.00) (b) REFERENCE EPIP-3.02 and 3.03. PWG-36 .E-Plan (2.9/4.7) ANSWER 8.22 (2.00)'
- a. l' f. 1
- b. 1' 3 none
- c. 2 h. 1 E0.2 lea.]
'd. 2 1. 1 e .~- 1 f. none REFERENCE' St. Lucie Tech-Specs Sec. 6
. North Annna TS sec.~6' PWG-23(Shift Staffing / Activities) (2.8/3.5)
ANSWER- 0.23 ( .50) 7.5 REFERENCE ~ St. Lucie Tech Spec. 4.0.2 North Anna TS 4'.0.2
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIDMS, AND LIMITATIONS PAGE- 54 ANSWERS -- NORTH ANNA 182 -86/06/23-CASTO, C ANSWER B.24 (1.00)
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- b. 5 Rem REFERENCE St. Lucie 10 CFR 20 North Anna ANSWER 8.25 (1.00)
C. REFERENCE St. Lucie TS definitiont North' Anna TS 1.5 ANSWER 0.26 (1.00) All full length control element assemblies shutdown and res, are fully inserted E0.5] except for the single assembly of highest reactivity. worth which is assumed to be fully withdrawn.CO.53 REFERENCE St. Lucie Tech Spec def. 1.29 North Anr'a TS def. ANSWER' O.27 (1.00) 1.02 (+.5 ea) 1.09 REFERENCE NA U1 TS 3.2.4 001/050; PWG-5 (2.9/4.3)
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- 8.. ADMINISTRATIVE PROCEDURESr1 CONDITIONS,-AND LIMITATIONS
.PAGE 55
' ANSWERS - NORTH ANNA'182 -86/06/23-CAST 0r.C-L
-ANSWER '8.28 (1.00)
~
;a. Report to the~onsite-OSC b.-Leave'the control room unless otherwise directed CO.5 ea.]
REFERENCE North Anna ADM ops.-
~ ANSWER 8.'29 ( .50)
Shift Supervisor REFERENCE Nov.th Anna ADH.11.3 ANSWER-' 8.30 ( .50)-
-4/ months REFERENCE' 10 CFR 55 PWG-23(Shift Staffins/ Activities) (2.8/3.5)
- ANSWER 8.31 (1.00)
LCO 3.0.3 applies REFERENCE North Anna LCO-3.0.3. !~ f 4 s _, ....am. -~ .. ,, . . , _
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