ML20147H334
| ML20147H334 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/09/1987 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20147H305 | List: |
| References | |
| ND3VPN:5249, NUDOCS 8803080456 | |
| Download: ML20147H334 (74) | |
Text
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)Vg 75 4 Telephone (412) 393-6000
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Shipp+ngport, PA 15077 6
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December.9, 1987 ND3VPN:5249' Mr. Robert M. Gallo 0perations Branch
-Division of Reactor Safety
-U. S. Nuclear Regulatory Commission
-Region 1 , ,
631 Park Avenue ' King of Prussia, PA .19406 v
Reference:
Beaver Valley Power Station, Unit #2 i Docket 50-412, License NPF-73 ! License Examination Report l
Dear Mr. Gallo:
t
.Please find enclosed comments generated by our Training Section-
. associated with the written examination administered December 2, 1987 at our facility.
If you have any questions concerning this report please contact , Mr. T. W. Burns at (412) 393-5751. ! t Very truly yours ! Y b
/
J. D. Sieber Vice President Nuclear [ i JDS/TWB: cal ! Enclosure cc: Central File (2) L 1 8803000456 800226 L e PDR ADOCK 05000412 i V PDR ; l J' . _ ~ . _ _
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" j FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 1.02 (2.00)
Beaver Valley Unit 2 has a total reactor coolant flow rate of 9.566 E7 lbm/hr at a hot leg temperature of 606 degrees F and a cold leg temperature of 546 degrees F. Steam generator pressure is 770 psig. Assume the feedwater entering the steam generators is at saturation. Calculate the total feedwater flow rate in Ibm /hr. Neglect all other heat inputs and losses. Assume the ' specific' heat capacity-of the reactor coolant is 1.0. State all other assumptions. SHOW ALL WORK. ANSWER 1.02 (2.00) Qrx - mic[T(hot)) - T(cold)] where al - mass flow rate of RCS (0.25 for formula) Qsg - m2[h(stm) - h(feed)) where m2 - mass flow rate of feedwater (0.25 for formula) (Qrx - Qsg, therefore,) mic(T(hot) - T(cold)) - m2[h(stm) h(feed)) (0.25 for relationship) therefore, m2 niciT(hot) - T(cold)1 h(stm) h(feed) from steam tables, h(sta) - 1200 BTU /lbm (0.25) h(feed) - 507 BTU /lbm (0.25) therefore, m2 9.566 E7 (1) (606 - 546)1bm 1200 - 507 hr therefore, m2 - 8.28 E6 lbm/hr (0.75) REFERENCE BV EXAM BANK, 1 32 BV LP-TMO-5, LO. 3 BV LP-TMO-3, LO. 7 3.3 3.1 4.0 KA VALUE(S) 002000K501 002000K511 193003K125 ...(KA'S) R0 EXAM REVIEV (12-2-87) 0 L,02 Point values of 0.25 are established in the answer key for stating the formulae for Orx and Qsg. The relationship between these two formulae can be established without needing to state each individually. Therefore, it is requested that grader discretion be used and that an overall understanding of the answer development be used as a guide instead of a particular response (i.e., the two formulae).
,e 3 si FACILITY REVIEW OF WRITTEN EXAMINATION-OUESTION 1.04 (2.50)- !
1
.. Explain HOW and WHY reactor power AND Tave respond during and for ONE (1) hour .I following ONE (1) minute of Emergency Boration at the following power levbts until equilibrium conditions are attained. No other operator actions are "
taken. Assume rod control is in manual and no manual or' automatic _ protection signals lare generated,
- a. 1004 equilibrium rated power (1.50)
- b. Critical at 1 E 8 amps following a refueling outage (1.00) 4 ANSWER 1.04- (2.50).
- a. Power decreases initially (0.10) due.to negative reactivity added by
- boration (0.20), but will subsequently increase (to ' match secondary power) ,
(0.20) due to (positive reactivity added from) decreasing Tave (0.20). Tave decreasee initially (0.20) due to primary to secondary power mismatch (0.20) and continues to decrease until after boration is stopped (0.20) and will stabilize when reactor power equals secondary power (0.20). g
- b. Power decreases initially (0.10) due to (negative reactivity added by) boration (0.20), and continues to decrease (0.10) until it stabilizes at a
[
' level caused by equilibrium subcritical multiplication (0.20). l Tave does not change (0.20) because it is independent of power (at power 1cvels < POAH). .(0.20) t REFERENCE BV EXAM BANK, 5 28 I BV LP-RT-7, LO. 3,4 -
3.8 KA VALUE(S) 192008K120 ...(KA'S) f RO EXAM REVIEW (12-2-87) 1.04.b. The answer reficcts that the examinee must state that Tavg does not change due to its power independence. Other answers which indicate Tavg independent of power, (i.e. , steam dump setting), should also be . . accepted. (Refer to attached copy of BV Exam Bank, 5 28). ' I, l
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Quahn, l.04. b . .
- 5. Theory of Nuclear Power Plant Ooeration. Fluids. and Thermodynamics
( - 5-28 a. Explain the response of reactor power and Tave during and after 2 minutes of Emergency Boration at 100% power. Assume' rod control is in manual,
- b. Explain the response of reactor power and Tave after 2 minutes of Emergency Boration at 10E.8 amps and no load Tave.
5 ANSWER: 5-28 a. Power decreases initially due to the boron addition. The primary to secondary mismatch causes Tave to decrease. The decrease in Tave inserts positive reactivity and
. restores reactor power to the same as initial power level,
- b. Tave does not change due to the boration. M dema=44byttheremeumewf ymythose* esse *twswesem>euer eeestag79 After thh initial transient, power decreases at a negative 1/3 DPM rate.
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TEN EXAMINATION FACILITY REVIEW OF WRIT SIX dissimilar i it is' (1.50) vable parameter, nameensure the DN 1.06 controls to QUESTION is- not a directlyand/or obser Since the DNBR operator monitors parameters the
- not violated.
(1.50) f ANSWER 1.06 each)' 0.25 points RCS temperature (any 6 at 2 Rx power 4. RCS pressure 6, QPTR
- 1. ing, overlap, alignment)
- 3. RCS flow
- 5. AFD Rod Position (sequenc CASE-BY CASE BASIS) 7 (CONSIDER OTHERS ON CH. 7, P. 17,18 H25 REFERENCEVALLEY THERMODYNAMICS, B BEAVER V ALLEY TECH, . SPECS .,
KA VALUE(S) BEAVER 3.4 . . (KA' S) BV LP THO-7, LO.11 2.9 193008K105 001000G006 s as 871 alignment) is given a items ing, overlap, d that each of the rep by it RO EXAM REVIEW Rod Position (sequenc (12-2 It is
- an individual an requeste answer.
- 1. Of acceptable can be considered asanswer ses: of "rod pos parenthesis singlee individual respon Therefore, the h the following t re rod sequencing properproper bank overlap rod alignment 7.
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c , ,;8(e - FACILITY REVIEW OF URITTEN EXAMINAllQH QUESTION 2.06 .(1.00)
.Given the following methods /pathslof secondary heat _ removal:- -
- 1. One rasidual heat, release. valve, HCV-104.
2.. One:SG atmospheric steam dump valve, PCV-101A
- 3. J0ne condenser steam dump valve, TCV-106H s a. Which ONE method / path has the greatest capacity for heat removal?
(0.50)
.b. .;Which ONE method / path can be controlled from the Alternate Shutdown Panel? (0.50)
ANSWER 2.06 (1.00)
- a. 1.- (0.50)
- b. 2. (0.50)
REFERENCE BVPS~2LP.SQS 21.1, PP. 5-8, 17 BVPS - 2LP.SQS 21.1, ELO 4,15 3.3 3.1- 3.6 2.9 KA VALUE(S) 039000K102. 039000K100 041020A202 041020G009 (KA'S) RO EXAM REVIEW (12-2-87) H 2.06.a The question asks for the one method / path of greatest heat removal
'; capacity. The choices are:
l- 1. Residual Heat Release Valve l
- 2. SG Atmospheric Dump Valve
- 3. Condenser Steam Dump Valve The answer found in the answer key is: 1. Residual Heat Release Valve. This is the incorrect response. The correct answer is: 3.
Condenser Steam Dump Valve. Included is a copy of the Interoffice Correspondence, dated 8/5/86, which lists the capacities of the Residual Heat Release Valve at various pressures. The capacity of the Condenaer Steam Dump Valve, as per 0.M. 21, has a maximum value of 890,000 lbm/hr. Therefore, in comparison, the Condenser Steam Dump Valve has a larger capacity for heat renoval. (Refer to attached copy of OM 2.21.1, p.10, 11; interoffice correspondence dated 8/5/86)
~
QoesW 2.06.s _ B.V.P.S. - 0.M. 2.21.1 MAJOR COMPONENTS Condenser Steam Bypass Valves [2HSS-TCV106A, 106B, 106C, 106D, 106E, 106F, 106G, 106H, 106J, 106K, 106L, 106M, 106N, 106P, 10601, [2 MSS-PCV106A, 106B, 106C.1 s Eighteen 8-inch steam dump valves are located on the south side of the main condenser neck, nine on each half. The steam dump valves are designed to . pass up to 907, of full load steam flow. Main steam pressure is always on I the stearn dump valve inlets. To prevent moisture collection in the valve inlets and steam c' ump. lines, the inlets of each steam dump are piped to a drain line. Fifteen of the valves are designated temperature control valves [2 MSS-TCV106A through 106Q). These valves are blocked or locked out by Lo-Lo Tavg. The remaining three valves are designated cooldown valves [2 MSS- > PCV106A,B,C). The Lo-Lo Tavg interlock may be bypassed for these three valves to allow cooldown of the reactor coolant system. l
- The eighteen steam dump valves, are divided into four banks for control, l The first and third banks consist of five valves each, Banks two and four l have four valves ea h.
I The steam dump valves are Copes-Vulcan. eight inch, D100-160-3 reverse acting tandem trim valves equipped with direct acting top works and a direct acting Moore positioner. Increasing air pressure pushes the valve stem down to open the valve. Decreasing air pressure allows the spring to ' lift the valve stem, closing the valve. The positioner uses a booster relay for faster valve response. Diaphragm operating air is routed through four solenoid valves. The first two solenoid valves control the Lo Lo Tuvg interlock, Train A and B. The third solenoid valve acts as the arming solenoid for condenser available, Rx trip, load rejection, and steam pressure signals. The fourth solenoid is the trip open or modulate solenoid. l d acher s tessWueMa lv e =1 s =me ch an i c e l l y-res t edet e+ 1e rrwerteuma stroke-to prevenscamsifegie valve vflow ofe-greater-when'890ie00@ This limits the consequences of a stuck open valve. Each steam , dump valve discharge is piped into the condenser neck to discharge at an elevation just above the sixth point heater bottom. The l . steam is prevented from impinging the sixth point heaters by a deflector which directs the steam out, up, and down from the* heater. The steam dump l discharges are staggered the length and width of the condenser to distribute the aeat load, l L 10 7 ISSUL 1 REV 2
h oj . o . !b-.. ' B.V.P.S. - 0.M. 2.21.1 T HAJOR COMPONENTS
!as?an Data
[2 MSS-TCV106Al (Typical for all) Type' 8" D100-160-3 6 6 Pressure (Inlet / Outlet), PSIG 1085/-15 Temperature, F 556 , Action, air-to-open Fail position (air / electricity) Closed / Closed The steam dump system has two automatic modes of operation, steam pressure mode and Tavg mode. The operational a, ode is operator selected by the Steam Dump Control Mode Selector Switch on the benchboard. In steam pressure mode, only the first two banks of valves are operational and they modulate to maintain the steam pressure setpoint set by the operator, using the l benchboard mounted steam pressure controller. In Tavg mode, two steam dump controllers are availabis. The reactor trip controller operates the steam dump valves to restore no load Tavg following a reactor trip. Only the , first two banks of valves are operational after a reactor trip. The load rejection contro11~re operates all four banks of valves for large load rejections and the first two banks for small load rejections, to restore Tavg to program value. ( All 18 valves trip closed if Tavg reaches Lo-Lo Tavg. If it is desired to cooldown the reactor plant, the Lo-Lo Tavg interlock may be manually defeated for the three cooldown valves only. 31nce the Lo-Lo Tavg interlock is dual train, two Steam Dump Control Interlock Selector Switches are provided, one for each train. The. Steam Dump Control Interlock Sele.ctor Switches are also used for manually blocking the steam dump control system. All 18 steam dump valves are blocked when the condenser is not available. To be available, the condenser must have sufficient vacuum and at least one cooling tower pump running. (' Load rejection is sensed by turbine first stage pressure. First stage pressure transmiter [2 MSS *PT447 sends a signal to bistables [PC447A] and [PC447B]. These bistables trip on rate of change of first stage pressure. [PC447A) trips on a rapid reduction in firstjstage pressure equivalent to a loss of load between 15% and 50%. [PC447B] trips on a 50% load rejection. ThebistablesaredesignedtolatchONsince!therateofchangesignalwill disappear as soon as first stage pressure reaches its new value. [PC447A] unblocks or arms the first and second bank of valves. [PC447B] unblocks or arms the third and fourth bank of valves. Both arming signals are negated if the condenser is unavailable or 1( Tavg reaches Lo-Lo setpoint. [PC447A and B) are reset by momentarily placing the Control Mode Selector , switch tn RESET. The switch spring returns to TAVG. L l i 11 : ISSUE 1 REY 2
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. FACILITY REVIEW OF VRITTEN EXAMINATION l
OUESTION 3.02 (1.50) ' Indicate how the affected OT delta T and OP delta T SETPOINTS will INITIALLY -i change (INCREASES. DECREASES, or NO CHANCE) if the following events occur with reactor power at 50%. Consider OP delta T SETPOINT and OT' delta-T SETPOINT for each event separately,
- a. Pressurizer pressure decreases from 2235 psig to 2150 psig.
(0.50)
- b. N-41 lower detector fails low. (0.50)
- c. The narrow-range loop 3 T-hot RTD output fails low. (0.50)
ANSWER 3.02' (1.50)
- a. OT delta-T setpoint DECREASES (0.25)
OP delta-T setpoint NO CHANCE (0.25) *
- b. OT delta T setpoint DECREASES (0.25)
OP delta-T setpoint DECREASES (0.25)
- c. OT delta-T setpoint INCREASES (0.25)
OP delta-T setpoint INCREASES (0.25) REFERENCE BVPS TECHNICAL SPECIFICATION, TABLE 2.2.1 BVPS 2LP SQS-1.1, ELO 7 3.1 2.9 KA VALUE(S) 01200A205 012000K611 (KA'S) i RO EXAM REVIEW (12-2 87) 3.02.b The answer for OP delta T is incorrect. The answer key states that the OP delta T setpoint will decrease as a result of N-41 failing low. However, the delta flux input to the OP delta T setpoint is set ' to zero (for all delta-I) and will not be affected by the N-41 lower ; detector failing low. Therefore, the correct response is no change. (Refer to attached copy of T.S 2.2.1, p. 2 9,10) [ 3.02.c The answer for OP delta-T is incorrect. The answer key states that ! the OP delta setpoint will increase as a result of the narrow range E i loop 3 T-hot RTD output failing low. However, the setpoint is > unaffected by a decreasing average temperature and/or temperatures s , 576.2*F. As verified by the OP delta T setpoint formula of Technical { Specification 2.2.1 (p. 2 9,10), K5 is O for decreasing average ' temperature and K6 is O for T $ T" (T" - 576.2'F). Therefore, the ' correct answer is no change. (Refer to attached copy of T.S. 2.2.1, !
- p. 2-9,10) l ,
I s F y ,, v-r- --,m --
w. J J s TABLE 2.2-1 (Continued) h085 OA 3.02.b. -
- h. REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 9 .
, NOTATION (Continued)
<: _ _ - _ s
# NOTE 3: f. OVERMWENTT)
- - - n S
O 1+135 $ AT, M4 - KS 1+15 (1 + 1 6) 5 T-K 6 U 1+156 - T9 7
~
Where: AT = Measured AT by RTD Manifold Instrumentation; 4 1+t5 1 = lead-lag compensator on measured AT; . 1+1 2S - - - - '-- ' ' " * - ' - - - - - -
=
17, r2 Time constants utilized in lead-lag compensator for AT, ty = 8 s,1, 2 = 3s; 7 1 = Lag compensator on measured AT; 1+T53 r = 3 Time constant utilized in the lag compensator for AT, r3 = 0 s;
= Indicated ai at RATED THERMAL POWER; AT, K
4
= 1.0781; K = 0.02/*F for increasing aver ge temperature and 0 for decreasing average temperature; 5
tS 7 = The function generated by the rate-lag compensator for T '9 dynamic compensation; T+ ty 5 . 1 = 7 Time constant utilized in rate-lag compensator for T,yg,17 = 10 s; -
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TABLE 2.2-1 (Continued) M - E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 9 NOTATION (Continued) <= 1 = f Lag compensator on measured T,,9; m 1+1S6 .
= '_ r 6 Time constant utilized in the measured T,,g lag compensator, 16 = 0 s; z
K 6
= 0.0012/'F for T > T" and K6 = 0 fof T 5 T"; .
T = Average Temperature, 'F; T" = Indicated T,,g at RATED THERMAL POWER (Calibration temperature for AT instrumentation,1576.2*F); 5 = Laplace transform operator, s 3; and
${.ill. ~ ~.". =H_ 02a11MI U NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.6% of AT span.
NOTE 5: The sensor error for temperature is 1.72% and 0.73% of span for pressure. NOTE 6: The sensor error for steam flow is 1.0%, for feedwater flow is 1.0%, and for steam pressure is 0.83% of span. .
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- C-TABLE 2.2-1(Continued]
E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)
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k_ NOTE 3:.fDfE.W.O.WE.R_ _T 7 m 0 1 t 1+13 5 I ob 2 z ~ Where: AT = Measured AT by RTD Manifold Instrumentation; 1+15 1 = Lead-lag compensator on measured AT; I+152 , , , , , , , , . _ . . , , , , , , , ,
=
17, t2 Time constants utilized in lead-lag compensator for AT, ty = 8 s ,12 = 3s; 7 1 = Lag compensator on measured AT; 1+1 35
=
1 3 Time constant utilized in the lag compensator for AT, T3 = 0 s; AT, = Indicated AT at RATED THERMAL POWER; K 4
= 1.0781; .
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IS 7 = T+t 75 The function generated by the rate-lag compensator for T**9 dynamic compensation; 1 = 7 Time constant utilized in rate-lag compensator for T,yg, 17 = 10 s; -
^ r '
m . TABLE 2.2-1 (Continued) QuesNovs 3.07.c. R - E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 9 NOTATION (Continued)
.c 1 =
f I*Ib6 Lag compensator on measured T,yg; O - T = 6 Time constant utilized in the measured T,yg lag compensator, r6 = 0 s; Z ]if
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= X0.0012/*F- for,T">f f"[eadh{0"for Ti?]
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bIh a _m =}f; Averaigleeigwra Q ;( { g l'~ ' { = 7:Iridicites T_'N'POWI[(CilibratTon']@~tFyffk2 13strumentatiea,1916;_2*Dd S = Laplace transform operator, s 2; and f 2(al) = 0 for all AI.
$ NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.6% of AT span.
NOTE 5: The sensor error for temperature is 1.72% and 0.73% of span for pressure. NOTE 6: The sensor error for steam flow is 1.0%, for feedwater flow is 1.0%, and for steam pressure is 0.83% of span. . e a 6
FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION 3.04 (2.00) Tha plant is operating at 80% turbine load with all control systems in automatic. A reactor protection system surveillance test of Train A is about to commence.
- a. Which reactor trip bypass breaker will have to be racked in and closed to prevent a reactor trip during the test? (0.50)
- b. List TWO (2) annunciators provided in the control room that will alert the operator when the reactor trip bypass breaker is closed. (1.00)
- c. How will the Reactor Protection System respond if a technician attempts to rack in and close both reactor trip bypass breakers concurrently? (0.50)
ANSWER 3.04 (2.00)
- a. B (0.50)
- b. REACTOR PROTECTION SYSTEM TRAIN A(B) TROUBLE SAFETY SYSTEM TRAIN A(B) INOPERABLE (any 2 of first 3 @ 0.50)
REACTOR TRIP BYPASS BKR A(B) RACKED IN/ CLOSED Ceneral Warning alarm (accept for half credit)
- c. Both reactor trip bypass breakers will trip. (0.50)
REFERENCE BVPS OM-2.1.1, P. 3 BVPS OM 2,1.4, P. AAF1, AAll, ABY1 BVPS 2LP-SQS-1.2, ELO 12,14,16 4.0 3,2 2.8 KA VALUE(S) 012000A307 012000K406 012000K408 (KA'S) RO EXAM REVIEW (12-2-87) 3.04.c The answer given, both reactor trip bypass breakers will trip, is only a portion of the correct response. The correct answer to the question is that an automatic reactor trip will result (i.e., two general warnings occurring simultaneously in Train A and Train B). Referring to OM 2.01.1, p. 43, 44, a list of inputs to the "Protection System Train A(B) Trouble" (general warning alarm), includes "7. either bypass breaker closed. " However, it further states that "if trouble in both trains should develop simultaneously, the reactor will be tripped automatically by the alarm system (general warning circuitry). Therefore, if an attempt to rack in and close both reactor trip bypass breakers concurrently is cade, an automatic reactor trip will result. (Refer to attached copy of OM 2.01.1, p. 43, 44) k
B.V.P.S. - 0.M. 2.01.1 INSTRUMENTATION AND CONTROL ( rear panel of the Central Board Display. This test checks the ability of the circuits within the Control Board Display to signal the applicable warnings to the operator. While performing this test the non-urgent alarm will be eliminated but an urgent alarm will be created (CR 19 flashes on each central control card). Also the GW LED flashes, step 0 LED (RB) is lit on every display card, and ROD DEVIATION LED (CR20) is lit on each Centrl Control Card. Rod Error Test This test checks the ability of the circuit within the display to signal the applicable warnings to the operator. Rod error codes (all l's) are manually presented at the output of the display I/O Card by means of the ETA and ETB pushbutton on the rear panel of the Control Board Display. While performing this test a non-urgent alarm will be created and RPI rod-at-bottom on RPIZ or more codes-at-bottom conditions. Also the step 0 LED (RB) is lit on every display card, the GW LED flashes on every display card and the DATA A or B FAILURE LED (CR17) flashes on each central control card. Simulated Data Transmission Tests for Normal and Error Codes may be done while the system is not in normal operation. In theses test the detector / encoder card outpn are inhibited. The normal position or error codes are applied .o each I/O Data Cabinet I/O Card (A021) by means of thu switches on the associated test / monitor card (A101). 3 Alarm System Two annunciators are provided in th'e control room labeled: i% -
. B4 These annunciators are not operated through the multiplexing scheme but are signaled ~ direct from the alarm system in the trains. {ThIaiinun'ciators"aia'6piraied by]tb{foI16Eing~failuresf
~
t 6pWrat!81ff?*{---i
- 1. Loss of either of two 48 volt DC power supplies
- 2. Loss of either of two 15 volt DC power supplies
- 3. Any printed circuit card not properly inserted
- 4. Input Error Inhibit switch in.the IhTIBIT position f
( 43 ISSUE 1 REV 1
B.V.P.S. - 0.M. 2.01.1 l INSTRUMERTATION AND C0tfrROL ! f
- 5. Logic A, Permissives, c r Memor switch not in 0FF position.
- 6. Slave relay tester Mode Selector switch in the TEST position 1
- 8. Multiplexing inhibited
- 9. Blown ground return fuse.
- 10. Mode selector switch not in the OPERATE position on the output relay test panel. f l
( l Loss of one of four 120 volt AC vital instrumentation busses is monitored b the li htin of multi le status la on the affggte' f channel. theToad ' {Cir6dit M Ir~ the alarm system are l'cated o on part of the semi-automatic tester card. An alarm relay, shown in Figure 1-40 , energized when none of conditions ( this ),g l l
~ .1h: :n se=
b.
. - - - ~._ i., egoactor. Another contact on the a arm re ay activates a Control Room annunciator.
(
\,
Overvoltage protection is supplied on the 48 volt and 15 volt DC power supplies. The device is a silicon controlled rectifier that clamps the output by providing an ar'tificial load. One key will be provided for all locks on train A d,bors. A different key will be provided to operate all train B locks. l Power Distribution ; Train A and train B both receive power from the four 120 volt AC vital instrumentation busses. The channel I through IV busses enter their respective input cabinet compartcents through fuses in , the compartments. In the input compartments, the busses are used to operate relays driven by external contacts. Two of the four k busses are run through line noise filters at the rear of the input compartment into the DC power supplies in the logic cabinet as shown in Figure 1-41. In train A, busses I snd II feed the power supplies .ind bus I feeds the slave relays. In train B, busses III and IV feed the power supplies and bus IV feeds the slave relays. Separate feeds are brought into tha output cabinet for the slave relays to avoid running unfiltered lines through the logic cabinet. The two 48 volt DC and 15 volt DC power supplies in one train are auctioneered to form one 48 and one 15 volt DC bus. A zero volt bus or circuit common bus is formed by. connecting the (-)48 and 44 ISSUE 1 REV 1
FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 3.06 (2.00) ; The plant is operating at 704 turbine load with the 'lirst stage pressure
. transmitter used for T ref generation (PT446) failed low. Rod control is in
. Manual and the Steam Bypass System is in the Steam Pressure mode,
- a. Describe the INITIAL response of the control rods (direction and rod speed) if the Rod Control System is placed in Automatic. (1.00)
- b. Describe the INITIAL response of the steam dump valves if the Steam Bypass System is placed in the T ave mode. (1.00)
ANSWER 3.06 (2.00)
- a. The control rods will insert (0.50) at 72 spm. (0.50)
- b. All steam dump valves (0.50) will fully open. (0.50) CHECK AT FACILITY!
REFERENCE BVPS LER 08/16/87 017 BVPS 2LP-SQS-1,3, ELO 11 BVPS 2LP SQS-21L1, ELO 18 3.1 3.0 KA VALUE(S) 001000A102 -016000A201 (KA'S) RO EXAM BANK (12-2-87) 3.06.b The answer that all steam dump valves will fully open is in:orrect. The question states that the transmitter used for T-ref generation (PT-446) failed low. This will generate a Tref-Tave mismatch for the load rejection controller of the steam dump system; however, the steam dumps will remain closed due to the absence of an arming signal provided by first stage pressure transmitter PT-447. Therefore, the correct response is that the steam dump system will not respond when placed in the Tave mode. (Refer to attached copy of OM 2.21.1, p. L 11) o
Ques % 3%.\o. B.V.P.S. - 0.H. 2.21.1
\
MAJOR COMPONENTS Design Data [2 MSS-TCV106Al (Typical for all) Type 8" D100-160-3 Flow (nor/ max), PPH 636,000/890,000 Pressure (Inlet / Outlet), PSIG 1085/-15 . Temperature, F 556 Action, air-to-open Fail position (air / electricity) Closed / Closed The steam dump system has two automatic modes of operation, steam pressure mode and Tavg mode. The operational mode is operator selected by the Steam Dump Control Mode Selector Switch on the benchboard. In steam pressure mode, only the first two banks of valves are operational and they modulate to maintain the steam pressure setpoint set by the operator, using the benchboard mounted steam pressure controller. In Tavg mode, two steam dump controllers are available. The reactor trip controller operates the steam dump valves to restore no load Tavg following a. reactor trip. Only the first two banks of valves are operational efter a reactor trip. The load rejection controller operates all four banks of valves for large load rejections and the first two banks for small load rejections, to restore Tavg to program value. All 18 valves trip closed if Tavg reaches Lo-Lo Tavg. If it is desired to cooldown the reactor plant, the Lo-Lo Tavs interlock may be manually defeated for the three cooldown valves only. Since the Lo-Lo Tavg interlock is dual train, two Steam Dump Control Interlock Selector Switches are provided, one for each train. The Steam Dump Control Interlock Sele.ctor Switches are also used for manually blocking the steam dump control system. All 18 steam dump valves are blocked when the condenser is not available. To be available, the condenser must have sufficient vacuum and at least one cooling tower pump running, ry r -'*g--- --- '-- g- - ~ ' - - - - - - -
,g _
ad en n 5 The lifstXb1EaTe "designe~d' to latch ON since the rate of change signal will disappear as soon as first stage pressure _ reaches its new value. W _ g o thTs'h r e negated' if the condenser is unavailabic or if Tavg reaches Lo-Lo setpoint. [PC447A and B] are reset by momentarily placing the Control Mode Selector switch to RESET. The switch spring returns to TAVG.
\
11 ISSUE 1 REY 2
. j FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 3.07 (2.00)
The plant is operating at 45% turbine load with all control systems in ! automatic. A feedwater flow transmitter (FT496) used for control room feedwater flow indication and steam flow / feed flow mismatch for steam generator 21C is out of service and its associated protection bistables are tripped. A malfunctioning emergency trip header drain valve causes a turbine trip. The reactor trips several seconds later and the main generator output breakers trip even later. Assume no operator action.
- a. Why didn't the reactor trip directly as a result of the turbine trip?
(0.50)
- b. What is the most probable cause for the reactor trip? (0.50) !
- c. How soon after the turbine trips will the main generator output breakers trip? (0.50) ,
- d. What is the basis for the time delay associated with tripping the main generator output breakers? (0.50)
ANSWER 3.07 (2.00)
- a. Reactor power was below P 9 (49%) (0.50)
- b. SG 21C low level coincident with SF/TP mismatch (0.50)
- c. 30 seconds (after turbine trip) (0.50)
- d. Prevent turbine overspeed (0.50)
REFERENCE BVPS LER 08/25/87 019 BVPS OM-35.1, P. 37 BVPS 2LP SQS-1.2, ELO 21 (No ELO for parts e and d) 4.3 3.7 3.0 2.5 KA VALUE(S) 015000K405 015000K407 062000G007 OG2000K402 (KA'S) u.-
, '. u.
FACILITY REVIEW OF VRITTEN EXAMINATION RO EXAM REVIEW (12-2-87)
-3.07.d The answer "prevent turbine overspeed". is only one of three possible correct responses. Other acceptable answers arei
- allow for extended RCP flow prevent missile generation inside containment from RCP flywheel destruction
.(Refer to attached copy of W. Bird letter dated 2/29/80)
a s
@G3 0 A. 3.07,d ,
'd1C/ Zion
"- 523-4041 W 2/2.9/80 590 Turbine Trip / Generator Trip Time Delay All SHUPPS & Zion Instructor's k-There Ppparently have been se - ..isccnccptions float ing around regarding the reason (s) for a _ time. delay for the generL tor trip sub-sequent to most turbine trips. The following is a sun..ary of a bulletin from Mart Merrian - Reactor Protection Analysis, entitled "Functional Requirements for Continuity of Electrical Power to Reactor Coolant Pumps." These requirements will be included in lesson materials for all programs. It is. each instructors responsibility to ensure that he understands and passes along these requirements to the students.
Bases: A reactor trip results in a turbine trip which would result in a generator trip irmediately if no time delay were incorporated.
- r. .., .. .
._n -...n (Th_ere.,are..t,
- w. o._.regireme_nts m.
which are safely related which requ.ir - i alei)~e' rat 6r ~ trip , tiie^ del ay g ec - --- - -- -
...-q---
ure,sJ [ Q. Qf the,, reactor trijs due to overpower, overtemperaI M c M 43 k s hon,di. tion, an ix.edif d fid[ d M M .# _
.- - . _ __ -,r_
Frl p3dificTd e'5FL.Q. __ _. .. .f a,il,ure of a_utomatic.. bus _.itcasier_of.f.)1chay
,b;6{e~sfcoJ1Qesujy.ina1oss_.ofBCPflow. ...Ib i Llc 55lofiffp[,c[Suldg pj((hjj,a@f5AonsegeILces more severe ,than _thatl repor,tedhi,rt thf
[$iifiQ,g,. ii Report... .Howeyer, if_ pumping .poy_er' is 1ost withi,t_im/ n ,, e fdil u_ifE3?p.iie',ratorit, rip,:the loss of flow is noLconsidered-- serious J'
~
.f' tT bRTiifeT.a_ieWeR t'_orda s . b e~~ c n s'hu t d ownl fo r'To'mTt'l rr,e ./ . - --
. a. m u. ..
Mfkf
/
p
. "h.::.;n*c s w tec?n x.tue c -~~n ~+ v , ~ m w na
':Tw
;, ' , .;; J n',vtg QL.is u. m . L ra '.~. . : ;^ X. 2'y ] ;
(;
/
' gre%y q , ,,
[frEqatiiEJ.~._unt1TTulo 1,m m.. .- b_us trE5sTe7T.RC)JV'er.sntedicould _m. - result inf [..M,- M,LdastMclion?forpj,.,., _ _ - . .ne.mlssRe51Wirbicoulddamastihriontai,q/ [ bnNinef oWECTCSiclomponenthWithissonafass@ l, The generator is effectively motorized for the normal gen trip time delay following most turbine trips. This feature minimizes the , consequences of Reactor trips from overporter, overtemperature, and t k low pressure and minimizes RCP overspeed following major LOCA's. 1 [I5.~53.Fe.a~.tTr w bTnE.d e s'i g n_eir5~ V_E_p_o i n ._
- ~
t ,awthe generator: mau..a . tiip't~iiire"delay w._ 4' [a l Wge Fe'htT tWb~GE'_'6v'2 rife eT$'sTr e sWol. .s t.<.-..%
-__a___.. e a n) w i tlii~n' t h e t u rbTny ,
m-u m w __.-m. g.h' elf. 6xpanding t.o. tee? g
._ - . . . _ uuacon.denserij This is a vital concern but is Secondary to the reactor s'afety considerations mentioned earlier, f i
It may be noted that some turbine trips result in immediate { oenerator trip. The probability of these events coupled with failure of electrical buses to auto-bus transfer is considered very unlikely. 0W . k William Bird / Senior Instructor SNUPPS Phase III . WB/ki l 9
/
FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION 3.12 (2.00)
- a. List the FIVE (5) interlock signals / conditions required for a cold leg isolation valve to open when its control switch is taken to OPEN. Assume electrical power is available. Include setpoints. (1.00)
- b. On which Motor Control Centers are the motor breakers for the Loop B Hot Leg Isolation Valve and Loop C Cold Leg Isolsrion Valve located? (1.00)
ANSWER 3.12 (2.00)
- a. 1. Hot leg temperature within 5/20 degrecs F of auctioneered temperature of operating loops. 90.20)
CHECK TEMP AT FACILITY.
- 2. Cold leg temperature within 5/20 degrees F of auctioneered temperature of operating loops. (0.20)
- 3. Isolation valve vent relief line flow >/- 200 gpm for 90 minutes.
(0.20)
- 4. Isolation bypass valve open for 90 minutes. (0.20)
- 5. Hot leg isolation valve open for 90 minutes. (0.20)
- b. Hot leg valve - MCC-2-19-1 (0.50)
Cold leg valve MCC 2-18 (0.50) REFERENCE BVPS OM 6.3, PP. 6,7 BVPS 2LP.SQS-6.2, ELO 8 3.2 3.1 KA VALUE(S) 002000K409 062000A204 ...(KA'S) RO EXAM REVIEV (12 2-87) 3.12.a The correct temperature to be used in the answer is 5'F. (Refer to attached copy of OM 2.6.1, p. 48)
B.V.P.S. - 0.H. GocWon 2.6.1 312.a. -- INSTRUMENTATION AND CONTROL , .
.I The flow through the relief lino is low so that the temperature and boron '
(i ' concentration are brought to equilbrium with the remainder of the system at a relatively slow rate. , e.,
, , m p. , ; .
...,,.n ..
. . . . . .., _m g& ,
- p ,
_ ' WJ30 -* ' b ede% VidW$f Meset(igegred"_tMU$re 'of]hEa't]
..- m m . - .. - .
.m= v~.--,~. --
g g
. -,te.mperat,ure _withia;,SF;iof /sve= ..tgpielod_ temper 3tureoQpergign
- c. Isolation valve vent relief line flow satisfactory (200 gpm) for 90 minutes (Trains A and B)
- d. Isolation bypass valve open for 90 minutes (Trains A and B)
- e. Hot leg isolation valve open for 90 minutes (Trains A and B)
Cold Leg Isolatien Valve [2RCS*MOV591] may be closed provided both of the l following conditions exist:
- a. Control switch in CLOSE i
- b. No motor thermal overload Isolation Bypass Valves (2RCS*MOV585, 586, 5871 C
Refer to Figure 6-34 The isolation bypass valves are contro11eid from Benchboard - Section B. Switch positions are CLOSE-OPEN with red (open) and green (closed) indicating lights. The operation of Isolation Bypass Valve [2RCS*M0Y585) is described below, and which is also typical for [2RCS*MOV586 and 587). l ( Isolation. Bypass Valve [2RCS*MOV585) will, open provided both of the l following conditions are satisfied: ; , t
- l
- a. Control switch in OPEN j
- b. No cotor thermal electrical protection trip Isolation Bypass Valve [2RCS*MOV585) will close provided both the following l conditions are satisfied: ,
' r.
- a. Control switch in CLOSE
)
48 ISSUE 1 REV 2
FACILITY REVIEW OF URITTEN EXAMINATION OUESTION 4.09 (2.25) Refer to the attached figure, Axial Flux Difference Limits. Given each of the following indications, state whether axial flux difference (AFD) is being maintained INSIDE or OUTSIDE the target band. BRIEFLY EXPLAIN your answer. (0.75 each) POWER AFD AFD AFD AFD IEVEL CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL 4
- a. 754 14 12 -13 17
- b. 65% 1 +2 -2 +1
- c. 55% -11 INOPERABLE 15 -11 6LTE .M (2.25)
- a. Inside the target band, (0.50) because less than 2 AFD channels are outside the target band. (0.25)
- b. Outside the target band, (0.50) because 2 AFD channels are outside the target band. (0.25)
- c. Inside the target band, (0.50) because less than 2 AFD channels arc outside the target band. (0.25)
REFERENCE BVPS TECHNICAL SPECIFICATIONS, 3/4.2.1 NO EL0s PROVIDED FOR TECH. SPECS. 3.7 3.4 1.5 KA VALUE(S) 001000c005 001000c011 014000A104 (KA'S) RO EXAM REVIEW (12-2-87) 4.09.a.c The following should also be considered a correct response for full credit:
- Inside the target band because none of the AFD values are outside the target band.
-FACILITY REVIEW OF WRITTEN EXAMINATION
_ QUESTION 4.11 (2.75).
- a. What is the normal limit for annual nonemergency Whole Body radiation exposure to an NCO (on shift) in accordance with the BVPS Radiation Control Manual? (0.50)
- b. List TWO (2) Individuals (by title) who are authorised to INCREASE the above (part A.) annual nonemergency Whole Body radiation exposure limit.
(1.00)
- c. A 20-year old radiation worker (BVPS employee) has received 1 rem during the current quarter and has a lifetime accumulated whole body dose of 8.6 Rem (including current quarter exposure). Assuming this worker has a Form NRC 4 on file, calculate how long (in hours) he could remain in a 250 mrer/ hour gamma radiation field without exceeding any noaemergency Whole Body exposure limits of 10 CFR 20 or BVPS Radiation Control-Manual. SHOW ALL CALCU1ATIONS AND ASSUMPTIONS. (1.25)
ANSWER 4.11 (2.75)
- a. 5 rem (0.50) b, 1. Senior VP, Nuclear Group
- 2. VP, Nuclear Group (any 2 @ 0.50 each)
- 3. Senior Manager, Nuclear Operations
- c. LIMITED BY 5(N 18)-10 rem lifetime whole body limit (0.50 for correct limit) 10 rem - 8.6 rem - 1400 mrem (0.25 for correct method) 1400 mrem 250 mrem /hr - 5.6 hours (0.50 for correct answer)
REFERENCE BVPS RADIATION CONTROL MANUAL, PP 6, 7 NO ELOs PROVIDED FOR RAD CON 2.8 KA VALUE(S) 194001x103 ...(KA'S)
FACILITY REVIEW OF VRITTEN ES' AMINATION RO EXAM REVIEW (12 2 8i) 4.11 h The qu+.stion should be deleted. It is not required knowledge of an R.O. to be able to state those individuals who may authorize an increase above an annual Whole Body radiation exposure limit. This is both an administrative management concern and Radiation Control Department concern only, llowever, it must also be noted that the answer key had omitted an additional correct response to the question. According to EPP/IP 5.3.D.2.3, the Emergency Director may also authorize "exposures in excess of normal guide / limits" (i.e., non emergency whole body limit). (Refer to attached copy of EPP/IP 5.3, p. 2; RCM, Ch.1, p. 7)
~ ^
Qhe.stibn H .Il. 6,; m 2: .. y = ; % . .- Emergency Exposure Criteria and Control I 2.0 Precautions 2.1 The provisions of this procedure are applicable only in actual emergency conditions, and are applicable to BVPS personnel performing assigned emergency functions and' emergency volunteers (eg.: fireman) if applicable. 2.2 The exposure of personnel during emergency operations shall be maintained as low as reasonably achievable, and should be maintained less than the administrative guides established Lt the BVPS Radiation Control Manual (RCM), and/or less than the Federal radiation exposure standards established in 10 ' CFR 20. Administrative means used during normal operations to minimize personnel exposure (sue.h as radiation work permits, radiation clearances, and ALARA measures) should remain in force to the extent consistent with timely implementation of emergency measurns. Mk? I l a ra on con sta ons \ ... w ..w .. . - .. . . - . - ~ ~ ~ > < - ~ . [2 73' "~~ Dire'iit o r~~has'~the "'~ ~ ~ ' ' ' ~ ~ ~ ~
.__ . 3 i
l t ,provided .the _ pre ,
**E51tFN re are met. - - - - - '
I tw w.ww x.--- i 2!4 Personnel shs11 not enter any area where dose rates are unknown or unmeasureable with instruments and dosimetry irimediately available. I 2.5 Appropriate dosimetry equipment, which is capable of measuring the anticipated maximum exposure and type of
- radiations, shall be worn, l
l l 2.6 Extremity dosimaters shall be worn if anticipated exposure is
. greaterthanaboutfive(5) times.thatofthewholebody.,
{} 2 Issue 8 Rev. 0
+ ee e%..ee===.m e -e e-ene .mo es-semuum e w ww .w-,w . - en e = = = = * - . e- ++-=*--+=*e - .. . .
a v. w --
, , ..,i CHAPTER 1 - STANDARDS AND REQUIRDiLNTS
( Column 1 - Column 2 - l Assumed Assumed exposure in exposure in Part~of Body rems for rems for calendar calendar l quarters prior quarters to Jan. 1, 1961 beginning on or after Jan. 1, 1961 Whole body, gonads, active - blood-forming organs, head and trunk, lens of eye 3 3/4 1 1/4
- 7) If calculation of the individual's accumulated occupational dose for all periods prior to January 1, 1961 yields a result higher than the applicable accumulated dose value for the individual as of that date, 3 rems per quarter, the excess may be disregarded.
Cumulative Quarterly Annual Part of Body Dose (rems) Dose (rems) Dose (rems) Whole Body 5 (N-18) 3 *12 Skin of Whole Body ---- 7.5 30 Hands and forearmsg ---- 18.75 75 feet and ankles
' ' ~ ' ~ ~~
Atti f[y.. -{c'umulativeir. pu sne 3bi v ' t j i
$ e1 ej
~ ~ ~ ~ ~ ~ ' ~ ~
- b
!Th8 cumUlattve,. wh6H p'6d ~ " 'em)
TcoWt'G&dF,7erid6Q
*fiiPWTc'ENndEi' iia . T{h tarpeiRINE"EtMlWarat'all i otur"feell t'Lue, on a- -
.eas -
MiWA90NtWTllHPffWif
<Prasident,-Nuclear;' may:4et p
4"2 hoye_L%K.a..but eak.40.*xceed.Vapplicableiesrgency exposur g A must suppV ,~1n writing, authorization for the exposure. (Refer , to SAP 23 for guidance and the required forms.) ; i Individual exposure guidelines are upper guidance levels and do
, not authorize unwarranted or unr{ecessary dose, f '- '
ISSUE 4
e
. k b6W%O fV 85. Ol k , ,,,,,,g, ,g,,,
g ,,g i DAtLT 5 TAT SA!.MCt
' Daft !!Mt fleet emot be' at steady state for 30 enestes. Te vertip thte trend all senputer totate betas used at i steuse tatsreats for 30 stoutes and vertly that they teosta withte e 21 band respeettvely.
% p.62 % s 43 % p*44 t' U1150 %
[N . Fever t.evel - . B 61 (515 Cettset) ( ,
" 1. 2&DG FIT 1008 %. ilDC FIT 100C %
- 1. Resord the ladtstewat blowdows flew readtrase ll0G FtT100A malt tple eaan % flev (in aestaan format. to. 22%
- 0.22) tLees too to entate tne Gth for each 11aea A CFit t CPN C CPN A44 ell tutee flows to ettate total tiewtown flevi CFM. If lese than 100 CPM. soeves tieweows to te sete -
nr plattes a sete to sine S. 9. end 11. toce a uct s 400 c
- 3. Feeswate trss ttivoa er 2rws ftt5 4 er 2WS tits *c or Temperature (7)* T06184 TO4)SA 704$$A
- 3. scene Prese.se 2 mss Pief. er inss eta 6* er 2rss tt*** er (PstA)** W 16 00*14 00654
( + 16.7 ( + 16.7 (
- 14.7-
- P11A)
- FS1A)
- PSIA )
2FWS*Ft478 or 2fv5+Ft444 et 2WS Ft494 or Mpv (a10g P M 00610 U0410 00*10 f.
$- ax.,tmCA_%
- 0 m.. m00._%*0 nx,mc=_t*0 0 #7 ~
;,, , .10. ,,, ,
- 4. : 0.07
- _ s 10*Prt 0 a 0.01
- _ a 10*PPE 0 0.07
- _ a 1C FP4 1
- e. .
4 Mst
- Pr4
- Ms: t u t0 Pn
< n- --
- 8. i tW e
*ST g )
,0 ('iq Ib ng; b5T a108 E}
wr I u-
.,, .,0 (.,0. m )
- 4 prv hpv fs10
,,. . . . . , t.t.p 0 . .t.p .
1' '. *IO DrE Step 12
- f
- 14. Loop 9 Power t$tep 10
- 8tep kl *
( [s104 m\ Stat it
- _ ) y nr i f
( l (Step 50
- 6tep 11 *
- 15. Leap C Fower gg p gg e .)
l a106g i vr I* 4 RCSOutputfs10>rI Q }
- Sua of Steps 13. 14. and 11
- _
I'* Fat Aeacter Power * {tC8 Output (Step 16)
- RCP Outpet] * (
- 47.31 a 10( $
- 3 1C6g -
, >r vr t II* Wth
- pet teatter Power (step 17) LO*4 Wut
- Wry 3,61) STV N' % toaster Pmer
- Weg (Stes 181 a 10C%
- s 2e'l
- 20. If teatter Power le > 2412 W uT tannedtetely reduce pewer te ( 2653 Mh ty Cats AMVST p-62 t 44 camn n 61 W 4) smi stasch Tavs & tret pries to a qwettes 51 Detester currents en s!'s. MEtta % % % %
Adjvat CA15 (fi DritCTca a ersatwT ve .4 =a es
- 4. ut meter teedtess enseeds 10C% ett M m ten 6
- 2. If meter seedtese entfor by ~* 1% f ree toisolsted '# ** ** **
pewer, een estavleted power te t 54 ett
- 3. 18 meter readtege 46fier by ett free satsulated
*0% Calculattene performed tyi power when eslaulated power le } 911.
I AT1/164 If any of the eleve tend &ttens estet 44)v44 CAtu en the front of the power r eage itsver 9 setti the tailsator Calevlatica reviewed and approved tys tende the esse as the tattulated tatue. Record flaal 9eedtage if ad petseat to regatted. E*II'II4 Operet tag lap. rvleet Sovlevi
~
batt/T N 155:;1 1 RTV !
.--_________-______--__a
B.V.P.S. - O. 2.56.4 C1 (Pcg] 2 cf 2) DAILY HEAT BALASCE , at Balonce Calculation Instructions
-O b R: cord the individual loop blowdown flow percentages. Calculate the GPM's for each loop using the decimal. equivalent of the percentage. Calculate the total blowdown f1:w. If total blowdown flow is less than 100 GPM, assume blevdown to be zero by using zero in steps 5, 9, and 11.
. Obtrin feedwater temperature (F) from the appropriate instrument.
- 3. Obtain steam pressure from the appropriate instrument and convert to absolute prcasure.
- 6. Obtain the mass flow rate of feedwater ( ,) from the appropriate instrument. O
- 5. Ob tnin the flow rate of blowdown (unless total blewdown < 100 GPM) from the apprcpriate instrucent and convert to 100 PPH (91D)
- 6.
TJ -K M *P Calculate mass flow rate of steam (M37): ST BD*
- 7. Obtain feedwater enthalpy (hpg) from saturated steam table using hf at feedwater tccperature.
.8. Obtcin steam enthalpy (hST) r m saturated steam table using hg at steam pressure.
- 9. Obtain blowdown enthalpy (hBD), unless total blowdown < 100 GPM, from saturated stc:s table using hf at steam pressure. e O. Multiply ST and h ST r applicable loop. ,
- 1. Multiply F!
BD and h BD r applicable loop (unless total blowdown < 100 GPM).
- 2. Multiply FQ, and lig for applicable loop.
- 3. Calculate Loop A Power: Step 10 + Step 11 - Step 12, 4 Calculate Loop B Power: Step 10 + Step 11 - Step 12.
- 5. Csiculate Loop C Power: Step 10 + Step 11 - Step 12.
,. Cciculate RCS output. /. Calculate Net Reactor Power.
5' . Calculate F"J g. J. Calculate % Reactor Power. ISSL'E I REV 1 0
FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION 5.01 (2.50)
- a. Calculate reactor power (MWt) using the following information. (1.50) feedwater temperature - 420 F.
feedvater flow - 3.60 E+6 lbm/hr (per S/C) S/G pressure - 785 psig condenser pressure - 2.5 psia
' drain pump flow - 3.0 E+6 lbm/hr 1st stage pressure - 735 psig
- b. What are WO (2) components which, if not accounted for, cause the calculated value for reactor power to be greater than actual reactor power 7 (1.00)
ANSVER 5.01 (2.50)
- a. h(feedwater) - 396.9 BTU /lbm (0.50) h(steam) - 1200 BTU /lbm (0.50) reactor power - (10.8 E+6 lbm/hr)*(1200 - 396.9 BTU /lba)
*(0.293 W/ BTU /hr)/(1.0 E+6 W/MW) - 2541 MW (0.50)
- b. RCPs (0.50)
PZR heaters (0.50) REFERENCE BVPS LP-TMO 6 Enabling Objective 11 BVPS Thermo Text Chapter 6, p. 37; Question 48 BVPS LP TMO 6 page 18 K/A 193007 Kl.08 3.4 193007K108 ...(KA's) SRO EXAM REVIEW (12-2-87) 5.01.b Another acceptable answer,should be blowdown flow. The attached daily Heat Balance shows a steam calculated by subtracting ablowdown fr * "feedvater-
FACILITY REVIEV OF WRITTEN EXAMINATION OUESTION 5.04 (3.75) Complete Parts B D, and E of the attached Estimated Critical Position Calculation (Attachment 1) given the following information: A reactor trip occurs at 2030 on 11/28/1987 from 1004 power after the plant had been operating for 30 days with Bank b rods at 190 steps. Boron concentration was 930 ppa before the trip and has been increased to 1190 ppa and is now steady at this value. The estimated time for criticality _is 1000 on 11/30/1987. ANSWER 5.04 (3.75) 3.1.I- 1400 pcm
~B.2.1 -9068 pcm B.2.II .-11231 pcm B.2.III +2163 pcm B.3.I 3000 pcm B.3.II -1500 pem B.3.III 1500 pcm B.4.1 610 pcm [15 x 0.25]
B.4.II 760 pcm B.4.III +150 pcm D.II 80 pcm D.III 667 pcm D.IV Bank D Step 98 E.IV Bank D Step 180 E.V Bank D Step 50 REFERENCE BVPS LP RT-9 Enabling Objective 6 BVPS - OH 1.50.4 K/A 192008 Kl.07 3.6 192008K107 ...(KA's) m.
FACILITY REVIEW OF VRITTEN EXAMINATION SRO EXAM REVIEV (12-2-87) 1Ei Make allowances for differences in graph interpretation. Answer B 2II should be 11186 pcm Answer D II should be -140 pcm Answer D III should be 754 pcm Answer D IV should be 88 steps on D Answer E IV should be 160 steps on D Answer EV should be 42 steps on D I
d FACILITY REVIEW OF VRITTEN EXAFIINATION OUESTION 5.05 (2.00)
'Using the attached Shutdown Margin Calculation (Attachment 2), complete the-form and calculate SHUTDOWN MARGIN in percent delta K/K. Assume boron concentration is 900 ppa and the reactor has been at 1006 power for 5 days.
ANSWER 5.05 (2.00) 1.b. +1225 pcm (0.40)
- rod worths: SDA 2589 SDB 1287 (0.40) for right group of numbers CBA 1167 r CBB 1960 (0,40) for conversion from 4 delta rho to pcm CBC 1324 CBD 1300 (0.40) for Bank D at 190 steps shutdown margin = .4.3694 delta k/k (0.40) 2 REFERENCE BVPS LP.RT.9 Enabling Objective 4 BVPS . OH 2.55A.4 l K/A 192002 Kl.13 3.7 192002K113 ...(KA's) - SRO EXAM REVIEW (12-2 87)
- 2 Q1 Shutdown Margin Calculation should allow for a different assumed Rod Height (not given in question). BVPS runs with all rods out, not with Bank D at 190 steps as per answer key. Therefore, calculations g done with all rods out should be considered acceptable, f
4 4
L FACILITY REVIEW OF URITTEN' EXAMINATION l QLILFION 5.09 (2.25). HOW (Increase, Decrease, No Change) and WHY.would each of the following parameters affect the margin to DNB7 Pressurizer temperature increase 5 degrees a.- b ',' Mass flow rate through the core-increases 10%
- c. AFD increases to +10%
r EigR 5.09 (2.25)
- a. Increases (0.25) as PRZR temperature rises, so does saturation pressure (0.50)
'b. Increases (0.25) because core delta T will decrease (to keep power constant) (0.25) reducing T(hot) (0.25) c.
Decreases (0.25) the core (0.25) because causing the botmore power channel is being factor boundsp)roduced p in this areaintothe be top half o i approached (0.25) g 4 , REFERENCE BVPS LP TMO.7 Enabling Objective 12 BVPS Thermodynamic Text Chapter 7 page 17 K/A 193008 Kl.05 3.6 193008K105 ...(KA's) r SRO EXAM REVIEW (12-2 87) { i 5.09.e Delete reference to hot channel factors. This is not needed to answer the question, t t 1 [ I l i f I l L L i [ I
' FACILITY REVIEW OF VRITTEN EXAMINATION (2.05)
Q'd ESTION 5. lQ a .- WHICH initial temperature, 300 F or 547 F, will give the largest change ?,n the magnitude of MTC as boron concentration is lowered from 1000 to 500 ppe? Justify your answer. (1.35)
- b. . WHY does power defect become more negative as the' core ages? (0.70)
ANSWER 5.10 (2.05) an. 547 (0.35) because the change in density / degree F is much greater than at. Iower temperatures (0.50) so more boron atoms enter or leave the core causing a larger reactivity change for a given temperature change (0.50),
- b. MTC, a component of the power defect, becomes more negative JOAJ7 due-te daerammed baron conenntration (0.35). (o,7,}
~
REFERENCE BVPS LP RT 6 Enabling Objectives 1, 2,10 BVPS Reactor Theory Text Chapter 6, pp. 6.13,16 K/A 192004 Kl.06 3.1 4 SRO EXAM REVIEW (12 2 87) 5.10.b Delete second part of answer because it is not asked for in the question (i.e. due to decreased boron concentration), i 1 1 I l 4
FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 6.01a (2.50)
- a. List THREE (3) conditions / signals that will cause a steam generator feedwater pump to automatically trip. DO NOT include manual or motor / bus electrical trips). (0.75)
- b. The reactor operator tries to start a main feed pump by placing the control switch to the START position and immediately releasing it. The pump does not start. Briefly explain WHY the pump did not start and how the operator should have manipulated the control switch. (0.75)
I
- c. VHAT is the maximum flowrate for a single feedwater pump operating alone?
WHY does it have this limit? (1.00) ANSVER 6.01a (2.50) ./
/
5% e # .{, 44 (lMa,f.'
- a. - feedwater isolation signal (a
- sustained low suction pressure (0.25 x 3)
- lube oil pressure extreme low
- b. Switch must be held in the START position (0.35) until the feedwater recirculating water valve is fully open (0.40)
- c. 16,000 gpm (8.0 apph) (0.50) prevent runout conditions (0.50)
REFERENCE Beaver Valley 2LP SQS-24.1 Enabling Objectives 7,16 BVPS - OM 2.24.1, pp. 20,21; 2.24.2, pp. 2,3 K/A 059000 K4.16 3.2 K/A 059000 C0.09 3.1 K/A 059000 CO.10 2.9 059000r:416 059000G010 059000G009 ...(PA's) SRO EXAM REVIEW (12-2-87) 1.dia Accept answer that states a feedwater isolation comes from a safety injection or a hi hi steam generator level. See attached logic diagram.
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FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION M (2.50)
- a. During refueling operations radiation monitor 2HVR*RQ104A, Local Containment Purge, alarms high. Before the alarm, WHERE is this effluent being released, and WHAT automatic actions occur to prevent this release from continuing? (1.00)
- b. State the automatic actions which occur, if any, when 2 ARC-RQ100, Local. Air Ejector Discharge, alarms high. (0.50)
- c. Lanf are there two (2) sample flow paths for the Wide Range Gas Monitor, 2HVS*RQIl09B, and WHAT type of effluent does it monitor? (1.00)
ANSWER 6.02 (2.50)
- a. auxiliary building vent (elevation 773) (0.50) containment purge exhaust isolation dampers close (0.25) containment purge air supply dampers close (0.25)
- b. none (0.50)
- c. Samples different radiation levels at different flow rates (0.6)
Monitors noble gases (0.40) REFERENCE Beaver Valley 2LP.SQS 43.1 Enabling Objectives 2.p. , 4. ,12. BVPS OM 2.43.4, pp. ACN, ACX; 2.43.5. Table 43-1, p. 3 2LP.SQS 43.1, p. 27 of 48 K/A 072000 K4.01 3.6 K/A 072000 A3.01 3.1 072000K401 072000A301 . .(KA's) SRO EXAM REVIEW (12-2 87) 6.02.a During refueling the containment purge path must, by Tech. Specs., be discharging through the Main Filter Banks and out the Elevated Release on Containment, not out the auxiliary building vent as the answer key indicates.
G0nsnuna c.ci t
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REFUELING OPERATIONS CONTAINNENT BUILDING PENETRATIONS i LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. .The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or v ~
- 2. Exhausting at less than or equal to 7500 cfm through OPERABLE Containment Purge and Exhaust Isolation Valves with isolation times as specified in Table 3.6-1 to OPERABLE HEPA filters and charcoal adsorbers of the Supplemental Leak Collection and Release System (SLCRS). , _
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel'within i the contaunnent. .. ! ACTION: Withtherequirementsoftheabovespecificatiodnotsatisfied,immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of Spe(ification 3.0.3 are not ap'plicabl,s. l SURVEILLANCE REQUIREMENTS l I 4.9.4.1 Each of the above required containment penetrations shall be determined
' to be in its above required condition within 150. hours prior to the start of and at least once per 7 days during C0RE ALTERAT. IONS or movement of irradiated fuel in the containment. -t O
4.9.4.2 The containment purge and exhaust system shall be demonstrated OPERABLE by: !
- a. Verifying the flow rate to the SLCRS a't least once per 24 hours
- when the system is in operation. i i
- b. Testing the Containment Purge and Exhaust Isolation Valves per the applicable portions of Specification 4.6.3.1.2, and
( c. Testing the SLCRS por Specification 4.7.8.1 with the exception of item 4.7.8.1.c.2. i i SEAVER VALLEY - UNIT 2 3/4 9-4
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FACILITY REVIEV 0F VRITTEN EXAMINATION OUESTION 6.03 (1.75)
- a. VHAT TVO (2) chemicals are added to the RCS to control pH and oxygen when in MODE 57 Indicate which chemicals for which function. (0.50) b, During which type of dilution operation (alternate dilute or dilute) will reactor coolant hydrogen concentration become depleted faster? Justify your Answer. (0.75)
- c. WHY must the-VOT be maintained at a minimum pressure of 15 psig7 (0.50)
'l ANSVER 6.03 -(1.75) 4
- c. Li70H . pH (0.25); hydrazine 01 .v.25)
, b.. Alternate dilute (0.25) because this flow path partially bypasses the VCT (0.25) preventing the hydrogen cover gas from being absorbed by the dilution flow (0.25).
- c. Required backpressure for the reactor coolant pump seals (b..
REFERENCE Beaver Valley 2LP.SQS.7.1 Enabling Objectives 1,10,17 BVPS OM 2,7.1, pp. 3, 4, 37; 2.7.2, pp. 1, 5 K/A 004000 K4,01 3.3 K/A 004000 K4.02 2.6 K/A 004000 K4.08 3.2 K/A 004000 CO.10 3.4 004000K408 004000K402 004000K401 0040000010 0040000009
...(KA's) 1
, SRO EXAM REVIEW (12 2 87) a 6.03.a Answer should be given full credit for LiOH vice Li70H. Training material for BVPS uses LiOH. i
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-CHEMISTRY FUNDAMENTALS-LP. CHEM-18 f REACTOR PLANT CHEMISTRY '
- 1) ~ containment depressurization system (A-4)-
2); chemical and volume' control system III. Reactor plant system sampling requirements
. 'A. ' Sampling frequencies are listed in BVPS chemistry manual,
!- . " Chapter 3:"Sampling and Testing". '
. l.- ' Sampling requirements are listed.by system in' body
of chapter as well as appendices. ,. a. Give students handout for RCS (A-5 thru A-9).
. 2. ' Sampling r.equirements come from Westinghouse )
recommendations, surveillance requirements of Tech;
. Specs., and= action statements of Tech. Specs.
Note: Tech. . Specs .will be covered. in .LP-CHEM-19 .
. r IV. - Analysis' of - coolant chemistry .
A. Predictable changes over fuel cycle. s
- 1. Start of cycle.
. m.
- a. Boron concentration N1000 ppm. -
- b. Li+ concentration as ~ necessary- (> mid range) .
- 2. During cycle ,
- a. Boron,"burndd out" '{ converted to ~L1+) or else y diluted'from system.
- b. Boron concentration can be increased or decreased to load follow.
- c. Li+ concentration will increase as boron is "burned ott". At the start of cycle when boron concentrations are high (and hence Li+ production is high) Li+ might have to be removed by cation exchanger.
- 1. B10 + nl
- 3Li +2e4 H
- 2 .~ : 3Li 7'+:H2O w'H A :LiOH A Li+ +/OH-f
- 3. End of cycle
- a. Boron <100 ppm
- b. Li+ as maintained FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION 6.04 (2.40)
- a. List FOUR (4) locations where SR indication can be found. (1.40)
- b. A reactor shutdown is in progress with the source range (SR) detector reading about 10,000 cps and both intermediate range (IR) detectscs reading 1 x 10E 11 amps. Ten minutes later the SR detectors resa about 1,000 cps but the IR detectors still read 1 x 10E-11 amps. WHY coes the IR detector output remain the same? (0.50)
- c. The plant is operating at 100% power with NI-44 out-of-service. If an automatic reactor trip occurs and NI-43 is failed as is, WHAT effect, if any, will this have on the nuclear instrumentation system's ability to monitor neutron flux? (0.50)
ANSWER 6.04 (2.40)
- a. NIS rack benchboard - section B emergency shutdown panel (SDP) (0.35 x 4) alternate shutdown panel (ASP)
- b. 1 x 10E 11 amp signal is used as a reference for gamma compensation (0.50)
- c. Source range detectors cannot be energized (0.50)
REFERENCE Beaver Valley 2LP-SQS-2.2 Enabling Objectives 4.g., 8.a., 16 BVPS - OM 2.2.1, pp. 11,16,27; 2.2, p. 1; 2.01.1, p. 11 K/A 015000 K5.01 3.2 K/A 015000 G0.06 2.8 K/A 015000 G0.07 4.0 K/A 015000 A3.03 3.9 015000K501 015000G006 015000A403 ...(K/A's) SRO EXAM REVIEW (12-2-87) 6.04.a There are two additional answers not given in key.
- Remote recorder 2NME-NR45
- PSMS computer (see OM 2, Section 1, p.16 and Section 2, pp. 4, 5 )
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. l B.V.P.S. - 0.M. 2.2.1 MAJOR COMPONENTS i SR Remote, Count Rate Meter [2NMS-NI31B, 32B, 31BA, 32BAl The SR remote meter indication is an analog signal proportional to the count-rate being received, and is obtained from the 0-1 mi111 ampere isolation amplifier output.
The meters are mounted on the Benchboard - Section B (2NMS-NI31B, '
)
32B)and the Emergency Shutdown Panel (2NMS-NI31BA, 32BA) and calibrated ' logarithmically from 1E+0 to 2E+6 counts per second. These meters give the same indication at the Benchboard and Emergency Shutdown Panel. as is displayed by the local meter on the corresponding source-range drawer. i
= Remote Recorder (2NME-NR45) ~ --
~ %
~~_
.s '
f This two-pen recorder located on the Vertical Board - Section B is . y capable of continuously recording any two NIS channels at a time. Each
- l pen receives its signal through a multiposition switch (1N45 and 2N45) located on the Benchboard - Section B which can select any one. of the eight nuclear channels. In the case of the source ranges, a 0-50 millivolt DC signal, proportional to the count-rate range of 1E+0 to 1E+6 counts per second, is supplied for recording during source range ,
4 operation.
~ ~ - _ _ _ _ _ _ _
_ _. __.m , , SR Startup Rate Circuitrv (N37) i The startup-rate (SUR) drawer receives four input signals (0-10 VDC), one from each of the source and intermediate range channels. Four rate amplifier modules condition these signals and transmit their rate signals to its respective SUR meters on Benchboard - Section B and the i Emergene.y Shutdown Panel. The indicators for source range channels NI-l 31 and NI-32 are (2NMS-NI31D, 32D] and [2NMS-NI31DA, 32DA) l respectively. A test module is provided which may be used to inject a j test signal into any one of the rate circuits. The test signal can be monitored on a test meter mounted on the front panel of the SUR drawer. l Two power supplies are installed in such a manner as to ensure rate l indication from at least one source and one intermediate range channel. Intermediate Range Channels 1 r Intermediate range (IR) output information is tabulated in Specific l Instrumentation and Control. Each IR channel receives a DC signal from l a compensated ion chamber and supplies a signal positive high VDC and a j negative compensating voltage to its respective detector. The l compensating voltage cancels that signal imposed on the detector l resulting from gamma radiation. Both voltage supplies are adjustable l through controls located inside the IR channel drawer. The detector l signal is received by the intermediate range logarithmic amplifier. l This modular unit, co= prised of several operational amplifiers and 1 16 ISSUE 1 REV 1
- Gh&Jw - L.ok e B.V.P.S. - 0.M. 2.2.2 1,
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?"~ SET POIWS
.,p-DELTA FLUX ALARM 60 penalty minutes in last 24 hours NIS CHANNEL IN 'IIST N/A NIS INT RNG NEUTRON FLUX HIGH Current Equiv to REACTOR TRIP 25% Power-NIS 2/4 PWR RNG LOW SETPOINT 25%
NE WRON FLUX HIGH RX TRIP NIS 2/4 PWR RNG HIGH SETPOIhT 109% "' NEWRON FLUX HIGH RX TRIP NIS SOURCE RNG NEWRON FLUX 10' counts /sec HIGH REACTOR TRIP NIS 2/4 PWR RNG NEUTRON FLUX 109% RATE'HIGH REACTOR TRIP P-10 PERHISSIVE 10% of full power- ' NOT P-7 <10% of full power NIS SOURCE RANGE TRIP BLOCKED N/A NIS IEERMEDIATE RANGE TRIP N/A > BLOCKED POWER RANGE LOW SETPOIh7 N/A TRIP BLOCKED NOT P-8 N/A l P-6 PERMISSIVE N/A NOT P-9 N/A Computer Digital' ' l DELTA FLUX OWSIDE TARGET power above 90% and l BAND delta flux outside target band DELTA FLUX IN ALARM 60 penalty minutes
- in last 24 hours PWR RNG CHN 1 P9 PERM 549% of full power l 4 ISSUE 1 REV 1
- l. ~
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.s ;i PWR RNG CHN 2 P9 PERM $49% of full powert :
PWR RNG CHN 3 P9 PERM 549% of full power PWR RNG CHN 4 P9 PERM 549% of full power
/ .
3 PWR RNG CHN 1 P8 PERM 30% of full power ! PWR RNG CHN 2 P8 PERM 30% of full power 1
~
PWR RNG CHN 3 P8 PERM 30% of ful1 power-PWR RNG CHN 4 P8 PERM 30% of full, power I!ff RNG 1 HIGH FLUX Current equiv to 25% power- , INT RNG 2 HIGH FLUX Current 25% power- equiv;to
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- j'
. SOURCE,RNG CHN 2 FLUX 10' CPS i
'~~
Computer Analog - SOURCE RNG DET. 1 FLUX LOG N/A SOURCE RNG DET. 2 FLUX LOG N/A IhT RNG DET. 1 FLUX LOG N/A INT RNG DET. 2 FLUX LOG N/A PWR RNG CHN 1 FLUX 109% PWR RNG CHN 2 FLUX 109% i PWR RNG CHN 3 FLUX 109% PWR RNG CHN 4 FLUX 109% ! 5 ISSUE 1 REV 1
l FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 6.06 (3.00)
- a. WHAT PCCW realignment occurs on a low-low PCCW surge tank level? (0.50)
- b. Component cooling water pump 2CCP*P21C is racked in on bus 2AE and is in auto. WHAT THREE (3) conditions need to be satisfied before the pump will automatically start? Assume all associated equipment fur.ctions properly.
(1.00)
- c. WHAT are SIX primary component cooling water (PCCW) containmont loads that are isolated by a containment isolation phase B signal? (1.50)
ANSWER 6.06 (3.00)
- a. 2CCP*MOV175-1, 176-1, 177-1, 178-1 (non-nuclear safety portion of PCCW)
PCCW system supply and return isolation valves close (0.50)
- b. - no containment isolation phase B signals present
- PCCW system header pressure low (3 x 0.33)
- diesel loading sequence signal present
- c. - reactor coolant pumps CRDM shroud cooling coils shield tank cooler
-excess letdown heat exchanger (0.25 x 6)
- residual heat removal heat exchanger residual heat removal pump seal coolers
- primary drains cooler REFERENCE Beaver Valley 2LP-SQS-15.1 Enabling Objectives 2, 4, 9 BVPS - OM 2.15.1, pp. 9, 17, 24, 25 K/A 008000 Kl.02 3.4 K/A 008000 K3.01 3.5 K/A 008000 A3.04 3.7 008000K301 00800K102 008000A304 ...(KA's) i
FACILITY REVIEW OF WRITTEN EXAMINATION SRO EXAM REVIEW (12-2-87) 6.06.a Valve numbers'should not be required-for correct answer. 6.06.b The second and third answers in the key'are incomplete. The correct answers are:
- diesel loading sequence and pump A "not racked in"
- System header pressure low and breaker 2E7 closed.
See attached diagram. 6.06.c Reactor Coolant Pumps is given as an acceptable answer. The RCP's actually have four separate cooling units: Thermal Barrier, Stator - Upper Lube 011, and Lower Lube 011. It is requested that each of these be considered a correct answer.
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FACILITY REVIEW OF VRITTEN EXAMINATION OUESTION 6.09 (3.00) For the following reactor protection trips,' state its Technical Specification basis AND at what point, if any, the trip is blocked.
- a. . pressurizer high water level
'b. low feedwater flow-
- c. low reactor' coolant flow
- d. overpower delta-T
- e. pressurizer low pressure
.f. power range high negative neutron flux rate
- g. reactor trip on a turbine trip ANSWER 6.09 (3.00)_
- a. RCS overpressurization (0.30)
. blocked below P-7 (0.20)
- b. -loss of heat sink (0.30)
- c. low DNBR protectic (0.30) 2 out of 3 loop trip blocked below P-7 (0.15) 1 out of 3 loop trip blocked below P-8 (0.15)
- d. fuel rod rating protection (0.30) e, low DNBR protection (0.30) blocked below P-7 (0.20)
- f. low DNBR protection (0.30)
- g. provides additional protection beyond that required (0.30) blocked below P-9 (0.20)
REFERENCE Beaver Valley 2LP-SQS-1.1 Enabling Objectives 5, 7 BVPS - OM 2.01.1, pp. 6 through 9 BVPS - Unit 2 Technical Specifications B2-3 through B2-6 K/A'012000 K4.02 3.9 012000K402 ...(KA's) SRO EXAM REVIEW (12-2-87) 6 cf_mg An acceptable answer should be: a negative rate trip protects against two or more dropped rods. See attached Tech. Spec. Bases B2-3. d
Q w s m oro c, .O S F LIMITING SAFETY SYSTEM SETTINGS BASES specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the detersiination of the trip setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not i met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, ' in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. , Manual Reactor Trip , The Manual Reactor Trip is a redundant channel to the automatic protective
- l. instrumentation channels and provides manual reactor trip capability.
Power Range, Neutron Flux . The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by
-( temperature and pressure protective circuitry. The low setpoint provides redund-ant protection in the power range for a power excursion beginning from low power.
The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10.becomes inactive (three of the four channels indicate a power level l below approximately 10 percent of RATED THERMAL POWER). Power Ranae, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low ( trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above 1.30 for control rod drop accidents; At high power a multiple rod drop accident could cause local flux peaking which, when inconjunctionwithnuclearpowerbeingmaintainedequivalenttoturbinepower by action of the automatic rod control system, could cause an unconservative l local DNBR to exist. The Power Range Negative Rate trip will prevent this from l occurring by tripping the reactor. No credit is taken for operation of the l Power Range Negative Rate trip for those control rod drop accidents for which l ONBRs will be greater than 1.30. l (. BEAVER VALLEY - UNIT 2 8 2-3 l l
FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 6.11 (1.909)
- a. For a large break LOCA, WHAT are the. minimum emergency ~ core cooling system pumps required to cover. exposed fuel and limit possible core damage?
(0.70) '
- b. Following SI reset,.WHAT operator action (s) must be performed in order to reinstate automatic re-initiation of SI? (0.50)
- c. The accumulators are isolated during a normal plant cooldown. WHAT RCS.
system safety limit can be violated if this is not done? (0.70) ANSWER 6.11 (1,90)
- a. 1 HHSI pump (0.35) 1 LHSI pump (0.35) b, close the reactor trip breakers (0.50)
- c. RCS overpressurization (0.35) at reduced RCS temperatures (0.35)
REFERENCE Beaver Valley 2LP SQS-11.1 Enabling Objectives 1, 4,12 BVPS - OM 2.11., p. 2 BVPS - E.O.P. 2.53.1, ES-1.1, p. 2 Beaver Valley - Unit 2 Technical Specifications, p. B3/4 5-1 K/A 006000 K6.02 3.9 K/A 006000 K6.03 3.9 K/A 006020 K4.06 4.2 006020K406 006000K603 006000K602 ...(KA's) SRO EXAM REVIEW (12-2-87) 6.11.c Delete last part of answer, question establishes the situation as already being at a reduced temperature. NOTE: BVPS has two safety limits per Tech. Specs.; neither would pertain to this question. See Tech. Specs. 2.1 and 2.1.2.
FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 7.01 ( 1.~ 50) , For the following questions, assume BVPS - OM 51, Station Shutdown Procedure, is in use.
- a. When using condenser steam dumps, WHAT operator action (s) must be taken to cooldown the RCS below the Lo-Lo Tavg setpoint? (0.50)
- b. With Residual Heat Removal System (RHS) in service, WHY should at least one reactor coolant pump remain in service until RCH temperature is less than 200 degrees F7 (0.50)
- c. If minimum RLS flow requirements CANNOT be met while in Mode 4, the operator's immediate response is to refer to WHAT procedure? (0.50)
ANSWER 7.01 (1.50)
- a. Place steam bypass interlock selection switch to the DEFEAT TAVG position (0.50)
- b. prevent reactor vessel void formation (maintain RCS subcooling) (0.50)
- c. BVPS - E.O.P. ES-0.2, "Natural Circulation Cooldown" (0.50)
REFERENCE BVPS 2LP-SQS-2~ E.O. 4; 2LP-SQS-50.51.52.1 E.O. 2, 3 BVPS - OM 2.51.4, pp. C9, D2, D4; 2.51.2 p. 3; 2.53C.4, p. 3 K/A 005000 GO.10 3.5 K/A 005000 G0.15 3.9 K/A 041020 A4.08 3.1 041020A408 005000G015 005000G010 ...(KA's) SRO EXAM REVIEW (12-2-87) 7.01.b This question should oe deleted. It is not required knowledge. The question is from a note in Procedure D, Chapter 51. The note also contains the reason the note is put in procedure. 7.01.c It is given that the plant is in !!cda 4, so RCP's and/or RHR may be in service. If you are using RHR and flow was lost, you would go to AOP 2.10.1, "Loss of RHR". Therefore, AOP 2.10.1 should be an additional acceptable answer.
y QUE%dd mOs 9 B.V.P.S. - 0.M. 2.51.4 Issue 1/ Revision 2 Page D4 of 16 D. STATION SHWDOWN - COOLDOWN FROM THE HOT SHUTDOWN (MODE 4) TO THE COLD SHWDOWN (MODE 5) (Continued) l NOTE: At least one reactor coolant pump l should be maintained in . operation j until reactor coolant system temperature is below 200F. . This will provide the reactor vessel head < and other stagnant areas with cooling when RMS is in service and ' maintain the entire reactor coolant system subcooled.
- 5. Adjust '[2RHS*HCV758A(758B)], Benchboard-A, l Residual Heat Removal Outlet Flow Control Valve, as necessary, to maintain reactor coolant system temperature at 250 i SF. , /
- 6. When the steam generator pressures reach 25 psig (as verified by computer points PO401A, PO420A, PO440A), place the steam generators in wet layup in accordance with OM-2.21.4.C, "Wet Layup From Hot Shutdown." /
- 7. Establish steam generator chemistry as follows:
- a. If necessary, shutdown the steam generator start-up feed , pump in accordance with OM-2.24.4.H, "Steam Generator Feedwater System Shutdown." /,
- b. Shutdown the steam generator blowdown l system as described in OM-2.25.4.C.
"Steam Generator Blowdown System - System Shutdown". /
1 CAWION: DUE TO DEAD WEIGHT. CONCERNS DO NOT EXCEED 98% LEVEL ON WIDE RANGE SG LEVEL INDICATORS.
- c. Coincident with the addition of wet layup chemicals to the steam generators, increase steam generator levels to at least 94% in the wide range in accordance with OM-2.24.4.I. "Feeding Steam Generators at Low Pressure And Little or No Steam Flow." This establishes a level above the primary separators so that the layup chemicals can mix by thermal circulation. /
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FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 7.05 '(2.00)
- a. 1404T would be TWO Reactor Coolant Pump indications that a loop isolation-valve was drifting closed? (1.00)
- b. WHAT TWO (2) physical actions should the reactor operator take.if loop A isolation valve.2RCS*MOV591 19 confirmed to be drifting closed? (1.00) 6NSWER 7.05 (2.00)
- a. low flow (0.50 low pump amps (0.50)
- b. trip'the reactor (0.50) trip RCP "A" (0.50)
REFERENCE BVPS 2LP-SQS 53C.1 E.O. 's 4, 5 BVPS - OM AOP 2.6.3 K/A 000017 EA1.10 2.6 K/A 000017 EA1.12 3.1 K/A 000017 GO.11 3.6 000017G011 000017A112 000017A110 ...(KA's) l SRO EXAM REVIEW (12-2-87) 7.05.a At BVPS, RCP amps increase as flow starts to decrease in the RCS, so amps could initially increase as loop stop valve drifts shut. This should be an additional correct answer. See attached RCP performance curves. I
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k ~i FACILITY REVIEW OF WRITTEN EXAMINATION
.OUESTION 7.07 -(2.50)
During a plant startup with reactor power at 12% a loss of 125 VDC from switchboard 2-2 occurs.
- a. WHAT FOUR (4) automatic actions will occur? (2.00)
- b. What would be the most likely cause of a trip during this transient?
(0.50) ANSWER 7.07 (2.50)
- a. -
Feedwater flow control valves (0 ;0) and feedwater bypass flow control valves (0.40) fail closed (0.20) Letdown isolation valve (20HS*A0V204) (0.40) fails closed (0.20) Steam driven auxiliary feedpump starts (0.40)
- b. -
Steam generator it.v level in coincidence with feedflow/steamflow mismatch (0.50) REFERENCE BVPS - OM AOP 2.39.5, p. 1 K/A 000058 EK3.02 4.2 K/A 000058 EA2.03 3.9 000058K302 000058A203 ..(KA's) SRO EXAM REVIEW (12-2-87) 7.07.b The given answer is impossible since it requires a 40% of full flow steam feedflow mismatch and the problem states that the plant is at
- 124 power. The likely cause of the trip would be on low-lov steam 1
generator level (15.5%). i
FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 7.10 (3.00) For the following questions, assume that a main steam line break from steam generator 21B has occurred inside containment.
- a. As directed in BVPS - EOP E-2, "Faulted Steam Generator Isolation, " WHY are_the supply valves to the turbine-driven auxiliary feedwater (AFW) pump closed? (0.65)
- b. Because of elevated secondary radiation levels, the operator transitions to E-3, "Steam Generator Tube Rupture." After asking for steam generator (SG) samples to be taken, the chemistry department informs the operetor that SG 21C has high activity. If the turbine-driven AFW pump is the only available source of feed flow, which SG should be used to supply it with steam? Justify your answer. (1.35)
- c. If pressurizer pressure control is lost and the operator transitions to ECA-3.3. "SGTR without Pressurizer Pressure Control" (Attachment 3), WHAT are TWO (2) reasons for NOT using auxiliary spray at step 3 when normal PRZR spray cannot be established? (1.00)
ANSWER 7.10 (3.00) i i '
- a. minimize RCS cooldown (0.65) ;
OR i to isolate steam from the S/Gs (0.65)
- b. 21C (0.50) 4 faulted SG 218 should remain isolated unless needed for RCS cooldown.
SG 2LA is available, but 21C must be used to steam the AFW pump (0.85)
- c. possible spray nozzle failure (0.50) ,
insufficient pressure differential (0.50) REFERENCE BVPS LP-LRT-VII-65 E.O.s B.3, B.4; LP-LRT-VII 67 E.0 B.3 BVPS - E0P 2.53A.1 Procedure E-2, p. 2; Executive Volume 2.53B.4 Background Information for Procedure E-2, p. 14 BVPS - E0P Executive Volume 2.53B.4, p. 14 , K/A 000038 EK3.06 4.5 K/A 000040 EK3.04 4.7 000040K304 000038K306 . ..(KA's) ! l l
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~ FACILITY REVIEW OF WRITTEN EXAl@ATION SRO EXAM REVIEW (12-2-87) 7.10.b The correct answer should be SG-A since it is neither faulted or ruptured (intact) and all three SG's can supply the turbine driven
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7.10.c An alternate correct answer based on plant conditions should be: the plant is still'in SI configuration and charging and letdown are isolated. l l
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u BVPS - E0P 2.53B.4 rv.,urgy3 golum. E-3 To provide a more rapid means of depressurizing the RCS and rupturad steam ganarators, a post-SGIR cooldown method using steam dump, ES-3.3, was developed. This is the fastest of the three inethods. However, the radiological consequences must be considered particularly if steam dump to the condenser is unavailable. The NRC has indicated that controlled radiological releases from the affected staam generators should not exceed 10 CFR 20 limits. In addition, if water exists in the steamline, steam release may cause wacer hemmer effects resulting in damage to secondary side equipment. Consequently, this method should not be used if water may exist in the main steamlines. Although the three post-SGTR cooldown procedures are presented as alternate methods, they are similar. Consequently, one could begin with the backfill method, continue with the blowdown method, and complete the recovery using steam dump, provided the limitations of each method are observed. Similarly, for multiple - tube failures, one could execute combinations of the three methods at the s ame time. For example, one could depressurize one ruptured steam generator using blowdown and another using the backfill method. These procedure,s provide the flexibility necessary to cool down and depressurize the plant to cold shutdown conditions for a wide variety of SGTR scenarios. The operating shift must evaluate the three methods to establish a preferred post-SGTR cooldown method and prioritize the alternate methods. The actual recovery method must be determined on an event specific basis with consideration of the available equipment and evolution of the event. For example, normal letdown may be needed to complete recovery using the backfill method, ES-3.1, while steam generator blowdown is needed to implement ES-3.2. The procedure ES-3.3 should not be used if water may exit in the main steamlines or steam releases may lead to unacceptable radiological releases.
- 2. Select Faulted or Ruptured Steam Generator For RCS Cooldown In the unlikely event-that no intact. steam generator is available, one must select either a faulted steam generator, i.e., one with a secondary side break, or a ruptured steam generator for cooling the RCS to RHS operating conditions'. This decision should be based upon consideration of the concerns created by each method and an evaluation of the parameters that effect them. A secondary side break leads to uncontrolled steaming of the affected steam generator and possible overcooling of the RCS. Continued feed flow to this steam generator Will increase the amoust of steam discharged and can increase the uncontrolled cooldown of the RCS.
The potential consequences of continuing to feed a faulted steam generator depends on the size and location of the secondary side break. For breaks inside containment, feed flow to the affected steam generator will result in additional discharge to contain=ent and potentially higher containment pressures and temperatures. ISSUE 1 PAGE 13 0F 136 REVISION 2
L
. FACILITY REVIEW OF WRITTEN EXAMINATION OUESTION 8.02 (2.00)
Answer the following questions TRUE or FALSE:
- a. The NSS, NSOF, and the STA (or NCO) all must sign the "Authorization for Removal from Service" lines of the Emergency Safeguards Equipment Clearance Checklist.
- b. Only the NSS needs to sign the Equipment / Radiation Clearance Log for the clearance to become effective,
- c. A Master Clearance can be used to cover maintenance that requires equipment to be operated in order to perform the necessary work.
- d. A Caution Tag may be removed by Test Group Personnel without obtaining NSS/NSOF's permission.
ANSWER 8.02 (2.00)
- a. TRUE
- b. FALSE
- c. FALSE
- d. TRUE REFERENCE DLC SAP Chapter 41, pp. 17, 47, 50; Chapter 42, p. 6 K/A 194001 Kl.02 4.1 194001K102 ...(KA's)
SRO EXAM REVIEW (12-2-87) 8.02.d The correct answer is FALSE. The test crew can only post and/or remove caution tags without NSS/NSOF permission if authorized by an approved test procedure. The question did not state that the test crew member was performing an approved procedure. See attached SAP 42, pp. 2, 3, and 6.
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'n' p,n< )C,N Utilization of' Caution Tags.
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. s 1 i' ',j 1. .; Caution,' Tags' are ' . net to be used for personnel safety- , p .'
. a . pr.otectic7 in111eu of a Red.Dangerl(T g e for .~ performing
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u up ~, , t.A, ,,r vork , ;-on eny plant component. {
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Q'l ; , j ' ; ' . ;O* ' ! ;2. , n.Cautiori i Tags ' are n'ot 'to be'used for lifting leadsLorj.(
.5 ;
' - J' Etemporarily' modifying electrical circuits in lieu l ~of Na-- ,1
- ' ;f Jwaper ' and Lif ted Leads Tag. 1 .
~
- 3. .The yellow Caution Tag may be utilized to alert sta cion ' '
k
' ~
, persennel of an unusual operating condi . ion.
s . ( -
- . . * . *g
- 4. An apparatus tehich has a Caution Tag posted shall not be.
operated until a supervisor in charge of the operation
.is familiar with the condition under which the tag was posted and authorizes steh operation.
. ~i. ; ,
, 5. Yellcw Cautie, stickers should be used on centrois mounted on th. control boards instead of Caution Tags .' -
(Ref: Attac ar..ent 1), when a tag has the potential to obscure indicating lights, nameplates.. annunciators, or indications.
- 6. Any Caution Tag used to fulfill instructions dictated by
.," approved procedure does not need docuoented in' the D, - . Caution ~ Tag Log provided the procedure (s) address.
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Page 2 of 10 Revision 0
y,, y ; 77 ._, w a, - . s , s .. s . ..:, L :
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O a . . t );. .t.. . . . . . . s V cf Nucisar: Group nSitei. Administrative Procedures y , Chapter 42 "n.
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.3 .'t.M'L,?.gincludingi..
W.$ g, g rcturningLthe placement 'and ' removal. system or component of the Caution Tag (s) . to:its normal st - M; ~ s + 'c:7 c . b.T .iThe. procedure number or test number shall be placed %on ,
. a .
Ic.$ 4f @p R. .
-t hy,a. [m fd h{d.f[y~6the trag< fins, v. e. o f,s., +
1'ac the slog . tag;--num.bera;a- +f Qi' %; , i fh 5""~cy n- * %~ ? ' L4 u' % wS O M.M. AW'. n6 p.i % @+< a ! with:a. . w h 'p. ,4 j'. Yf-y4p nt - . . - f.r,1;.w.,3yry
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_a . t e ;, b w N *. . .w 4 A w Caution.?; Tag 'u. sed in: conjunction =
- clearance &.'N+[.doest - . .<,,
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t p notyneed', documented in1theiCaution$Tage,Logi 'sincey-l the ' G'. ' . a' J M ;' l j - i Q. d Caujien%Tygr 2is 4 documented-Gbn W the " Clearance's 7 9 >' [ ,
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b' Q{ W) Valving / Switching.( Procedure .using'J the DForm; g .;; J $
, f- ;D d h C[ ' w" j dp %j % M. gCautionyTag 2,with - a; Caution. Tag. entries, clearancel tthe,in' W- _ ,
-:, this
" < p NT ' J:?f,g.gj).pshall)be; filled.out. as 1 described , ,
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?..yMK:3. fGM i*NfggeXCePt the(Clearance ' Permit, number:shall~ be ' indicatedoon- Caution? Tag?; Log; Index* ,
YdM MI 3 P d D hc sequent'i Alinumber. ' 'i .!e M 0 f 4y Q[5P. k.S o h } Caution 3 4 Tags D N-E utill:ed for ' jf9 .f .' p procedu 1 E 7,9;< ', . . '; 9 %;?wN.8.$y;%)N D3
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M i M Una[n.ed.
.. nn y g e b stored.?for. reuse! .if .the. procedure. is . used y,, ' - .
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s j^qq.*f[9 g Caution f3"agsj aust? bespositioned.such.{that they,7 ojno 7 A 4 1.p gobscurejindicatingt11ghts danterif aces , '. annunciators ,j or,4 ' cd F. s *_ k ; C ,., f f ' ?f 4,i)
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' ' [7 {j, y.u g< p y. 4$Cautiou g[11st ofjTagY; t and1removedifrom pobrinent';cautionQtags ' will: ;bethe' Caution by .the Tag jLog. ;Th kept :
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<t 3s g . ~s 6 N.S.S. + 1 p. + . , . +
J: . ;. #l $. ' .y/111,"g Special;.'. Caution Tag / Sticker markings may be utilized as,
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\',.Edirected'by7the rUnit 'NSOS.. .These specialf markings. - -
fa: ^Jdesig'nate specific *, con'ditions' of annunciator window;or , F ~ f
~ Wcontrol'. switch status as 'shown' on Attach cent 4
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C 12.- l Circumstances which should. require Caution Tagging: _ . r
;. , , 3 - 4
. a. A 'Caut' ion Tag may'be posted on a component to L
, instruct an operator' tc notify siecii c station '
personnel . prior to the operation et the component. -
- b. A' Caution Tag may be used to identify a component that-is in a teeporary configuration or to inform ,
persennel of the use of temporary equipment that is required to support the operation of the tagged component.
- c. A Caution Tag r.3y be used to inform an operator of special additional manual actions that are required to operate the tagged componint.
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, Page 3 of 10 Revisfon 0
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NuclearTGroup - Site 1 Administrative Procedures. S
~' Chapter 42Y { h ,@. ~ ..K
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. , c. 480V.'and 4160V Breakers, install the. tag in' clear 1 .; MW
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n .. ,- 1 , . s c view on the outside of the breaker door.> . < A 2 ,, -. g?c y [ d 1. 9 . J e,Ga
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. %;, .f, w' 'NSS/NSOF's1 permission to . remove y ,.g the! Caution' Tag, J '
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- 2. ~ Remove the Caution Tag ~from the:componenti . . s-3- . ,
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. 3'. k Till out the remaining porticn'of hhd .'ap'plicablesUnit'sjY M.h[. @;% ~
*;< - . . 7Castion Tag' Log' (Ref. Atcachment \ 2)3forithe'lCautiiinsTag M W'd.? O h F f d I. 1 $ N removed,E. including: % i N w; , $ s'h.ipIM@h MM.+QMS.kNM'h@
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reas'en the Jaution Tag,was tremovedJ . r . t t M.rI'- U . '- . ,. .J".s,- f 3 ,K s .- a . . Tht.
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, 'w i.k. . ,. ,' c 1. N The. name of..the .persen c. removing /the Caution Tag. %,, .u .,M.-.',"!. iY. . ,D... .:7 :.c.
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- b. '...,.. ,. "i45 Slice.. . out sche Cautien Tag Log
. . - ladez > for .th2. Caution _.
_';: Tag p @p M-
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- b. . T,he system number.
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; The' item' tagged . '
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D. . Periodic ?"Audit of Caution Ta,gs. v ., y,_ . x.- e .
+. . . . > . 6, ,g c.$_ -
The. Caution Tag Log Review,- (OST 1/2.4F. 6) shall be'
- 1. ;,-
performed to ensure that Operations conform to thel ; , , administrative requirements f.>r Caution Tass. * '
- o. -
- 2. Caution Tagged components inside containment'.tre exempt. [* j.
from the. Caution Tag Log Review when containment ovacuum , is establist ed.
' J-
[
- 3. The Caution . Tag Log Review will be completed prior.to leaving Mode i following an extended mainu. nance outage.
- 4. The review shall be con'ucted d a minimum of 3 times per- P year. -,
~
k. Page 6 of 10 Eevision 0
FACILITY REVIEW OF WRITTEN EXAMINATION OU2STION 8.07 (2.40) For the following questions, utilize Attachment 5.
- a. WHAT would be the effect on the Safecy Injection System during a LOCA is CONTROLLED LEAKAGE had exceeded the Technical Specification limit prior to the event? (0.60) I
- b. Would Limiting Condition for Operation 3.4.6.2 be exceeded if containment sump discharges totaled 625 gallons during the last hour? Justify WHY or WHY NOT. (1.40)
- c. Is reactor-to-secondary leakage included as part of IDENTIFIED LEAKAGE?
(0.40) i I ANSWER 8.07 (2.40) l a. Safety injection #1ow would be less than assumed by the FSAR accicent l analyses (0.60) I
- b. no (0.40) l 600 gallons (10 GPM x 60 mins) of IDENTIFIED LEAKAGE (0.50) l plus 60 gallons of UNIDENTIFIED LEAKAGE (0.50) is the limit for 1 hour
- c. yes (0.40)
PEFERENCE < l BVPS - Unit 2 T.S 3.4.6.2; Definition 1.14; B 3/4 4-4 K/A 006000 G0.05 4.2 K/A 006000 G0.06 4.0 006000G006 006000G005 ...(KA's) SRO EXAM REVIEW (12-2 87) 8,07,b This question should make allowances for assumptions as to where the water was coming from since it wasn't specified in question. If the student assumed that the entire supply was either identified or unidentified, the answer would be yes. They also could have assumed that the water was not from the RCS but from other sources such as main feed, chilled water, CCP, fire header leakage, PG water, etc., in which case the answer would be no, since the T.S. only refers to RCS leakage.
*~
,p 1 i:
ATTACHMENT 4 NRC RESPONSE TO FACILITY- COMMENTS The following statements address the NRC's disposition of comments on the written examination submitted by the facility (see Attachment 2) and changes made to the answer kay during the grading process. REACTOR OPERATOR EXAMINATION
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l.02 Comment accepted. No answer key change required.
1.04.b Comment accepted. No answer key change required. 1.06 Comment accepted. No answer key change required.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.06.a Comment accepted. Answer key changed to reflect additional information provided by the facility.
Inf ormat i on should also be incorporated into training mtterial.
- 3. INSTRUMENTS AND CONTROLS 3.02b Comment accepted. Answer key changed.
&-c 3.04.c Comment accepted. Answer key changed to reflect information provided by the facility. Information should also be incorporated into training material and procedures.
3.06.6 Comment accepted. Answer key changed to reflect information provided by the facility. 3.07.d Comment accepted. Answer key changed to reflect information provided by the facility. Information should also be incorporated into training material and procedures. 3.12.a Comment accepted. Answer key changed to reflect information provided by the facility. Material originally l provided was inconsistent and should be corrected. l l l L.
y E- e
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL 4 '.' 0 9. a Comment ~not. accepted. In both cases one AFD value is 8< c outside .the target band.
'4.11.b-' Comments not accepted. Knowledge of facility ALARA program is covered by the KA Catalog, NUREG-1122, KA:#
~
194001K104, with an importance to safety factor of 3.3. An essential element of the ALARA program should be who 6 is authorized to establish and extend exposure limits. The question neither stated nor implied that emergency i exposure was involved. ! SENIOR REACTOR OPERATOR EXAMINATION
- 5. THEORY OF NUCLEAR POWER PLANT OPERATlON. FLUIDS AND THERMODYNAMICS 5.01 Comment accepted. Blowdown flow was accepted as an j
. alternate answer.
5.04 Comment accepted. Allowances. wore incorporated into the-
- answer key.
5.09 Comment noted. The answer key was changed to eliminate "hot channel factor" as'part of'the required response. I "Thermal limits" or "DNB bounds" are, however, required t for a full credit-answer. i 5.10 ' Comment accepted. , P
- 6. PLANT SYSTEMS DESIGN.' CONTROL AND INSTRUMENTATION ,
, 6.01.a Comment accepted, l , 6.02.a Comment accepted. The answer key was changed to "Containment purge (elevation 702)." j 6.03.a Comment accepted. 6.04.a Comment accepted. The answer key.was changed to accept l alternate answers. 6.06 Comments noted. The answer key for 6.06a. was mouified [ per the facility's request. The answer key for 6.06.c , was modified to accept any of the RCP components as a i correct answer. i t [ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _s
Tf: W 6.09 Comment noted. The answer key was modified as follows, "low DNBR protection CO.153 for control rod drop
, accidents CO.153."
6.11 Comment accepted.
- 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY. AND RADIOLOGICAL: CONTROL 7.01.b Comment not accepted. BVPS enabling objective 2LP-SGS-50.51.52.1 number 2 states the requirement to be able to explain the basis for "Cautions" and "Notes."
7.01.c Comment not accepted. The "Immediate" actions required by BVPS-OM 51 is to refer to BVPS-E.O.P. ES-0.2. 7.05.a Comment accepted. 7.07.b Comment accepted. The answer key was changed to, "low-low steam generator level." 7.10.b Comment accepted. The answer key was changed to, "S/G 21A [0.503 because it is neither faulted or ruptured [0.85]." 7.10.c Comment accepted. The follow).ng alternate answer was incorporated into the answer *< e y , "normal charging flow is isolated."
- 8. ADMINISTRATIVE PROCEDURE _S. CONDITIONS. AND LIMITATIONS 8.02.d Comment accepted.
8.0/.b Comment accepted. Grading was based on the candidates' assumptions. The answer key was modified to incorporate al ternate correct responses.
j ~ t _r i w Attachment 5 NRC COMMENTS ON FACILITY REFERENCE; MATERIAL-Generic
'Many lessonLplans-(LP's) focus only on' differences betw'een units.- +
For future exams either the Unit 2 material should stand alone or , Unit l' references should be-provided. Figures referred'to in LP's as TP's were-not included.
-Specific 1 RHS- LP ' wa s. ve r.y ' sparse. Contained no Learning Objectives (LO's). ;
L
- 2. Reactor Control- LP- provides only LO's. Operating Manual (OM) does not support LO's. . Figures do not match OM references.
- 13. P LP-SQS-1.1 shows 70%, OM-2.1 shows 4%, alarm response OM-2.26 shows 70%. i f
- 4. LP-SOS-26.2. stated' "Same'as Unit 1", but Unit 1.information not provided.
- 5. LP-SQS-24.1 - LP for SG Feedwater does not discuss SG WLC system.
m
- 6. RCS Isolation' Valve Interlock- OM-2.6.3, p 7,9 lists 20 F; ,
OM-2.6.1,p 48 lists 5 F; LP-SOS-6.2.2 lists "predetermined value."
- 7. RIL Curve - Figure in Curve Book is different from Tech ;
Specs. i
- 8. RCP Start Interlock - OM-6.3 states that RCP start requires Tc isolation valve. Closed with loop by-pass isolation' valve
.open; LP-SOS-6.2, p 6 states tnat pump start requires Tc and loop bypass valves both open. ,
- 9. Main Generator and Transformers- No LP provided. .;
- 10. OM References Figure 26-20 for Turbine Runback; actual ,
reference should be 26-22. i
- 11. OM-7.. -
Figure 7-7 shows that either 460A or 460B not open ; will auto close ADV200A,B,LC. LP-SOS-7.1 states that 460A ; and 460B must both be closed to auto close ADV200A,B&C. j
- 12. ' OM-43.1 - References Figures 43-2 thrcugh 43-12 which were ;
L not provided. r b
- .}}