ML20151W011
| ML20151W011 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/14/1988 |
| From: | Nauman D SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Dean B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20151V985 | List: |
| References | |
| NUDOCS 8805030268 | |
| Download: ML20151W011 (49) | |
Text
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. ., gth w Ekctric & Gas Company nA n Cou Nuclear Operations t a,g29218 SCE&G
"" ENCLOSURE 3 March 14, 1988 Mr. Bill Dean License Examiner U.S. Nuclear Regulatory Commission Region II, Suite 2900 101 Marietta Street, N.W.
Atlanta, Georgia 30323
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 License No. NPF-12 Operator License Examinations
Dear Mr. Dean:
Enclosed are the utility comments to the NRC Operator and Senior Operator examinations administered at V.C. Summer on March 7, 1988. Your consideration is appreciated. ,
Very truly yours, h Ag in~
D.A. Nauman GAL:0AN Enclosures cc: (without attachment)
- 0. S. Bradham M. B. Williams A. R. Koon NPCF (with attachment)
K. W. Woodward M. Morgan R. Aiello P.Isaksen(w/atuchment-R0only)
File (814.04) 8805030268 880427 hDR ADOCK 05000395 DCD
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REACTOR OPERATOR EXAM QUESTIONS 1
QUESTION:
1.15 List 3 purposes for the Rod insertion Limits.
ANSWER KEY RESPONSE:
- 1. Adequate SDM upon trip
- 2. To minimize the amount of positive reactivity inserted during an ejection accident, and
- 3. To minimize radial flux telt (peaking)
REFERENCES.
VCS, RT B K lli, RT- 14, P. 20 SUGGESTED CORRECT RESPONSE:
Either the above answer, of To ensure that:
- 1) Acceptable power distribution limits are maintained,
- 2) The minimum SHUTDOWN MARGIN is maintained,and
- 3) The potential effects of rod misalignment on associated accident analyses are limited.
REASON:
Although the answer key response lists the "standard" reasons for maintaining rods above the Rll, the alternate answer is found in V.C. Summer Plant Technical Specifications, page B 3/41-3 (attached) as the bases for control rod insertion limits.
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, , Sub l. tS REACTIVITY CONTROL SYSTEMS BASES 80 RATION SYSTEMS (Continued)
MARGIN from expected operating conditions of 1.77% delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs from full power equilibrium xenon condi-tions and is satisfied by 12475 gallons of 7000 ppe borated water from the boric acid storage tanks or 64,040 gallons of 2300 ppe borated water from the refuel-ing water storage tank.
With the RCS temperature below 200*F, one injection system is acceptable without single failurs consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be OPERA 8tE and the Surveillance Requirement to verify all chartling pumps except the required OPERA 8LE pump to be inoperable below 775'F prov des assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
The boron capability required below 200*F is sufficient to provide the required SHUTDOWN MMtGIN of 1 percent delta k/k or as required by Figure 3.1-3 af ter xenon decay and cooldown free 200*F to 140*F. This condition is satisfied by either 2000 gallons of 7000 ppe borated water free the boric acid storage tanks or 9690 gallons of 2300 ppe borated water from the refueling water storage tank.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
The OPERA 8ILITY of one boron injection system durinti REFUELING ensures that this system is available for reactivity control whi's in M00E 6.
3/4.1.3 MOVARI CONTR04. ASSDSLIES The specifications of this section ensure that (1)'ecceptable power .
distributten~1tsits~are-enintained. (2)-theminious SWTOOW MMGIN is main-tained, and-(3)_1iett thopetentiai~effacts'sf~ red ~aisaiigneont on associated accident analyses. OPERASILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.
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QUESTION:
2.08(b) Other then relief valves, what feature prevents overpressurization of the CCW System if a thermal barrier heat exchanger tube ruptures? (include setpoint,if applicable).
ANSWER KEY RESPONSE: -
Thermal barrier isolation valve isolation automatically closed (0.5 points) when respective downstream flow exceeds 65 gpm (0.5 pts.). (1.0)
REFERENCE:
VCS,18-2, CCW System, p 4 & 14 SUGGESTED RESPONSE:
As given above with 0.8 points for action and 0.2 points for setpoint.
REASON: Knowing that the CCW system is protected from RCS pressure upon thermal barrier railure by the closure of the isolation valve on high flow is far more important than knowing the associated setpoint. Only reactor trip, safety injection, permissive and control interlocks, steam and feedwater isolation signals. R8 spray signal, and other SSPS setpoints should carry such high weighting. An 80%/20% split between action and setpoint is more amenable for a feature such as this.
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QUESTION:
2.13 (a) Describe TWO automatic actions associated with the instrument air system l
which serve to mitigate a loss of air pressure. Include any associated setpoints.
ANSWER KEY RESPONSE:
- a. 1.
- 2. Service air compressor Standby air isolated (0.5) at 90 starts 0.5)p(sig at 70(0.25) (IPV-8324) psig (0.25)(if lead compressor breaker is closed).
REFERENCE:
SUGGESTED ADDITIONAL RESPONSE: )
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As given above with 0.6 for action and 0.15 for setpoint.
- REASON: Knowing that service air isolates is far more significant than knowledge of l
the setpoint at which the action occurs. Only reactor trip, safety injection, permissive and control interlocks, steam and feedwater isolation signals, RB l spray signal, and other SSPS setpoints should carry such high weighting. An 80%/20% split between action and setpoint is more amenable for a feature of this nature.
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QUESTION:
2.17 What automatic action (s) should occur to the Reactor Building Cooling Units upon receipt of an SI AS signal? Assume an initial normal at power lineup.
ANSWER KEY RESPONSE.
All fans would start / shift to low speed (0.5) and service water system booster pump starts automatically (0.5)
REFERENCES:
VCS, TS 3.6.2.3 and AB-9 SUGGESTED RESPONSE:
Selected fans (1 in each train) would start / shift to low speed (0.5) and service water system booster pumps start automatically (0.5) (Industrial cooling water supply to RBCU's isolates).
REASON: Main control board selector switches allow the operator to choose one fan in Train A and one in Train 8 for the RBCU's. Only these fans will shift to slow speed upon a safety injection signal. See attached drawing showing selector '
switches for each train.
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QUESTION:
2.21 Which statement regarding the Main Generator Protection System is NOT CORRECT.
a) Opening the generator output breakers ALWAYS results in a turbine trip when the generator is loaded.
b) Once the generator is loaded, a turbine trip ALWAYS results in a generator trip.
c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a reactor trip.
d) A reactor trip ALWAYS results in a turbine trip.
ANSWER KEY RESPONSE:
a) Opening the generator output breakers ALWAYS results in a turbine trip when the generator is loaded.
REFERENCE:
VCS, IC-9, P. 44, 59 ; TB-5, P. 87 S!JGGESTED ADDITIONAL RESPONSE:
(c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a reactor trip. i REASON: While choice (a) is certainly a correct answer to the question, choice (c) is also correct. The question asked which choice is NOT TRUE. At V.C. Summer, a turbine trip above P-7 will not always result in a reactor trip. At one time this was the case. However, now, the P 9 interlock (>S0% power) enables the reactor trip upon turbine trip. This is documented by the attached references, IC 8, Nuclear instrumentation, p. 29, and IC-9, Reactor Protection and Logic, p. 45.
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Goh 2. .y more of tne four power range chann@ls exceed the tr ip setpoint, the bistable will cause a reactor trip and actuate the reactor trip first-out annunciator "PR FLUX HI SETPT HI" on C8P 10. This trip function cannot be blocked. I o P-8 permissive circuit.
When the outputs of two of four power range channels exceed the P 8 setpoint (38 percent power), a loss of flow, 'i as sensed by two of three flow transmitters in any one loop, will cause a reactor trip.
Between readings of 38 percent power (P-8) and 10 percent power (P-7), flow must be lost in at least two loops to cause a reactor trip. Below P-7 there are no loss of flow trips. As each power range channel exceeds 38 percent oower, the respective "CHAN I (II, !!!, IV) P8" light on the reactor pemissives panel energizes. When two of four channels exceed 38 percent power, the "P8" light on the reactor pemissives panel on CSP 9 energizes, o P-10 permissive circuit. This permissive is automatically enabled when two of four power range channels exceed 10 percent. Actuation of P-10 allows the manual blocking of the power range low setpoint trip, the intermediate range trip, and the intermediate range rod stop.
It also blocks the source range trip and provides a portion of the P-7 signal .
As each power range channel exceeds 10 percent power, the respective "CHAN I (II, III, IV) P10" light on the reactor pemissives panel energizes.
When two of four channels exceed 10 percent power, the "P10" light on the reactor permissives panel on C8P 9 energizes.
The "P7" light also energizes because its logic is satisfied den either P-10 or P-13 (1/2 turbine impulse pressure greater than 10 percent) is satisfied.
o P-9 permissive circuit. As reactor power exceeds 50% indication for two of the four power range channels, a turbine trip will activate a reactor trip.
At power levels less than 50% (3/4), the reactor trip resulting from a turbine trip will be blocked.
The differential amplifier provides an output signal proportional to the rate of change of reactor power to two separate bistables.
The first bistable will trip when a sudden increase in neutron flux occurs
(+5 percent of full power within 2 seconds). When this bistable trips, it l
will actuate "CHAN I (II, Ill, IV) PR FLUX RATE HI" on the reactor bistable 29 l 1068S:4 i
Qo W.o n Ltg A separate permissive, P-8, will automatically block' the trip that occurs when flow is lost in ona loop. This occurs when thrce of four power range channels are below the setpoint of 38 percent. Again, the reactor is adequately pro-tected below the setpoint without the trip. If allowed in the future, it would be possible to continue limited power operation with an inoperable loop. The AND gate of the P-8 permissive logic denoted with a 2 of 4 coincidence actually energizes above the P-8 setpoint and the NOT gate will extinguish the permis-sive status light. To deenergize the AND bistable, 3 of 4 power range channels ' '
must be below the P-8 setpoint. The P-8 permissive status light energizes and the zero input to the AND gate of the low flow logic blocks a reactor trip frcm a loss of flow in one loop. P-7 works in a similar manner. P-9 will prevent the reactor trip that normally results from a turbine trip. The setpoint for this additional permissive is 50% power with a coincidence of 2 out of 4 power range channels.
Rod Stops (Sh. 4 and Sh. 9)
The C-1 high neutron flux rod stop will block both automatic and manual rod withdrawal. This occurs if the oparator does not block the rod stop prior to either a normal or an unexpected power increase. Any out motion is then stopped if one of two intermediate range channels exceeds an amperage output that is equivalent to 20 percent of rated thermal power ( =10-Samps). This rod stop is manually blocked at the same time the operator blocks the inter mediate range high flux reactor trip. In fact, the same switches that are used to block the trips are also used to prevent the rod stop. Again, attempts to block the C-1 function prior to 2 of 4 power range channels exceeding P-10 will be unsuccessful. The rod stop is automatically reinstated when 3 of 4 power range channels fall below 10 percent.
The C-2 overpower rod stop also blocks automatic and manual control rod with-drawal . The block action occurs when one of four power range channels rises above 103 percent power. This is the only function associated with the Reactor Protection System that actuates on a 1 of 4 coincidence. Because of
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this, special switches called Rod Stop Bypass Switches are provided on the flux deviation, miscellaneous control and indication drawer of NIS. These switches allow continued normal rod withdrawal if a power range channel should fail high or if a channel should require testing.
45 11555:4 Rev 1 5/85
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QUESTION.
3.04 What plant conditions will cause the Cold Overpressure Protection System (COPS) alarm to actuate?
ANSWER KEY RESPONSE:
Low auctioneered wide range T,, or T, less than 350*F AND any RHR Suction valve not fully open.
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REFERENCE:
VCS,IC-3, PZR PRESSURE AND LEVEL CONTROL SYSTEM, p. 33 l
SUGGESTED CORRECT RESPONSE:
Any wide range T,, or T,less than 300 F AND any RHR Suction valve not fully open.
REASON:
Annunciator Response Procedure, ARP-001 XCP 610 (attached) clearly specifies the conditions which cause actuation of the alarm. This is backed up by Technical Specification 3.4.9.3 (attached), which is the basis for the alarm.
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. . ARP-001 REVISION 3 1 1/22/35 PANEL XCP-0610 (NSSS) l ANNUNCIATOR POINT 2-4 RCS TEMP LO AND RHR SUCT SETPOINT: ,
VLVS NOT OPN ORIGIN RCSEtemp'<300*F and! TY/410J ary -RHR suotion valve TY/413K rnot~ fully'open 33X-RH04 PROSA3LE CAUSE: 33X-RH06
- 1. Failed low temperature instrument.
2.
RRR suction valve (8701A, 8701B, 8702A, 8702B) not fully open. ;
I AUTOMATIC ACTIONS None ,
I IMMEDIATE ACTIONS:
- 1. I
- 2. Verify RCS temperature < 300*P as read on ITI-410 and ITI-413.
Verify RHR suction valves 8701A, 8701B, 8702A, and 87028 fully open. .
SUPPLEMENTAL ACTION: '
- 1. 1 Ensure by at least one an onservice RRRtrain Cold relief suction Overpressure valve. Protection being provided
- 2. Refer to Technical Specification 3.4 9 3.
REPERENCES:
- 1. GAI DWG. B-208-082, sh. 73 l
- 2. GAI DWG. B-208-084, RH-03
- 3. GAI DWG. B-208-084, RH-04 4
GAI DWG. B-208-084, RH-05
- 5. 3AI DWo. B-208-084, RH-06
- 6. V.C. Summer Technical Specifications
- 7. SOP-115 PAGE 11 of 17
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ADDLICA8ILITY:
M0004 ween the temoerature or ~
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e ; 1-c any c:r ecti.e action necessary to :revent r et i- 1 ;
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QUESTION:
3.06(b) After resetting safety injection, automatic actuation is inhibited until s J at signal is cleare d ? (0.5)
ANSWER KEY RESPONSE:
- b. Rx Trip 8krs closed (P-4)
REFERENCES:
VCS, IC-9, RPS, p 52. EOP-1.2,1.4 SUGGESTED RESPONSE: l
- b. P-4 must be cleared. (It is cleared by closing the reactor t;ip breakers).
REASON: The intent of the answer key appears to be correct. The question asked only ,
"what signal must be cleared" However, to provide clarification, a logic l drawing is attached showing that the P-4 signal must be removed in order to <
enable a subsequent auto Si signal. The P-4 signalis cleared by closing the l reactor trip breakers ;
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QUESTION:
3.18 a. What are the FIVE inputs to the CCM? Be specific.
ANSWER KEY RESPONSE: '
Wide range Tc i Wide range Th incore TC s Narrow range PZR press.
Wide range RCS hot leg press. (0.25 each)
REFERENCES:
VCS, IC-12, CCM, p 4,5,10. EOP-1.2,1.4 SUGGESTED CORRECT RESPONSE- l l
Wide range Tc l Wide range Th incore TC s Narrow range PZR press. ;
Wide range RCS press. (0.25 each) l REASON.
I The only wide range pressure transmitters on the RCS (PT 402A, PT-403A) are j located on the loop A and C RCS hot legs (see attached drawings). Therefore, requiring the words "hot leg" in conjunction with the wide range pressure j instruments is both redundant and unnecessary.
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QUESTION:
4.17 a. What are the V.C. Summer Nuclear Station Administrative exposure limits for Whole Body, Skin, and extremities for occupational workers? (exclude fertile females)
ANSWER KEY RESPONSE:
Skin 6 rem per qtr (0.25)
Ex 12 rem per qtr (0.25)
REFERENCES:
VCS Radiation Fundamentals, p. 9,1017 SUGGESTED CORRECT RESPONSE:
WB 1 rem per qtr. (0.33)
Skin 6 rem per qtr. (0.33)
Ex 12 rem per qtr. (0.33)
REASON: Plant limits are expressed in terms of maximum quarterly exposure.
Requiririg 4 rem per year as well as 1 rem per qtr. is redundant. Neither the skin or extremities limits were extrapolated to yearly values. Quarterly limits .
are more restrictive and should be all that is required. Deletion of this item and reassignment of points to one-third point for each limit is requested.
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[ _ _ _ _ _ _ _______ _ _ _ - - - - - - - - - - - - -
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l SENIOR REACTOR OPERATOR EXAM QUESTIONS l
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QUESTION:
5.07(B) State the change (INCREASES, DECREASES, or REMAIN THE SAME) in the magnitude of the fuel temperature coefficient (FTC) for each of the following:
B. Aging of the core ANSWER KEY RESPONSE:
B. Decreases (0.5)
REFERENCE:
VCS, RT B K lil, R T-11, P24 SUGGESTED RESPONSE:
B. Increases or Decreases (0.5) or Remains the same REASON:
While generally, it is established that the fuel temperature coefficient becomes more negative (increases in magnitude) over core life, this concept is often confused with the behavior of Doppler power coefficient, which decreases with core age. This seemingly contradictory relationship is due to the buildup of Pu-240 as the core ages (see attached reference RT-10, Reactivity and Fuel Temperature Effects, p. 32, 33). However, power coefficient decreases due to overall fuel temperature running relatively cooler at EOL vs. BOL ,
it is requested that either answer be acceptable due to the attached figures FND-RF-224 and VS FND-20 from reference RT-10. Figure VS FND 20, "Doppler Temperature Coefficient at BOL and EOL" graphically underscores this. Depending on the effective fuel temperature, the EOL value can be more negative, the same as, or less ne gative then the BOL value.
e.oh 5,o, W i
Variatiqns in Donoler Power Coefficient Over Core Life Over corG life, the Doppler power coefficient and Doppler only power defect are affected by four phenomena. Fast fissioning of U-238 and conversion of U-238 to Pu-239 decreases the quantity of U-238 and subsequently, the Doppler effect. The effect of the decrease in U-238 is nearly of fset by the production of Pu-240 from the conversion of Pu-239. Pu-240 has a single resonant peak of l
4 approximately 100,000 barns. This peak tends to make the Doppler temperature coefficient more negative. Refer to Figure FNO-RF-234.
Three other important competing factort influencing the effective fuel temperature as a function of power with core life are helium gap conductivity, fuel densification, and clad creep. j Initially the fuel rods are pressurized with helium gas. The gap l between the pellet and clad possesses a given gap thermal conductivity coefficient. However, fission gases such as xenon and krypton are produced. These gases tend to pollute the helium gas, causing a reduction in the gap conductivity coef ficient, resulting ,
in an increased fuel temperature change for any given power change.
i Fuel densification is an effective decrease in the dimensions of the fuel pellets. This change in pellet dimension causes an l
increase in the fuel-to-gap dimension, and a icwer gsp thermal !
conductivity coefficient. Hence, fuel densification tends to increase the Doppler power coefficient is magnitude. However, the '
densification rate slows with core age, and the clad creep factor overcomes the effects of both fuel densification and gap conductivity early in core life.
Clad creep is the predominant factor affecting the effective fuel temperature at a given powcr level. Clad creen is the phenomena in which the fuel cladding effectively shrinks and comes in closer contact with the fcel. The rate of clad creep is a 32 0026V
Q ,. J. . e 5,o, ( b) function of the irradiation and the differential pressure across the clad. In this process, the clad inner diameter tends to decrease and the clad tends to wrap itself around the fuel pellet.
The net result is a decrease in the fuel-to-clad gap dimension, an increase in the gap thermal conductivity coefficient, and a marked decrease in effective fuel temperature over core life.
I The decrease in the rate of effective fuel temperature change with power
( h is the predominant ' actor affecting the Doppler power coefficient. Recall that:
h= (h) x (h)
Doppler Power Doppler Temp. Effective Fuel Coef ficient Coefficient Temperature Change (pcm/% power) ( pcm/
- F) With Power (*F/% power)
Therefore. at EOL the magnitude of the Doppler power coefficient is always less than that at B OL . Refer to Figures FND-RF-224 and VS-FNO-3.
Variation in Doppler Power Defect with Core Life g !p/I Since the magnitude of the Doppler _ oower coeffici nt is consistently less at EOL, the magnitude of thMppler hwer) defect 1 al so less. However, when considering the total DolihiTer oniy 1
power defect, the decrease between BOL and EOL is generally not significant. Refer to 'igure VS-FNO-4.
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0 0 20 40 60 80 10 0 POWER LEVEL (% FULL POWER)
EFFECTIVE FUEL TEMPERATURE AS A FUNCTION OF POWER FND-RF-224 REV I
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FIGURE VS-FND-20 Deppler Temperwure Coe#Islere et SOL and EOL Rev. 2 Cycle 1 j i
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r QUESTION:
5.15(b) What is the SDM range for modes 3,4, and 5? . , l
, -ANSWER KEY RESPONSE:
i i 1770 pcm to 2000 pcm ( 130 pcm).
REFERENCE:
j VCS, TS, p. 3/41-1-1-3 4
j SUGGESTED RESPONSE: Delete question, part b.
REASON: Examiner changed question during administering of' exam. Examiner i acknowledged that clarity of wording is a significant problem on this question.
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QUESTION:
6.01 Which statement regarding the Main Generator Protection System is NOT CORRECT.
a) Opening the generator output breakers ALWAYS results in a turbine trip when the generator is loaded.
b) 0,nce the generator is loaded, a turbine trip ALWAYS results in a generator inp.
1 c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a reactor trip.
l d) A reactor trip ALWAYS results in a turbine tnp.
ANSWER KEY RESPONSE:
a) Opening the generator output breakers ALWAYS results in a turbine trip {
when the generator is loaded.
REFERENCE:
l VCS, IC-9, P. 44, 59 ; TB 5, P. 87
)
SUGGESTED ADDITIONAL RESPONSE:
(c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a reactor trip.
REASON: While choice (a) is certainly a correct answer to the question, choice (c) is also
)
correct. The question asked which choice is NOT TRUE. At V C. Summer, a turbine trip above P-7 will not always result in a reactor trip. At one time this was the case. However, now, the P 9 interlock (>S0% power) enables the reactor trip upon turbine trip. This is documenteci by the attached )
references,IC-8, Nuclear Instrumentation, p. 29, and IC-9, Reactor Protection and Locic, p. 45.
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I more of the four power range channels exce0d the trip setpoin ;, the
(
bistable will cause a reactor trip and actuate the reactor trip first- I out annunciator "PR FLUX HI SETPT HI" on CBP 10. This trip function cannot be blocked.
o P-8 permissive circuit. When the outputs of two of four power range channels exceed the P-8 setpoint (38 percent power), a loss of flow, as sensed by two of three flow transmitters in any one loop, will cause a reactor trip.
Between readings of 38 percent pwer (P-8) and 10 percent power (P-7), flow must be lost in at least two loops to cause a reactor trip. Below P-7 there are no loss of flow trips.
As i
each power range channel exceeds 38 percent power, the respective !
"CHAN I (II, III, IV) P8" light on the reactor permissives panel energizes. When two of four channels exceed 38 percent power, the "P8" light on the reactor pemissives panel on C8P 9 energizes.
o P-10 permissive circuit. This permissive is automatically enabled when two of four power range channels exceed 10 percent. Actuation i of P-10 allows the manual blocking of the power range low setpoint !
trip, the intermediate range trip, and the intemediate range rod stop. It also blocks the source range trip and provides a portion of the P-7 signal . As each power range channel exceeds 10 percent power, i the respective "CHAN I (II, III, IV) P10" light on the reactor pemissives panel energizes. When two of four channels exceed 10 percent power, the "P10" light on the reactor permissives panel on C8P 9 energizes.
The "P7" light also energizes because its logic is satisfied when either P-10 or P-13 (1/2 turbine impulse pressure greater than 10 percent) is satisfied, o
P-9 permissive circuit. As reactor power exceeds 50% indication for two of the four power range channels, a turbine trip will activate a reactor trip. i At power levels less than 50% (3/4), the reactor trip resulting from a turbine trip will be blocked.
The differential amplifier provides an output signal proportional to the rate of change of reactor power to two separate bistables.
The first bistable will trip when a sudden increase in neutron flux occurs
(+5 percent of full power within 2 seconds). When this bistable trips, it will actuate "CHAN I (II, III, IV) PR FLUX RATE HI" on the reactor bistable 29 10685:4
co .b lo )
A separate permissive, P-8, will automatically block the trip that occurs when flow is lost in one loop. This occurs when three of four porer range channels are below the setpoint of 38 percent. Again, the reactor is adequately pro-tected below the setpoint without the trip. If allowed in the future, it would be possible to continue limited power operation with an inoperable loop. The !
AND gate of the P-8 permissive logic denoted with a 2 of 4 coincidence actually I energizes above the P-8 setpoint and the NOT gate will extinguish the permis-sive status light. To deenergize the AND bistable, 3 of 4 power range channels '
must be below the P-3 setpoint. The P-8 permissive status light energizes and the zero input to the AND gate of the low flow logic blocks a reactor trip fran {
a loss of flow in one loop. P-7 works in a similar manner. P-9 will prevent the reactor trip that normally results from a turbine trip. The setpoint for this additional permissive is 501( power with a coincidence of 2 out of 4 power range channels, i
Rod Stops (Sh. 4 and Sh. 9) i l
The C-1 high neutron flux rod stop will block both automatic and manual rod l withdrawal. This occurs if the operator does not block the rod stop prior to t
either a normal or an unexpected power increase. Any out motion is then stopped if one of two intermediate range channels exceeds an amperage output that is equivalent to 20 percent of rated thermal power ( =10-5 amps ) . This rod stop is manually blocked at the same time the operator blocks the inter mediate range high flux reactor trip, in fact, the same switches that are used to block the trips are also used to prevent the rod stop. Again, attempts to block the C-1 function prior to 2 of 4 power range channels exceeding P-10 will be unsuccessful. The rod stop is automatically reinstated when 3 of 4 i
power range channels f all below 10 percent. !
The C-2 overpower rod stop also blocks automatic and manual control rod with-dr awal . The block action occurs when one of four power range channels rises above 103 perec3 power. This is the only function associated with the Reactor Protaction System that actuates on a 1 of 4 coincidence. Because of '
this, spr-cial switches called Rod Stop Bypass Switches are provided on the flux deviation, miscellaneous control and indication drawer of NIS. These switches allow continued normal rod withdrawal if a power range channel should fail high or if a channel should require testing.
45 11555:4 Rev 1 5/85
1 QUESTION:
6.07 Match the RCS penetration in Column A with the appropriate RCS loop segment in Column B (Answers may be used more than once).
Column A
- a. Normal Letdown
- e. RHR Suction 4
ANSWER KEY RESPONSE:
(a) Loop A Cold leg -(1)
(e) Loop A Hot Leg - (2)
REFERENCES:
VCS, P&lD, RCS, DWG F-21 1 ABS, AB-2, FIG. AB2.3 AB-2, p.16-17 SUGGESTED CORRECT RESPONSE:
(a) Normal Letdown -
Loop Aintermediateleg(3) I (e) RHR Suction -
Loop A hot leg (2) and Loop C hot leg (8)
REASON: Per attached reference drawings E-302 601 and E-302-673, normal letdown penetrates the Loop A intermediate leg. RHR suction penetrates the Loop A and Loop C hot legs.
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t QUESTION:
6.10 What signal, other than an 51 signal, must be present to initiate an AUTOMATIC switchover of the RHR system from the RWST to the reactor building recirculation sumps? Give setpoint(s)and coinciaence(s),if applicable.
I ANSWER KEY RESPONSE:
AND (0.3) 2/4 (0.3) RWST level (0.3)less than 18% (0.3).
REFERENCES:
SUGGESTED CORRECT RESPONSE:
This question was amended during administering of the exam. Originally worth 1.5 pts, it asked for both signals. Since the proctor gave the examinees the 51 signal after amending the question, we request deletion of the word "AND" from the answer key, and assignment of points as below: ,
2/4 (0.5) RWST level (0.5)less than 18% (0.5)
REFERENCE NRC Answer Key 4
4 1
5
. o QUESTION:
6.14 What plant conditions will cause the Cold Overpressure Protection System (COPS) alarm to actuate?
ANSWER KEY RESPONSE:
Low auctioneered wide range T,, or T, less than 350*F AND any RHR Suction valve not fully open.
REFERENCE:
VCS,IC-3, PZR PRESSURE AND LEVEL CONTROL SYSTEM, p. 33 SUGGESTED CORRECT RESPONSE:
Any wide range T,, or T, less than 300"F AND any RHR Suction valve not fully open.
REASON:
Annunicator Response Procedure, ARP-001 XCP 610 (attached) clearly specifies the conditions which cause actuation of the alarm. This is backed up by Technical Specification 3 4.9.3 (attached), which is the basis for the alarm.
_ _ ~ . . . - ... - - - . -. - . __
4
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ARP-001 !
REVISION 3 1/22/85 ;
PANEL XCP-0610 (NSSS)
ANNUNCIATOR- POINT 2-4 t
AND RRR SUCT SETPOINT:
VLVS NOT OPN ORIGIN
! RC3"toepT(300*FCand TY/410J any.RER.suotion: valve TY/413K
- not. fully ~open 33X-RH04
- PROBABLE CAUSE
- 33X-RH06 i
- 1. Failed low temperature instrument.
2.
RHR suction valve (8701A, 87018, 8702A, 87023) not fully open.
i 1
AUTOMATIC ACTIONS None IMMEDIATE ACTIONS:
1.
- 1. Verify ROS temperature < 300*F as read on ITI-410 and ITI-413. '
Verify RHR suction valves 8701A, 8701B, 8702A, and 87028 fully open.
SUPPLEMENTAL ACTION:
1.
Ensure by an onservice at least one train RHR Coldrelief suction Overpressure valve. Protection being provided
- 2. Refer to Technical Specification 3.4.9 3 -
REFERENCES:
1
' l
- 1. GAI DWQ. B-208-082, sh. 73
, 2. GAI DWG. B-208-084 RH-03
- 3. GAI DWQ. 3-208-084, RH-04 .
1
- 4. GAI DWG. 3-208-084, RH-05
- 5. GAI DWG. B-208-084 RH-06 l
- 6. V.C. Summer Technical Specifications
) 7. SOP-115 t
1 4
i 5
PAGE 11 of 17 i
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l QUESTION:
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6.15 With the pressurizer level control switch in Position 2, describe how a high failure ,
of LT 459 will affect actual pressurizer level. Assume normal charging and letdown '
system lineups exist and no operator actions are taken. Continue the description l
until pressurizer levelis constant or a reactor trip occurs, include setpoints where -
applicable. l ANSWER KEY RESPONSE:
1 (B/U heaters come on). Charging flow reduced to minimum. Level decreases (0.25). '
At 17% level, letdown isolated (0.25). (Heaters turn off). Level increases. High level af arm at 70% (0.25). High level trip at 92% (0.5).
NOTE: setpoints worth 0.1 points each.
REFERENCES:
1 VCS, Control 5ystem Failure Analysis, IC 15, P. 27. 50P 401.6, P.2 1 SUGGESTED RESPONSE: 1 (8/U heaters come on). Charging flow reduced to minimum. Level decreases (0.25). ,
At 17% level, letdown isolated (0.25). (Heaters turn off). Level increases (0.25). I High level trip at 92%.
Delete "High level alarm at 70%"
REASON: Alarms were not solicited as part of the question. The question asked only how actual pressurizer levelis affected. Had alarms been solicited, this and several other alarms would have been required for a complete answer.
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i QUESTION:
List 3 purposes for the Rod Insertion Limits.
ANSWER KEY RESPONSE:
4 1 1. Adequate SDM upon trip 4
- 2. To minimize the amount of positive reactivity inserted during an ejection accident, and
- 3. To minimize radial flux tilt (peaking)
REFERENCES:
l VCS, RT BK 111, RT-14, P. 20
] l SUGGESTED CORRECT RE SPONSE:
Either the above answer, of:
j To ensure that:
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- 1) Acceptable power distribution limits are maintained, 1 2) The minimum SHUTDOWN MARGIN is maintained, and j 3) The potential effects of rod misalignment on associated accident analyses are limited.
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I i REASON:
) Although the answer key response lists the " standard" reasons for maintaining '
j rods above the Ril, the alternate answer is found in V.C. Summer Plant Technical ;
j Specifications, page B 3/41-3 (attached) as the bases for control rod insertion i
limits '
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REACTIVITY CONTROL SYSTEMS l
BASES 80 RATION SYSTEMS (Continued)
MARGIN from expected operating conditions of 1.77% delta k/k or as required by Figure 3.1-3 after xenon decay and cocidown to 200*F. The maximum expected boration capability requirement occurs free full power equilibrium xenon condi-tions and is satisfied by 12475 gallons of 7000 ppe borated water from the boric acid storage tanks or 64,040 gallons of 23M ppe borated water from the refuel-ing water storage tank.
With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be OPERA 8LE and the Surveillance Requirement to verify all chartling pumps except the required OPERA 8LE pump to be inoperable below 275'F prov' des assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
The boren capability required below 200*F is sufficient to provide the required SHUTDOWN MARGIN of 1 percent delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown free 200*F to 140*F. This condition is satisfied by either 2000 gallons of 7000 ppe borated water free the boric acid storage tanks or 9690 gallons of 2300 ppe borated water from the refueling water storage tank.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
The OPERA 81LITY of one boron injection systes durinji REFUELING ensures that this systes is available for reactivity control whi e in N00E 6.
i j 3/4.1.3 NOVA KE CONTROL AS$DSLIES i
i The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minieue SHUT 00WN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated l
! accident analyses. OPERASILITY of the control rod position indicators is !
l required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.
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QUESTION:
7.11 What are the 2 Immediate Corrective Actions required if the charging flow control valve (FCV-122) fails closed while in automatic control and will not respond to a manual open signal?
ANSWER KEY RESPONSE: '
- 1. Reduce letdown to 45 gpm
- 2. Control P2R level using (local) FCV-122 bypass valve (XVT 8403)
REFERENCES:
VCS, SOP 102, CVCS l SUGGESTED CORRECT RESPONSE:
I Either the above answer, or:
- 1. Isolate letdown
- 2. Place excess letdown in service l
REASON: This question appears to be outdated. At one time, SOP 102 included specific ,
actions to be taken in the event that FCV 122 failed closed. However, this is l no longer the case. There is no SOP off normal which addresses this failure in '
the current SOP revision. Therefore, the alternate answer should be ,
acceptable as well as any other answer which demonstrates integrated I knowledge of the CVCS system with regard to the failure. j 1
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QUESTION:
7.13 Other than identifying and correcting the malfunction, what == the 3 Immediate I Corrective Actions for a diesel generator trip following ar. emergency start. l l
ANSWER KEY RESPONSE' l
- 1. Reset mechanical overspeed,if applicable. l
- 2. Momentarily place EXCITER switch in SHUTDOWN.
- 3. Depress GEN RELAYS RESET pushbutton.
) ]
REFERENCE:
VCS, SOP-306. EM ERGENCY DIESEL GENER ATOR, p. 4,5,24,2 5
_SUGG ESTED CORRECT RESPONSE:
Procedurally, none exists. Request deletion of the question.
REASON: None of the referenceslisted matches the answer ke action which specifically addresses this failure. y.This There is no procedural question appears outdated because, formerly, these actions were listed in an earlier revision of )
SOP-306 They have since been removed and no procedure has been developed in its place with regard to this particular failure. l l
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f QUESTION- f l
7.17 in accordance with EOP-1.3,"Natural Circulation Cooldown", what are the five (5) parameters (conditions) that support or indicate natural circulation flow during .
cooldown? Which way should they be trending? l ANSWER KEY RESPONSE:
- 1. Differential Temperature (0.2) 25 80% of full power (byWR RTD) (0.2)
- 2. 5/G Pressure (0.2) Stable or decreasing (follow RCS temp)(0.2) l 3. Hot Leg RTD (0.2) Stable or decreasing (0.2)
- 4. Core Exit T/C (0.2) Stable or decreasing (0.2)
- 5. Cold Leg Temp (0.2) Near saturation temp. for l S/G press OR constant or slowly decreasing (0.2)
REFERENCE:
l l VCS, TH SCl, TS-14, P 14-24 GENERAL PHYSICS, HTFF, P 356 VCS, EOP-1.3 SUGGESTED CORRECT RESPONSE:
Responses # 2,3,4 and 5 are correct.
Response #1 is incorrect. The correct response is "RCS Subcooling based on core exit thermocouples > 30*F" REASON: EOP 1.1, "Reactor Trip Recovery", pp 910 (attached) clearly lists those five conditions which should be present if natural circulation exists. Although differential temperature will behave as the answer key describes, it is not '
used for natural circulation cooldown verification by the EOP's.
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E0P-1.1 l REVISION 2 :
t ACTION ALTERNATIVE ACTION !
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CAUTION 13
- 4 On natural circulation, RTD bypa.ss temperatures * -
and associated interlocks will be inaccurate. *
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i NOTE 13 ..
Priority should be given to starting RCP A or C.
i i a. Refer to SOP-101, i j '
SYSTEM". !
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- b. If an RCP cannot be :
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1 started, then verify I
' natural circulation ;
from trended values: '
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- 1) RCS subcooling I i
I based on core !
exit :
l thermocouples greater than .
}' 30*F.
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- 2) SG pressures are !
stable or !
l3 decreasing.
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PAGE 9 of 12 GENERAL REVISION i
E0 P- 1.1 i j REVISION 2 :
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ACTION . ALTERNATIVE ACTION l i
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.3) RCS Tg stable !
4 or decreasing.
saturation temperature for l j SG pressure.
- 5) RCS core exit thercoccuples !
stable or decreasing-
- c. Il natural !
circulation not verified, then l increase steam dumping. .
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- 14. Check Intermediate Range l Powerbelo5P-6
( 7 5 x 10 %) . l
- a. Transfer NIS RECORDER i (NR-45) to both Source Range channels. ,
- 15. Maintain stable plant '
conditions:
(
- a. PZR pressure at 2235 I psig (2220-2250 psig).
- b. PZR level at 25%. l
- c. SG Narrow Range levels between 38% and 50%.
- d. RCS Tavg at 557'F.
PAGE to of 12 GENERAL REVISION
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- QUESTION
$ 7.23 How are the trip bistables of a failed power range detector placed in the trip j condition?
j ANSWER KEY RESPONSE:
4 Removin drawers.g the applicable control and instrument power fuses on the power range !
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REFERENCES:
- VCS, AOP-401.10, POWER RANGE FAILURE, P 2 SUGGESTED RESPONSE
Either the above answer, or:
Pulling the control power fuse on the power range drawer.
- REASON
- The answer key is correct in that, procedurally, both fuses are aulled. .
Stowever, technically, the removal of the control power fuse a one is !
sufficient to place all bistables in the trip condition. See attached reference.
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.e. o QUESTION: 8.22 The reactor is being refueled (>23 ft above Rx Vessel Flange), loop "C"is isolated for maintenance and RCPs "A" and "B" are out of service for breaker repairs. "B" RHR pump and its Heat Exchanger are being used to circulate reactor coolant, with "A" RHR pump INOPERABLE for routine maintenance. A request to take the "B" EDG out of service for about 30 minutes to do a surveillance on the generator is made by the electrical supervisor. Can this surveillance be performed? Support your answer by referring to the applicable TS. ANSWER KEY RESPONSE: Maintenance cannot be performed (0.5), TS 3.9.7.1 applies (1,0).
REFERENCE:
VCS, TS 3.9.7 SUGGESTED CORRECT RESPONSE: Maintenance can be performed,TS 3.9.7.1 applies. REASON: The action statement for TS 3.9.7.1 discusses suspending all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the RCS. Provided that only new fuelis being added to the core the action statement needs not be entered (there is no increase in decay heat load being introduced). In addition the loop continues to be in operation, and taken in conjunction with the note which applies to this specification (*The RHR loop may be removed from operation for up to 1 hour....), an acceptable operator action would be to permit t' : surveillance of the diesel. The question, as stated, provides less information than is really required for an operator to make an informed, conclusive response. Because of the numerous potential technical specifications which could be applied, and the additional information required to determine the applicability of each of these specifications, a conclusive response to this specific question is extremely difficult to provide. Therefore, the overall merit and acceptability of this test question, as stated,is deoatable, o
0 < d'.. , 8.11. a e. ,
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REFUELING OPERATIONS 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION l l HIGH WATER LEVEL : LIMITING CONDITION FOR OPERATION { 3.9.7.1 At least one residual. heat removal (RHR) loop shall be OPERABLE and ' in operation." APPLICABILITY: MODE 6 when the water level above the top of the rea: tor pressure vessel flange is greater than or equal to 23 feet. . ACTION: With no residual heat removal loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS l
- 4.9.7.1 At least one residual heat removal loop shall be verified to be in j operation and circulating reactor coolant at a flow rate of greater than or
- equal to 2800 gpa at least once per 12 hours.
1 I 1 i a 1 The residual heat removal loop may be removed froe operation for up to 1 hour l J per 8-hour period during the performance of CORE ALTERATIONS in the vicinity I of the reactor pressure vessel hot legs. , i i l i f i l } SUMMER - UNIT 1 3/4 9-7 I
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c I ENCLOSURE 4 i SIMULATION FACit!TY FIDELITY REPORT , facility Licensee: South Carolina Electric and Gas Company Facility Licensee Docket No.: 50-395 l i . Facility Licensee No.: NPF-12 Operating Tests administered at. V. C. Sumner Nuclear Station l Operating Tests Given On: March 8-10, 1988 i During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies ! were observed : i (1) Attempts were made to recreate recent site specific LERs involving i failures in the 120 VAC Vital Instrumentation System. (LER87-027,
! Loss of Channel IV process rack with failure of "C" MFP to trip and l LER 87-015. Failure of A/E inverter). There was no capability to i simulate the loss of the individual process rack nor the failure of j a MFP to trip. Some of the effects of the loss of the process rack , could be opproximated (loss of spray, loss of steam dumps) but recreating the event was an impossibility. The loss of an inverter was capable of being recreated, however, the fidelity of this simulation
, was poor. For example, instrumentation associated with other chanti6ls j would also fail, confusing the operators trying to detennine what event j occurred. l (2) When perfonning rod operability testing (driving rods in/out several 2 steps), it was noted that the DRPI indications for each group of rods j would not accurately track actual rod motion, j I i ) i i
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