ML20148L143
| ML20148L143 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/01/1988 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML17347A705 | List: |
| References | |
| L-88-49, NUDOCS 8804010181 | |
| Download: ML20148L143 (82) | |
Text
f P. O. BO'X 14000, JUNO BE ACH, FL 3340t>-0420 ENCLOSURE 3 AAL m
O 1
FEBRUARY 1 1983_
L-88-49 Dr. J. Nelson Grace Regional Administrator, Region II
,()
U.
S.
Nuclear Regulatory Commicsion 101 Marietta St.,
N.W.,
Suite 2900 Atlanta, GA 30323
Dear Dr. Grace:
Re:
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Senior Reactor Operator Exam Comments Florida Power & Light Company has reviewed the Senior Reactor Operator Upgrade examination presented to Turkey Point operators on January 26, 1988.
As discussed in the exit
^3 meeting following the examination period, FPL has prepared (d
comments on questions in the examination for NRC review and consideration prior to grading the examinations.
The comments are provided in the attachment.
Should you or your staff have any questions on this information, please contact us.
Very truly yours,
/s rW
'3 C.
O cody Execu ive Vice President
/
COW /PLP/gp Attachment cc:
Document Control Desk, USNRC Mr. J. A.
Arildsen, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant PLP/001. SOL 8904010181 080314 PDR ADOCK 05000250 V
9 O
1 have reviewed the NRC Exam Question Review Package and concur with the responses provided.
O 1
2d-V Mw
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t T.N. Finn 0
- a. Aries U
Training Department Supt.-N PTN Regulation and Compliance o
l t W>
C.J. Bak r h
J.S. Odom Plant Manager-N Site Vice President O
1
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NRC EXAM QUESTION REVIEW O
QUESTION:
5.11 A critical boron calculation has been performed prior ter startup. State how the calculated value changes for cach of the following. Answer INCREAdE, DECREASE or REMAIN TIIE SAME.
b). The desired critical rod height is increased from 160 steps withdrawn to 180 steps withdrawn.
RESPONSE
We request that the answer be changed from decrease to increase for the following reason:
If actual critical rod height has been increased by 20 steps this adds more positive reactivity than originally calculated. Critical boron concentration would have to increase by an amount that would insert enough negative reactivity to offset the positive reactivity.
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- lti 5/mf 01/2&SS IMF I m
NRC EXAM QUESTION REVIEW O
QUF,STION:
5:15 For each of the following parameters, state whether a reactor power increase from 50% to 75% will cause the parameter to INCREASE, DECREASE or REMAIN THE SAME, Consider each case separately.
NOTE:
Assume that system reaches equilibrium after power change.
c). Shutdown Margin. (power change by rod withdrawal only)
RESPONSE
We request that the answer be changed from increase to no change for the following reason:
Pulling rods to increase power level will increase RCS temperature. As temperature increases, power defect adds an amount of negative reactivity equal to the positive reactivity added by the control rods. By having power defect alone offset the positive reactivity addition the shutdown margin will remain the same.
O l
- l11 -5/mE01/28/88 l'ge2
NRC EXAM QUESTION REVIEW i
'-~'
QUESTION:
5:17 The following are plant parameters of Reactors "A" and "B" before and after simultaneous reactor trips:
Reactor"A" Reactor "B
Before trips-Steady state power 100 %
50%
Boron concentration 1500 ppm 1500 ppm After trips-Reactivity from rod insertion
-8100 pcm
-8000 pcm Average temperature 547 deg 547 deg Maximum single IRW
-2000 pcm
-2000 pcm Assuming that no operator action takes place, identify the plant that has the larger shutdown margin { Denote as "REACTOR A" or "RFsACTOR B"} at each of the following times.
NOTES:
1)
Figures 5.1,5.2,5.3 and 5.4 are enclosed for reference.
2)
Figures 5.1,5.2,5.3 and 5.4 are applicable to BOTH REACTORS.
a)
One (1) minute after the trip
(,)
b)
Fifty (50) hours after the trip RESPONSFs:
Part a We request that you accept"Both Reactor A and Reactor B are the same" as an additional answer for the following reason:
Using the curves supplied the following data was determined-Reactor A Reactor B Rods
-6100
-6000 Power Defect
+ 1120
+ 600 Xenon
-2000
-2200 Total
-7580
-7600 These results are very close and some allowance should be granted for slight differences in reading the curves.
O
- l115/mf 01/28/88 l' age 3
.NRC EXAM QUESTION REVIEW O
Part c We request that the answer be changed to Reactor A for the following reason:
.Using the curves supplied the following data was determined-Reactor A Reactor B Rods
-6100
-6000 Power Defect
+ 1120
+ 600 Xenon
-900
-400 Samarium
-680
-605 Total
-6560
-6405 ll O
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O
- lti-5/mf;01/28/88 l' age 4
NRC EXAM QUESTION REVIEW QUESTION:
5.18 Answer each of the fellowing statements concerning the count rate (inverse multiplication) plot for rod withdrawal. TRUE or FALSE.
c)
A count rate, which is taken before the reactor power level reaches steady state (i.e.
count rate is taken shortly after reactivity is added), will result in a HIGHER predicted critical rod height than if a steady state count rate were taken.
d)
If the count rate taken prior to the last rod withdrawal did NOT reach steady state, the predicted critical rod height would be HIGHER than if that count rate taken had reached steady state.
RESPONSE
We request that answer D be changed to True for the following reason:
After reading statement C, which is True, and comparing it to statement D we feel that both statements are saying the same thing.
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- lti-5/mf,01/28/88 Page 5
NRC EXAM QUESTION REVIEW p
A QUESTION:
6.02 Select the statement below which most correctly describes the effect of overcompensating ONE (1) Intermediate Range Nuclear Instruments (IRNI).
- a. The indicated IRNI power level is LOWER than the actual power level, and the Source Range Nuclear Instruments (SNRI) will automatically energize above their reset setpoint during a reactor shutdown.
- b. The indicated IRNI power levelis LOWER than the actual power level, but the SRNI will NOT automatically energize above their reset setpoint during a reactor shutdown.
c, The indicated IRNI power level is HIGHER than actual power level, and the SRNI will automatically energize above their reset setpoint during a reactor shutdown.
- d. The indicated IRNI power level is HIGHER than the actual power level, but the SRNI will NOT autoinatically energize above their reset setpoint during a reactor shutdown.
'1
RESPONSE
%)
We request that the answer be changed to B for the following reason:
An overcompensated intermediate range instrument will indicate lower than the actual power level which makes answers C and D incorrect. During a reactor shutdown the overcompensated instrument will decrease faster (indicated) but the logic for auto reset of the source ranges requires that both intermediate instruments have decreased below the auto reset setpoint. The source range instruments will not auto reset until the properly compensated intermediate range instrument has reached the auto reset setpoint. This makes answer B the correct response.
REFERENCE; SD-4, Excore Nuclear Instrumentation pgs. 33,38, Figure 23 0
wRI-5/mf:01/28/88 Page1
('i The lithium and alpha particle resulting from this reaction cause ionization in the outer N2 gas volume.
The electrons produced by the ionization are collected on the outer can wall.
This produces a
signal which is proportional to the neutron flux.
Electrons are also collected on the outer can nall from the gamma radiation whicn interacts with the outer gas volume.
This additicnal signal is proDortiona' to the gamma flux and is additive to the neutron flux signal.
The outer cadmber ocerates in tre ionization region thus all the charged particles produced n the initial ionizing events are i
collected on the electrodes.
In the inner can volume, the gamma fivA also reacts with the N2 gas procucirg a signal proportional to the gamma raciation.
The inner cnamber is ocerateo in the recombination region to permit ad iustment of the output carrent by varying the compensating voltage.
If the inner volume compensation voltage is set properly the outer can signal of gammas plus neutrons interacts uitn-tre inner can gamma only signal and the gamma signals cancel out.
This neutron only signal is then amolified before it is displayed on tne meter or sent to the protection and control circuitry.
f(
Gamma Compensation It bec0mes necessary to define the term compensation and the effects of under compensation and over compensation to clearly understand the process of neutron detection in the intermediate range.
Compensation is a term applied to tne negative voltage signal applied to the inner volume of the CIC which cancels or comcensates for the current signal produced by the gamma radiation interacting with the outer volume of the detector.
-!efer to Figure 22.
This becomes very important to the operator because an incorrect setting of compensating voltage i.e.,
overcompensation or under-compensation would cause an erroneous neutron level indication on the meters.
Refer to Figure 22.
As noted on the curve undercompensation results in an erroneous high neutron level reading about 10 minutes af ter shutdown; overcompensation results in an erroneous low-neutron level about 12 minutes af ter shutdown.
Compare the two initial ranges and also relate intermediate amps to Source Range count rate (below P-10).
If improper compensation is suspected advise I & C pecmptly.
O 50004-Rev.2-33
- cas
=
REMOTE RECORDER This is tne sarie recorder unit discussed previously with regard to tne Source Range.
This recorder through selector switches also serves to record the selected IR and PR.
It is a two-pen recorder.
It wi'l normally record one IR at a time.
A 0-50 mv de signal f cm the isolation amplifier is supplied to the recorder and is proportional to the IR ion chamoe" current of 10-11 to 10- 3 aneres.
IR STARTUP-RATE CIRCUITY As described earlier for the SR, the startup-rate crawer rec 3ives inout signals (0-!0V OC) from each of tne SR and IR channels.
Refer to Figures 16
& 29.
Four rate amplifier mooules condition each of these signals and' transmit four respective rate signals to the respective control rocm startup-rate meters.
The remote IR startuo-rate indicators for IR cnannels N-35 and N-36 are located on the console.
A test module may De Jsed to inject a test signal into any one of the cate circuits and can be monitored on a test meter mounted on the front panel of the SUR drawer.
Two poaer supplies are installed in such a manner as to ensure rate indication f rom at least are Source and Intermediate Range channel pair uoon tne loss Of one power supoly.
P-6 Permissive The P-6 permissive is energized during reactor startup ahen 1 out of the 2 Intermediate Range channels reaches 10-10 amperes.
Refer to Figure 25.
This 4
is equivalent to a source range count level of approximately 4 x 10 cos.
Once the P-6 permissive is satisfied (indicated by a status light on VPB status light panel 8, windows 9-1 and 9-2) tne source range high level trip can be marually blocked.
By blocking the trip on :ne console, the high voltage to tne source range detector is automatically removed.
The provision is only operational below permissive P-10 which is supplied by the Power Range c.hannels.
Above P-10 the defeat circuit is automatically bypassed and source range nigh voltage cutoff is maintained.
i When shutting down, ' source range high level trips and detector high -voltageJ are ' automatically reactivated when both Intermediate Range channels ~ drop below 10-10-amperes (P-6, light de-energized). If an undercompensation of the IR detector is present preventing clearing of the P-6 permissive, Figure 17, e cas 50CC4 Oev.2-38
CIC GAMMA COMPENSATION 10 ELECTRIC ALLY ADJUSTED COMPENS ATED ION CHAMBER to,
10 I
TYPICAL SHUTOOWN CURVE es!
,0-7
=
a e
b
~
10
=
a2 I
10
=
juNDER COMPENS ATION h
-10 10 OVER COMPENSATED CORRECTLY COMPENS ATED
-11 NEUTRON LEVEL DECAY OUE TO
%g%
10 54 SEC. HALF LIFE 03 LAYED NEUTRON n
%g%
E MITTE R.
(80 EEC. NEGATIVE PERIOD) g
-12 1
1 1
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1 1
(
J10
(
0 2
4 8
8 10 12 14 18 18 20 22 24 28 28 30 32 T"JE AFTER REACTOR SHUTOOWN (M4NUTES) i 2
TRACE IF UNDER I
COMPENSATED INT. RANGE
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s%
APPROX.
l TRACEIF OVER to SEC. PER100 18 MIN.
CCMPENSATED INT. R ANGE 2 DECADE PRCMP JUMP L-t 1
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1 10*
10 10
10 go jo go $
10 '
10'
-10
-8
-7 4
ICH CURRENT METER READING TYPICAL RECORDING CHART TRACE (TRIP FRCM FULL POWER)
FIGURE 23 Au i
NRC EXAM QUESTION REVIEW O
QUESTION:
6:06 State three (3) requirements which must be satisfied prior to enabling RHR isolation valve, MOV-750, to open. (Include setpoints as necessary.)
RESPONSE
W e request that you accept additional answers for the following reason:
The question asked for requirements but the answer listed interlocks. For this reason please accept the following as additional answers-1.
Breaker for MOV-750 closed 2.
Power supply to MOV-750 available 3.
The valve shall not be opened if reactor coolant pressure exceeds 450 psig or reactor coolant temperature exceeds 350 F.
REFERENCE:
O SD-21, Emergency Core Cooling System, pg.16, Figure 9 3-GOP-305, Hot Standby to Cold Shutdown, pg.8 step 4.3.1,pg.26 step 5.10 3-OP-050, Residual Heat Removal System, pg.7 step 4.1, pg.9 steps 5.1.2.8 and 5.1.2.9 O
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- 111-5/mf.0 I/28/88 1%Re 2 I
RESIDUAL HEAT REMOVAL SYSTEM
%)
For the purposes of this discussion, the normal and emergency functions of the RHR system will be covered separately.
Normal System Functions As was previously discussed, the RHR system is normally used to remove the:
decay and sensible heat from the RCS during. plant startup, cooldown, and refueling when RCS' pressure is 450 psig and tempe.ture is 350*F.
The 350*F restriction is based on the RHR pump seal limitations.
The 450 psig limitation is based on not exceeding RHR system design pressure.
The 450 psig system limitation plus the RHR pump shutoff head of about 150 psig equals the system design pressure of 600 psig.
The design heat transfer rate for the RHR system is 58.8 x 106 Btu /hr based on a RHR heat exchanger inlet temperature (RCS) of 140*F and a CCW inlet temperature of 107.4*F.
The design heat transfer rate is based on the decay heat removal requirements 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> af ter shutdown from an infinite period of reactor operation. With a total CCW flow of 20,400 gpm to the RHR heat exchangers, the RCS can be cooled down from 350'F to 140'F during the period between 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> af ter reactor shutdown without the CCW leaving the CCW heat exchanger exceeding 125?F.
The RHR system is capable of l
removing more heat than design during a plant cooldown, since design heat transfer is based on the temperature difference of 140*F (reactor coolant)
I and 107'F (component coolant) as stated above.
A rapid cooling rate may result in a significant increase in temperature of component coolant leaving the CCW heat exchanger.
(Remember, from 50-8, that the maximum CCW supply temperature to the RCP's is 130*F.
During a plant cooldown this becomes a limiting parameter.)
The RHR system consists of two independent, redundant trains.
Each train consists of a
- pump, heat exchanger,
- piping, valves and attendant the heat exchangers are instrumentation.
See Fiqure 8.
Each p mp pnd located in separate compartments on the L-4'] elevation of the auxiliary l
l building.
Y,,
A S021-Rev.2-16
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RHR ISOL AT/ON VALVE.S 750 oR 75/
n L.)
INrfARuPT CLOSE cs s Norts cs acs to PAESSURE O PB To V'A OPEN CLOSE
>%SPSIG:
VPB S R. ro OMS 84,3A(e)
Sco2 Ale) mic S w/ TcH RHA RwST
- A
- N
Cto SEO CLOSED MODE I
a T
a,,
S L~5 N
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VAivt 750(751)
VAWE 750(73/)
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,h VALVE OPEN No rf S.
/. aim BfA 1/ 7 > $ASPSM Ako DuR/NG Cort CYCLE c r VALVE L' G1 UE J /T <525PS/G
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3 B orn A n 3ER nko St uf L IT-PERMISSIVE ro /r rt9Ru pr
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C: c Mc c ycz r. OPERR 7[ C S Yo Fux1y opfu VAL vf U wirH R C S > 930 PS/G Mo v 7f0 f/69
} 751 A.?E CL O SE D A wa G kR SDRt gry. i
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8rocedw e No pro (edu e
- tie pg, r
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3.GOP 305 Hot Standby To Cold Shutdown
^oo+ C m 10/28,86 g
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4.2 Reactor Coolan t Pum ps 4.2.1 The precautions and limits listed in 3 OP 041.1, Reactor Coolant Pump, shall be observed.
4.2.2 Operation of reactor coolant pumps should continue until NIode 5.
Cold Shutdown, is reached to provide pressurizer spray flow (B or C RCP), to prevent the temperature difference between loops from exceeding 25 F, and to ensure cooldown of steam generators.
4.3 Residual Heat Removal Loop 4.3.1 The Residual Heat Removal (RHR) Loop Isolation Valvesc MOV-3-750 and MOV-3-751 shall not be opened if reactor coolant pressu 9 exceeds 450 psig or reactor coolant temperature exceeds-350*F..
1.
The RHR Isolation Loop Valves, SIOV-3 750 and SIOV 3 751, will automatically start to isolate when RCS pressure is 515 to 535 psig. The isolation signal is indicated by a yellow light present on VPB, and alarm on the Reactor Coolant Annunciator
?anel. When the pressure drops below the isolation.setpoint, there will be a blue light present on VPB. The isolation valves travel may be reversed by de aressing the push buttons located
(')-
below the yellow and blue lig ats VPB. The yellow light should go out and the blue light will remain on until the isolation valves are full open.
4.4 Shutdown Rod Banig 4.4.1 Both shutdown banks of control rods should be at the fully withdrawn position when going from 51 ode 3, Hot Standby, to 51 ode 5, Cold Shutdown. However, shutdown rods may be left inserted at the discretion of the Plant Supervisor-Nuclear as long as shutdown margin is maintained. Refer to the Plant Curve Book for applicable baron concentration.
4.4.2 At any time when moving shutdown or control rod banks, closely monitor group step counters, RPIs and all nuclear instrumentation channels.
4.4.3 Control rod drive mechanism cooling fan operation shall continue until RCS temperature is below 350 F, and should continue as long as control rod drive mechanisms are energized.
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3.G O P.305 Hot Standby To Cold Shutdown 12/31/87 l
OU INIT erature is less than 350*F and pressurizer pressure is When RCS loop temE.sce RHR in operation for cooldown in accordance with 5.10 less tht.n 450 psig,l Heat Removal System.
p 3.OP-050, Residua 5.10.1 Refer to Enclosure 1, Valve Exercising Reference, for valves to be exercised. (Mark N/A if exercisingjs not required per Step 5.1.1) 5.10.2 LIf proceeding to Mode 6, Refueling. commence performance of OP-3206.2, Residual Heat Removal System. Refueling Interval.
(Mark N/A if not proceeding to Refueling)I C AUTION Technical Specifications require the Overpressure Mitsgating System (OMS) to be in operation when the RCS temperature is less than or equal to 275*F.
5.11 Prior to RCS cooldown to less than 276 F, when RCS temperature is in the range of 276*F to 285*F and pressurizer pressure is in the range of 325 to 375 psig, establish and verify OMS operation in accordance with 3.OP.041.4.
Overpressure Mitigating System.
5.11.1 Refer to Enclosure 1, Valve Exercising Reference, for valves to be exercised. (Mark N/A if exercising is not required per Step 5.1.18 U,,
NOTE I
i This step is not part of the OMS test. but is performed at this time to monimote the t
I l
I number of containment entries requoted.
s.._..._._._...._...._...-.._..._.a l
5,11.2 In the containment (penetration 34) unlock and close the Containment Service Air Header Drain Regulator Bypass valve.
3 40 208.
5.11.3 When the OMS system has been placed into operation. inform the Construction Supervisor that construction activities may resume in the following areas:
l Turbine Deck Main Steam Platform Mezzanine Deck Pipe and Valve Room Turbine Building Ground Level Containment Spray Pump l
EDG Room Room AFW Area Safety Injection Pump Room l
Blowdown Area RHR Pump and Heat Exchange Feedwater Platform Room 4160V Switchgear Rooms BAST Room 480V Load Center Rooms Charging Pump Room 480V MCC'S Condensate Polishers DC Switchgear and Inverter Cable Spreading Room Room Auxiliary Building North and
,l p)
South Hallway L.
., m sic e, e
Procedw e No :
Procedure htle:
Page:
r 7
Approval oate 3 OP 050 Residual Heat Removal System 10/1/87 O
4.0 P REC A UTIONS/ LIMITATIONS 4.1 The RHR Loop Isolation valves shall not be opened if RCS pressur6 is greater than 450 psig or RCS temperature is greater than 350*F.
4.2 When the RHR system is not being used for cooldown, it shall be aligned for low head safety injection.
4.3 To prevent overheating and pump damage, do not operate the RHR pumps "dead headed".
4.4 Each RHR Loop required to be operable during modes 3,4,5 and 6 shall be supported by one CCW Heat Exchanger, one CCW Pump and one ICW pump powered from the same electrical source as the associated RHR pump. At least one of the required RHR Loops and its related support components shall be capable of being powered ' om an operable Emergency Diesel Generator. Only one ICW Header and CCW Basket Strainer are required to be operable to support both RHR Loops.
4.5 When rotating RHR pumps and RCS level is at mid-nozzle, stop the running RHR pump prior to starting the alternate RHR pump to prevent possible RHR pump cavitation due to a decrease in level.
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ProCTCWre NO ProCeow'e fetle Dagg.
9 ADDroval Date.
3 OP.050 Residual Heat Removal System 11/1'2/87 (y
s -
I N* l T I A L S CK'D VERIF 5.1.2 (Cont'd) 4.
Verify the following valves indicate closed on the SIS Recirculation Status - SI Lights panel:
a.
SI Pump Recire Phase Suct Stop, AIOV-3 863A b.
SI Pump Recire Phase Suct Stop,510V-3-863B 5.
Open the RHR To CVCS Stop Valve,3 205B.
6.
Slowly open the RHR Letdown to CVCS valve, HCV 3-142. to equahze RHR and letdown pressure.
7._._._._._._._._._._._.g.,._._._._._._._._._._.)
I I
The low Pressure letdown Control valve, PCV-3-145 shall be maintained in manual I
I until a RHR pump has been started.
,t._._._._._._._._._._._._._._._._._._._._._._._.J 7.
Adjust the Low Pressure Letdown Control valve, PCV-3-145 to match the RCS pressure.
8.
Unlock and close the following breakers to energize the RHR O
Loop 3C Suction Stop valves.
a.
30615(MOV-3-750) b.
30731(MOV 3-751) 9.
Open the following valves:
a.
RHR Loop 3C Suction Stop,310V-3 750 b.
RHR Loop 3C Suction Stop,310V-3-751
- 10. Unlock and open the HCV-3 758 Air Supply Isolation valve.
3 40-1019.
- 11. lOpen the following valves:
RHR Hx B By-Pass Hdr Isolation,3-757C a.
b.
RHR Hx A By Pass HdrIsolation,3-757D
. Il ;C es et er
NRC EXAM QUESTION REVIEW QUESTION:
6:11 State the four conditions / switch positions which must exist in order to place the Overpressure Mitigation System (OMS) in operation in the low pressure mode.
RESPONSE
We request that you accept the following additional answers:
1.
Nitrogen (backup air supply) charged to 2070 psig or higher 4
2.
RCS temperature must be in the range of 276 F to 285 F and pressurizer pressure must be in the range of 325 to 375 psig.
REFERENCE:
AP-0103.32, Reactor Cold Shutdown Conditions, pg.3 step 5.1 3-GOP-305, Hot Standby to Cold Sh utdown, pg.26 step 5.11 0
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11/21/87 ADMINISTRATIVE PROCEDURE 0103.32, PAGE 3 REACTOR COLD SHUT 00WN CONDITIONS 4.11.1 At least one Reactor Coolant Loop is operable; or 4.11.2 The RCS temperature is less than 140*F, boron concentration resulting in a shutdown greater than or equal to 10 percent aK/K margin and the refueling cavity is flooded to greater than 23 feet above the reactor vessel fl ange with the reactor vessel head iemoved.
4.12 The Pressurizer Safety Valves shall be installed and operable in accordance with 3/4-SMM-041.1, Pressurizer Safety Valve, Setpoint Testing prior to installing the reactor vessel head.
4.13 When using the charging pumps with a flowpath from the BASTS through the RCP seals, fl ow shall be gree.er that 15 gpm but less than 45 gpm to prevent RCP seal damage, 5.0 Related System Status:
j 5.1 When in the Cold Shutdown condition, the OMS shall be operable at the low setpoint range, or there shall be an opening of the RCS with an area of at least 2.2 square inches.
For an Overpressure Mitigating Channel to be operable and meet Technical Specification 3.15.3, the following conditions for the primary and backup channels must be met for each channel:
O oss ar' rv ^i'a"ed:
1.
Status light ON, i.e., OMS Primary Aligned 2.
M0V *-535 - OPEN 3.
PCV *-456 - Switch in AUTO position 4
Nitrogen (Backup Air Supply) Charged to 2070 psig or higher 5.
OMS Low Pressure Setpoint Position selected for *-456 OMS lackup Aligned:
1.
Status light ON, i.e., OMS Backup Aligned 2.
MOV *-536 - OPEN 3.
PCV *-455C - Switch in AUTO position 4
Nitrogen (Backup Air Supply) Charged to 2070 psig or higher 5.
OMS Low Pressure Setpoint Position selected for *-455C 5.2 After reaching Cold Shutdown, one of the two operating CCW pumps should be placed in Standby by perfoming Table III.
6.0 Reference s/Comi tment Document s:
6.1 References 6.1.1 Technical Specification 3.2a O
3/4-G0P-103, Power Operation to Hot Standby 6.1.2 6.1.3 3/4-GOP-305, Hot Standby to Cold Shutdown 6.1.4 OP-0209.1, Valve Exercising Procedure 6.1.5 3/4-0P-041.7 Draining the Reactor Coolant System
Pr0CeCu e "uo.
P'O<tdwr0 Titig r
- agg, 26 3:GOP 305 Hot Standby To Cold Shutdown n o w o ate 12/4/86 INIT.
O"#b' When RCS loop temperature is less than 350*F and pressurizer pressure is 5.10 less than 450 psig, place RHR in operation for cooldown in accordance with 3 OP 050, Residual Heat Removal Systern.
5.10.1 Refer to Enclosure 1, Valve Exercising Reference. for valves to be exercised. ' Mark N/A if exercising is not required per Step 5.1.1) 5.10.2 If proceeding to Sfode 6, Refueling, perform OP-3206.2, Residual Heat Removal System - Refueling Interval. (Mark N/A if not proceeding to Refueling)
C AUTIO N Technical Specifications require the Overpressure Motigating System (OMS) to be in operation when the RCS temperature is less than or equal to 2757 5.11 l Prior to RCS cooldown to less than 276 F, when I CS temperature is in the range of 276*F to 285'F and pressurizer pressu re is in the range of 325 to 375 psig, establish and verify OMS operation in accordance with 3 OP-041.4, Overpressure Mitigating System.l 5.11.1 Refer to Enclosure 1, Valve Exercising Reference, for valves to be exercised. (Mark N/A if exercising is not required per Step 5.1.1)
(
l NOTE I
I This step is not part of the OMS test, but is performed at this time to minimize the I
number of containment entries required I
t._.
-.J 5.11.2 In the containment (penetration 34) unlock and close the Containment Service Air Header Drain Regulator Bypass valve.
3 40-208.
1 5.11.3 When the OMS system has been placed into operation, inform the Construction Supervisor that construction activities may resume in the following areas:
Turbine Deck Main Steam Platform Mezzanine Deck Pipe and Valve Room i
Turbine Building Ground Level Containment Spray Pump i
EDG Room Room i
AFW Area Safety Injection Pump Room l
Blowdown Area RHR Pump and Heat Exchange 1
Feedwater Platform Room 4160V Sw.tchgear Rooms B AST Room 480V Load Center Rooms Charging Pum a Room 480V MCC'S Condensate Po ishers DC Switchgear and Inverter Cable Spreading Room i(
)
Room A uxiliary Building North and South Hallway erABmatem
NRC EXAM QUESTION REVIEW O
QUESTION:
6.12 State five automatic functions which are performed upon receipt of a Emergency Diesel Generator Lockout Relay Signal.
RESPONSE
We request that you accept additional answers because this is an open-ended question.
Some additional acceptable answers should be:
1.
Flashing blue lockout relay light on vertical panel A.
2.
Annunciator Targets come in a) F 8/1 b) F 8/5
REFERENCE:
5610 T-L1 sht. 9C, Diesel Generator A Lockout Relay and Breaker Logic 4
5610-T-L1 sht. 9C1, Diesel Generator B Lockout Relay and Breaker Logic i
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NRC EXAM QUESTION REVIEW QUESTION:
6.18 State the two SAFETY design bases for a maximum water inventory in the S/Gs.
RESPONSE
We requast that you accept additional variations to the answer "minimizes the effects of a RCS cooldown in the event of a major Steam Line Break". Please accept a Steam Line Break at EOL.
REFERENCE:
FSAR, Section 14, pgs.14.2.5-1 thru 14.2.5-8 O
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D 14.2.5 RUPTURE OF A STEAM PIPE A rupture of a steam pipe is assumed to include any accident which results in an f^/
uncont rolled steam release f rom a steam generctor.
The release can occur due to y
a break in pipe line or due to a valve mal function.
The steam release results in an initial increase in steam flow which decreases during the accident as the steam pressure falls.
The energy removal from the Reactor Coolant System causes a reduction of coolant temperature and pressure.
With a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin..
If the most reactive control rod is assumed stuck in its tui!y possibility that the core will become critical withdrawn position, there is a and return to power even with the remaining control rods inserted.
A retura to a: the power following a steam pipe rupture is a potential problem only because high hot channel f actors which may exist when the most reactive rod is annused stuck in its fully withdrawn position.
Assuming the most pessimistic combination of circumstances which could lead to power generation following a steam line break, the core is ultimately shut down by the boric acid in tne refueling water stcrage tank.
a steam pipe
(']
The following systems provide the necessary protection against rupture:
1)
Saf ety injection System Act uation f roc: any of the following*:
Two out of three pressurizer low pressure signals.
a.
b.
Two out of three dif f erential pressure signals beteeen any steam line and the main steam header.
c.
High steam flow in two out of three lines (one out of two per line) in coincidence with either low reactor coolant system I
average temperature (two out of three) or low steam line pressure (two out of three).
I
- The details of the logic used to actuate Safety injection are discussed in Section 7.
14.2.5-1 gev 4 7/86 i
i
Two out of three high containment pressure signals.
d.
Reactor trip occurring upon actuation of the Safety injection System.
tv 2)
Sustained high isolation of the main feedwater lines.
3)
Redundant thus in addition to the feedwater flow would cause additional cooldown, normal control action which will close the main feedwater valves, any safety injection signal will rapidly close all feedwater control valves, trip the main feedwater pumps, and close the feedwater pump discharge r
valves.
Trip of the f ast acting steam line stop valves (designed to close in less 4) of three lines than 5 seconds with no flow) on high steam flow in two out of two per line) in coincidence with either low reactor coolant (one out system average temperature (two out of three) or low steam line pressure (two out of three).
opening of a steam generator relief or Three cases are presented:
Inadvertent O safety valve and complete (double ended) severance of a steam pipe with and
\\_ /
without offsite power available.
Inadvertent Opening of a Steam Generator Relief or Safety Valve Identification of Causes and Accident Description severe core conditions resulting from an accidental depressurization The most of the main steam system result from an inadvertent opening of a single steam dump, relief, or safety valve.
results in an initial The steam release as a consequence of this accident as the steam increase in steam flow which decreases during the accident The energy removal from the reactor coolant system causes a pressure falls.
In the presence of a negative reduction in coolant temperature and pressure.
moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.
)
Rev 4 7/86 14.2.5-2
Th2 cnolysis is parforsed to demns tr a te thct tha following criterien is a stuck rod cluster control assembly, with offsite power satisfied:
assuming single failure in the engineered safety features, available, and assuming a
g(j there will be no consequential damage to the core or reactor coolant system ofter reactor trip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.
Analysis of Ef fects t nd Consequences A.
Method of Analysis The following analyses of a secondary system steam release are performed for this section:
1.
A full plant digital computer simulation using the LOFTRAN code (Reference 1) to determine reactor coolant system temperature and pressure, during cooldown, and the ef fect of safety injection.
2.
Analyses to determine that there is no damage to the core or reactor coolant system.
The following conditions are assumed to exist at the time of a secondary steam system release:
1.
E nd-of-li f e shutdown margin at no-load, equilibriun xenon conditions, and with the most reactive rod cluster control assembly stuck in its fully withdrawn position.
Operation of rod cluster control assembly banks during core burnup is restricted in such a way that addition of positive reactivity in a secondary system steam release accident will lead to a more adverse condition than the case analyzed.
I not l
,0 V
14.2. 5 - 3 Rey, 1-11/83 l
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2.
A negctive modarotor co2fficient correspeeding to the erd-of-lifo redded core with the most reactive rod cluster control assembly in the
(~ : fully withdrawn position.
The variatiot of the coefficient with is included in the LOFTRAN calculations, The temperature and pressure K,ff versus reeperatu-e at 1000 poi corresponding te the negative moderator temprature coef ficient used is shewn in Figure 14.2.5-1.
2.
Minimum capability for injection of concentrated boric acid solution corresponding to the most restrictive single failure in the safety injection system (Failure of one safeguards train).
This corresponds to the flow delivered by two safety injection pumps delivering their full contents to the cold leg header.
Low concentration boric acid must be swept from the safety injection lines downstream of the refueling water storage tank prior to the delivery of concentrated boric acid (2000 ppa) to the reactor coolant loops.
This effeet has been allowed for in the analysis.
The boron injection tank boron concentration is assumed to be O ppa.
(%.
The case studied is a steam flow of 280 lb/see at 1100 psia from one steam generator with offsite power available.
This is the maximum capacity of any single steam dump, relief, or safety valve.
Initial hot standby conditions with minimum required shutdown margin at the no-load T avg is assumed since this represents the most conse rvative initial condition.
5.
Should the reactor be just critical or operating at power at the time of steam release, the reactor will be tripped by the normal overpower Following a trip at protection when power level reaches a trip point.
power, the reactor coolant system contains more stored energy than at no load and no load, the average coolant temperature is higher than at the fuel.
Thus, the additional is appreciable energy stored in f
there l
stored energy is removed via the cooldown caused by the steam release before the no load condi' ions of reactor coolant system temperature and shutdown margin assumed in the analysis are reached.
After the the cooldown and reactivity additional stored energy has been removed, C
i Rev 4 7/86 14 2.5-4
insortiens proccod in ths scoe mannst cs in the analysis which casumes no load condition at time zero.
However, since the initial steam generator water inventory is greatest at no load, the magnitude and gV duration of the reactor coolant system cooldown are less for steam line release occurring at power.
6.
In computing the steam flow, the Moody Curve (Reference 3) for FL/D = 0 is used.
Perfect moisture separation in the steam generator is assumed.
7.
B.
Results Figure 14.2.5-2 and 14.2.5-3 show the transient results for a steam flow of l
280 lb/see at 1100 psia from one steam generator.
The assumed steam release is typical of the capacity of any single steam dump, relief, or safety valve.
Safety injection is initiai.ed automatically by low pressurizer pressure.
Miaimum safety injection capability is that corresponding to two out of four safety injection pumps in operation.
Safety injection flow used in the analysis is shown on Figure 14.2.5-11.
Boron solution at 2000 ppm enters the reactor coolant system providing sufficient negative reactivity to prevent core damage.
The calculated transient is quite conservative with respect to cooldown, since no credit than that of the is taken for the energy stored in the system metal other fuel elements or the energy stored in the other steam generators.
Since period of about 5 minutes, the neglected the core transient occurs over a stored energy will have a significant effect in slowing the cooldown.
I Following blowdown of the f alted steam generator, the plant can be brought to a l
feedwater flow and stabilized hot standby condition through control of auxiliary lO 14.2.5-5 Rev 4 7/86 1
fety icj ction flev ce dsscribsd by plcat cperating preceduras.
The cparating ccedures would call for operator action to limit reactor coolant systca ge.,e and pressurizer level by terminating safety injection flow and to entrol steam generator level and reactor coolant system coolant temperature to sing tha auxiliary feedwater system.
Any action required of the operator intain the plant in a stabilized condition will be in a time frame in excess f ten cinutes following safety injection actuation.
tise saquence of events is given in Table 14.2.5-1.
Ccnclusions ishe enclysis shows that the criteria stated earlier in this section are accidental depressurization of the main steam system, the sccisfied.
For an lcinitun DNBR remains well above the limiting value and no system design limits
'cre excoeded.
Steen System Piping Failure Identification of Causes and Accident Description in an stoaa release arising from a rupture of a main steamline would result i
accident as the steam initial increase in steam flow which decreases during the I
i The energy removal from the reactor coolant system causes a prossure decreases.
reduction of coolant temperature and pressure.
In the presence of a negative l
moderator temperature coe fficient, the cooldown results in an insertion of t
If the most reactive rod cluster control assembly (RCCA) positive reactivity.
I there is an fully withdrawn position after reactor trip, is assumed stuck in its increased possibility that the core will become critical and return to power.
Tha core is ultimately shut de by the boric acid delivered by the safety injection system.
O Rev 4 7/86 14.2.5-6
Results (3
%)
The results presented are a conservative indication of the events which would occur assuming a steam line rupture.* The worst case assumes th'at all of the following occur simultaneously.
1)
Minimum shutdown reactivity margin of 1.77%.
The most negative moderator temperature coefficient for the rodded 2) core at end of life.
The rod having the most reactivity stuck in its fully withdrawn 3) position.
restrictive f ailure of engineered safety features, 4)
The most i.e. only two safety injection pumps availabic and one motor operated valve available to deliver fluid through three cold leg lines.
Core Power and R2 actor Coolant System Transient Figure 14.2.5-3 shows the Reactor Coolant System transient and core heat flux f ollowing a steam pipe rupture (complete severence of a pipe) outside initial no load the containment, downstream of the flow measuring nozzle at The break assumed is the largest break which can occur anywhere conditions.
outside the containment either upstream or downstream of the isolation valves.
full reactor coolant flow exists.
Outside power is assumed available such that The transient shown assumes the rods inserted at time 0 (with one r i
in its full.y Vlthdrhn positicn) and steam release from only one steam generator.
l Should the core be critical at near zero power when the rupture occurs the initiation of safety injection by high dif ferential pressure between any steam generator and the main steam header or by high steam flow signals in l
coincidence with either low reactor coolant system temperature or low steam Steam release from at least two steam line pressure will trip the reactor.
I generators will be prevented by either the check valves or by automatic trip of the f ast acting stop valves in the steam lines by the high steam flow (V
signals in coincidence with either low reactor coolant system temperature
,D
\\
Even with the f ailure of one valve,
or low steam line pressure.
for current Results have been re-evaluated to assure that criteria are met shown in Appendix 14D.
cycle kinetics parameters, as 14.2.5-7
release is limited to no more than 5 seconds for two steam generators
)
while the third generator blows down.
(The steam line stop valves are 3
designed to be fully closed in less than 5 seconds with no flow through With the high flow existing during a steam line rupture, the them.
valves will close considerably f aster).
As shown in Figure 14.2.5-3, the core becomes critical with the rods inserted (with the design shutdown assuming one stuck rod) at 25 seconds.
Boron solution at 20,000 ppm enters the Reactor Coolant System from the Safety Injection System a t 66 seconds with a delay of 20.5 seconds required to clear the Safety Injection System lines of low concentration
- loops, boric acid, between the boron injection tank and the coolant after the pressure has fallen to 1350 psia.
The computer calculation used assumes the boric acid is mixed with and diluted by the water flowing in the reactor coolant system prior to entering The concentration after mixing depends upon the relative the reactor core.
n flow rates in the Reactor Coolant System and in the Safety Injection System. The variation of mass flow rate in the Reactor Coolant System due to water density changes is included in the calculation as is the variation of flow rate in the Safety Injection System due to changes The Safety Injection System in the Reactor Coolant System pressure.
flow calculation includes the line losses in the system as well as the pump head curve.
No credit has been taken for the 2,000 ppm boron which enters the Reactor System prior to the 20,000 ppm boric acid.
The peak core average Coolant 2200 MWt.
flux for this case is 15.4* of the value at heat O
14.2.5-8
i NRC EXAM QUESTION REVIEW (d
)
QUESTION:7.04 For each of the following, provide the Turkey Point guideline for whole body exposure limits in millirem.
c) Yearly limit:
RESPONSE
We request that the answer (s) to this question should read "4,500 mrem / year" or "should not exceed 5,000 mrem / year"
REFERENCE:
0-AD51-600, Health Physics Alanual, pg. 30, Bottom note and step 5.18.1.1.b.2)
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0 ADM 600 Hecith Physics Manual 6i28,85 5.18 Personnel Exposure Control 5.18.1 Radiation Protection Standards (Cont'd) 1.
b.
Florida Power & Light Company Guideline Values 1)
Each individual entering the Radiation Controlled Area shall be limited to 250 mrem per quarter unless one of the following actions have been completed:
a)
Individual has signed a statement to the effect that he has no occupational exposure for current quarter. He shall then be allowed to receive up to 800 mrem / quarter in accordance with Section 5.18.1.1.c.
7._._._._._._._._._._._.g.,._._._._._._._._._._.3 I
I Only in an unusual situation (e g., the onl/ person available for a specialty craft I
I function that well require greater than 250 mrem exposure) should a written l
statement of exposure, estor.tated by the ondividual, be accepted. In such cases. the l
unusual situatsoa shall be noted and sogned by the Health Physics Department
=
I I
Supervisor.
- ' - ' - ~ ' - ' - ' - ' - ' - - ' - ' - ' - ' - ' - ' - ' - '
l O' b)
Individual has documentation from former employers that gives an estimate of his total occupational exposure for the current quarter.
He shall then be allowed to receive a dose that will bring his total up to 800 mrem / quarter in accordance with the Section 5.18.1.1.c.
l c)
Individual has signed a statement to the effect that he has never received any occupational radiation exposure (completed Form NRC-4).
He shall then be allowed to receive up to 2150 mrem / quarter in accordance with Section 5.18.1.1.c.
._._._._._._._._._._._.g.,._._._._._._._._._._.
I I
Indivodual dose should not exceed 5.000 miemlyear from all sources of occupatronal l
I I
exposure without prior authorization of the Health Physncs Department Supervssor t..' "d .' " '."_' ?' e ? '!."?'E. _. _. _. _. _. _. _. _. _. _. _. _. _. _. _. 1 mrems mrems mrems wk.
a tr.
year 2)
Whole body: head and See
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trunk; active blood-5.18.1.1C forming organs; lent Items 2 & 3 of eyes or gonads 300 4.500 2 on.c
NRC EXAM QUESTION REVIEW O
QUESTION:7.05 I
While approaching criticality using the 1/M plot, a projected critical rod position must be calculated after the third doubling (i.e.,1/M is approximately.125). State the three times that the result of this critical rod position projection requires that control banks be reinserted.
RESPONSE
Please expand the answer key to accept such aiiswer as:
"The reactor shall not be made critical with a difference of greater than or equal to 1000 pcm between the projected critical height and the ECC rod position".
REFERENCE:
3-GOP-301, Hot Standby to Power Operation, pg. 22 step 4.28.5 O
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99 AD "'Csa Oett 3.GOP 301 Hot Standby to Power Operation i1/5/87 O
4.23 The operability of the 51ain Feedwater Control Valves, FCV 3 475. 488. 498 should be checked by stroking through one complete cycle prior to placing the Turbine / Gen on line.
t I
4.24 If at anytime, a Limiting Condition for Operation cannot be met.10 CFR 50.36 reqitires that the reactor be shutdown or any remedial action permitted by the Technical Specification be followed: AP-0103.8, Reactor Shutdown Rate Time Limits, provides guidance for the resultant reactor shutdown, 4.25 Serious damage to the Alain Generator windings can result from operation of the generator outside of the terminal voltage limits of 20,900 to 23,100 volts.
4.26 Safety Injection Signals shall not be in a blocked status for any reason other than for depressurization and cooldown of the Reactor Coolant System.
4.27 During a Post Trip Recovery at EOL, when startup is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of criticality, contact the Reactor Engineering Department for startup guidelines.
4.28 The following guidelines shall be employed while approaching criticality using the 1/m plot.
4.28.1 If after the third (3rd) doubling (i.e.,1/m approximately.125), the I
projected critical rod position is below the insertion limit (107 steps on Bank C), reinsert the control banks and borate the RCS as o
necessary.
4.28.2 If after the (3rd) doubling (i.e.,1/m approximately.125), it is projected that the reactor cannot be made critical at the current acron concentration, reinsert the control banks and dilute the RCS as necessary.
4.28.3 If after the (3rd) doubling (i.e.,11m approximately.125), the projected critical rod position deviates from the ECC rod position by more than 300 pcm but less than or equal to 400 pcm: permission to pull the Reactor critical shall be obtained from the Plant Supervisor Nuclear or designee.
4.28.4 If after the third (3rd) doubling (i.e.1/m approximately.125), the projected critical rod position deviates from the ECC rod position by greater than 400 pcm but less than or equal to 500 pcm. permission to pull to criticality shall be obtained from the Reactor Supervisor or his ~
l designee.
4.28.5 If after the third (3rd) doubling (i.e.,1/m approximately.125), the projected critical rod position deviates from the ECC rod position by greater than 500 pcm, the control banks shall be reinserted and the ECC re evaluated. If the error cannot be determined, permission of the Operations Superintendent and Reactor Supervisor (or their O
designee >eaeiiseo8teined grierto =>ximet8ere><tercriticeie= der the guidance of tne 1.m plot. The Reactor shall not be made critical with a difference of greater than or equal to 1000 pcm between the projected critical height and the ECC rod position.
- ; 7.y a w l
NRC EXAM QUESTION REVIEW O
QUESTION:7.06 If the Unit 3 auxiliary feedwater system fails completely but offsite power is still available, the Steam Generator (S/G) feed pumps can be used to supply feedwater to the S/Gs utilizing the feedwater bypass regulating valves. In order of priority, state the four other methods to supply feedwater to the S/Gs under these conditions.
RESPONSE
This question does not specifically address ONOP-7308.1, Malfunction of the AFW system.
For this reason we request that"In order of priority..." be removed from the question. If the question was answered with 3-EOP FR-H.1, Response to Loss of Secondary Heat Sink in mind, the order is not the same and in fact, at one point, the order is not as important as getting any source of flow to the steam generators established.
REFERENCE:
O 3-EOP-FR-H.1, Response to loss of secondary Heat sink, pgs. 410 i
i I
O
- lt!-5/jt 01/2S/SS l' age 4
.m
......s.
..s 4
RESPON.SE TO LOSS OF
. s.. :...
3 EOP FR H.1 SECONDARY HEAT SINK 1/7/87 STEP ACTION EXPECTED RESPONSE-RESPONSE NOT OBTAINED.
(]
C A U TIO N If total feed flow is Ie'ss than 130 GPM per SIG due to Procedural e
Guidance, this procedure should not be performed.
If wide range level.in any 2 SIGs is less than 8% [32%] OR PRZ e
pressure is greater than or equal to 2335 PSIG due to Toss of secondary heat sink, RCPs should be tripped and Steps 9 through 15 should be immedia tely initiated for bleed and feed.
Feed flow should not be reestablished to any faulted SIG if a e
non faulted SIG is available.
1 Check if Secondary seat Sink is aequired:
a.
RCS pressure GREATER THAN ANY a.
Go to E 1, LOSS OF REACTOR OR NON FAL}LTED S/G PRESSURE SECONDARY COOLANT, Step 1.
b.
RCS temperature GREATER THAN b.
Try to place RHR System in service while 324'F [324'F]
continuing in this procedure. Refer to OP 050, RESIDUAL HEAT REMOVAL SYSTEM. IF adequate cooling with RHR System estabhshed, THEN return to procedure and step in effect.
p 2
Try To Establish AFW Flow To At least One S!G:
a.
Check conUol room indications for cause of AFW failure:
e CST LEVEL e AFW STEAM SUPPLY MOV POWER SUPPLY i
e AFW VALVE AllGNMENT b Check total flow to S/G GREATER b Perform tr'e following-THAN 130 GPM cer SiG
- 1) Reset SilF necessary
- 2) Reset FW isolation
- 3) OPEN the Unit 3 stancby S/G ;W pump man'ual isolation valve DWDS-012
- 4) START standby steam generator feedwater pump. A
- 5) OPEN feedwater bypass valves to desired flow.
If Feedwater cannot be established to at least one S/G THEN consult TSC staff for possible use of Unit 2 or 4 feedwater while continuing with step 3 c.
Return to procedure and step in effect 3
Stop All RCPs
. C Av et a
,. s... v 5
RESPONSE TO LOSS OF
....e...
3 EOP FR H.1 SECONDARY HEAT SINK 1/7/87.
.5TEP ACTiGNiEXPECTED RESPONSE RESPONSE NOT OBTAINE D C A U TIO N
~
.O If offsite power is lost after SI reset, manual action may be required to restart safeguards equipment.
4 Try To Establish Main FW Flow To At least One $/G:
a.
Ch4K'k condensate system - IN SERVICE
,a.
Try to place condensate system in service.
'lLNO_T, THEN go to Step 8.
b Open feedwater bypass valves to b.
Pe iorm the following:
desired flow
- 1) Reset 51if necessary.
- 2) Reset FW isolation.
- 3) Open fredwater bypass valves
- 4) Start S/G feed pumps E no feedwater bypass valve can be opened, THEN go to 5tep 8 c.
Establish main FW flow E feedwater flow cannot be' established O
THEN go to step 6 5-check s/c Leveis:
E feed flow to at least one S/G verified, a.
Narrow range level in at least one S/G -
a.
GREATER THAN 6% [32%)
THEN maintain flow to restore narrow range level to greater than 6% [32%}. E NOT verified, THEN go to Step 6 b.
Return to procedure and step in effect e
3 w
a O
.e 6
RESPONSE TO LOSS OF f
3 EOP FR H.1 SECOND ARY HEAT SINK 1/7/87 1
l l
STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED C A U TIO N 1
Followinq block of automatic SIactuation, manualSIactuation may be required of conditions degrade.
6 Try To Establish Feed Flow From Condensate 5ystem:
a.
Depressurize RCS to less than 1950 PSIG:
- 1) Checkletdown IN SERVICE
- 1) Use one PRZ PORV. IF NOT, THEN use auxiliary spray. Go to Step 6b
- 2) Use auxiliary spray
- 2) Use one PRZ PORV.
b.
Block 51 signals:
e LOW PRESSURE 51 e HIGH 5/G FLOW COINCIDENT WITH LOW 5/G PRESSURE OR LOW TAVG
/3 51 O
c.
Depressurize at least one 5/G to less than 430 PSIG:
- 1) Dump steam to condenser at
- 1) Vanually or locally dump steam from' maximum rate 5/Gs using steam dump to atmosphere valves d.
Establish condensate flow:
d.
Go to Step 8.
- 1) Start all available condeasate pumps f
- 2) Verify flow 7
Check s/G tevels:
E 'eed flow to at least one 5'iG venfied, Narrow rarge level in at least ore 5/G -
a.
a.
GREATER THAN 6% [32%}
' THEN maintain flow to restore narrow range level to greater than 6% (37%] @
NOT vertfied, THEN go to Step 8 l
b Return to procedure and step in effect 3
O
. cr.v c' t
- m
...v
->.,n RESPONSE TO LOSS OF 3 EOP FR H.1 SECOND A RY HE AT SINK.
1/7/87 STEP l ACTIONIEXPECTED RESPONSE l
RESPONSE NOT OBTAINED l
8 Check For Loss Of Se:Andary Heat Sink:
'a Parimeter Wide range S/G level in a.
Return to Step 1.
any 2 S/Gsless than 8% [32%]
OR PZR pressure greater than or equal to 2335 EXCEEDED C A U TIO N Steps 9 through 15 must be performed quickly in order to establish RCS hea t removal by RCS bleed and feed.
9 Actuate SI 10 Verify RCS Feed Path:
Manually start pumps and align valves as becessary to establish feed path. 'IF a feed a Check high. head $1 pumps -
path can NOT be established, THEN continue AT LEAST ONE RUNNING attempts to establish feed flow Retdtn to Step 2.
b.
Check valve align [nent for operating O
PROPER EMERGENCY pumps AllGNM ENT (Amber Lights on VPB) 11 Reset SI:
a.
Verify Unit 4 51 pumps NOT REQUIRED a.
Perform the following:
- 1) OPEN the following Si pump manual suction intedie valves:
al 870A b) 8708
- 2) OPEN the following 51 pump manual recirculation tie valves:
a) 892A b) 8928
- 3) Unlock and rack in the following Unit 4 RWST outlet isolation valve breakers:
a) 40712 b) ' 40605
.cavote3
o.
g RESPONSE TO LOSS OF 3 EOP.FR H.1 SECOND ARY HEAT SINK 1/7/87 p
STEP ACTIONIEXPECTED RESPONSE l
RESPONSE NOT OBTAINED l
U 11 '
Reset SI: (Cont'd)
- 4) CLOSE the following Unit 4 RWS7 outlet isolation valves:
- 5) Remove the manual gag from salves and CLOSE the following Unit 4 Si test return to RWST valves:
a) CV 4 856A b) CV 4 8568
- 6) Go to Step 12.
b.
Stop the Unit 4 Si pumps and p! ace in standby e
c.
CLOSE the following Si pump discharge header isolation valves:
- 1) MOV 878A
- 2) MOV 878B 12 Reset Containment isolation Phase A And Phase B G
13 Verify Instrument Air To Containment Start all diesel powered air compressors O
j4 Establish RCS Bleed Path:
Verify power to PRZ PORV block' valves a.
Restore power to block valves a.
AVAILABLE b.
Verify PRZ PORV block valves ALL b.
Open block valves.
OPEN c.
Ooen all PRZ PORVs
. cv et c, r
em
~ s... u 9
RESPONSE TO LOSS OF 3.EOP.FR H.l SECONDARY HEAT SINK In/87
(
l STEP ACTION / EXPECTED RESPONSE l
RESPONSE NOT OBTAINED 15 Verify Adequate RCS 84eed Path:
a.
PRZ PORVs BOTH OPEN a.
Perform the following:
- 2) Depressurize at least one intact S/G to atrrospheric pressure using $/G steam dump to atmosphere valves.
- 3) Align any available low pressure water source to the depressurized 5/G(s).
C A U TIO N The RCS bleed ath must be maintained even if RCS pressure remains greater than high head SIpump shutoff head.
16 Maintain RCS Heat Removal:
- MAINTAIN 5t FLOW e MAINTAIN PRZ PORVs BOTH OPEN 17 Check Power Supply To Charging Pumps.
Verify adequate diesel capacity o un OFFSITE POWER, AVAILABLE charging purrps @ necessary, shec suffic.ent nonessentialloads (Refer to E 0, Attacnment O for comp'onent KW load rating) 18 Check if Charging Flo w Has Been Established:
Ch'arging pumps - AT LEAST ONE a Perform the following:
a RUNNING
- 1) @ CCW flow to RCP(s) thermal carrier is lost, THEN isolate seal injection to affected RCP(s) before starting charging pumps
- 2) Start. charging pumps, b.
Establish max @um chargmg frow:
O'
. ca et a.
10 RESPONSE TO LOSS OF
- 3. E O l'. F R H.1 SECONDARY HEAT SINK 6'16/87 STEP ACTION EXPECTED RESPONSE RESPONSE NOT OBTAINEO C A U TI O N If RWST level decreases to less than 1155,0001 GAL, the.51 system should be aligned for cold leg recirculation using ES 1.3, TRANSFER TO COLO LEG RECIRCULA TION.
19 Continue Attempts To Establish Secondary Heat Sink in At least One S/G:
AFW FLOW e STANDBY S/G FE EDWATER PUMP 5 e MAIN FW FLOW '
e UNITS 2, OR 4 FEEDWATER FLOW e CONDENSATE FLOW 20 Check For Adequate Secondary Heat Sink:
n a Narrow range level in at least one 5/G.
a Return to Step 19
(]
GREATER THAN 6% [32%)
Return to Step 19 21 Check RCS Temperatures:
e CORE ExtT TCs-DECREASING e RCS HOf LEG TEMPERATURES -
DECR E A51NG i
~
4
. ca r :t e. t-t
NRC EXAM QUESTION REVIEW O
QUESTION 7.12 Concerning the requirement of at least one boric acid now path to core for boron injection by CVCS, answer the following questions, c) State the piant conditions which must exist to utilize the acceptable ALTERNATE Dow path.
RESPONSE
We request that other answers be considered correct also. The procedure discusses a secondary alternate Dow path which is actually discussed on the page prior to the acceptable alternate now path. In reality, these are both acceptable now paths. Another 3
now path is by using the high head safety injection pumps which take their suction from the RWST. This path is addressed in the Technical Specifications.
O
REFERENCE:
AP-0103.32, Reactor Cold Shutdown Conditions, pgs. 7 and 8 Technical Specifications, Bases section, pgs. B3.6-1 and B3.4-2
):
O
- R 1 -5/j t:01/28/88 l' age 5
12/23/87 ADMINISTRATIVE PROCEDURE 0103.32, PAGE 7 REACTOR COLD SHUTDOWN CONDITIONS CAUTION:
When raising RCS level, verify proper operation of the VCT level makeup system.
Ensure that the makeup water contains the proper boron concentration to prevent the possibility of inadvertent RCS dilution, 8.6.2 To raise RCS level, decrease letdown flow by closing HCV *-142 (VPB)
OR_ by increasing charging flow by increasing speed of running charging pump (s) (console) or opening HCV *-121 (console).
8.7 ti!O N0ZZLE LEVEL OPERATIONS - 'Inen the RCS is drained to mid noz zl e, carefully monitor RHR pump amperage and RHR flow.
Erratic indications on either requires Nnediate operator action to raise RCS level.
If erratic amperage or flow are indicated, perfonn the following:
8.7.1 Adjust FCV *-605 and HCV *-758 (as necessary) to reduce RHR flow as seen on F1 *-605 (VPB).
THEN CAUTION:
Monitor RCS temperature and RHR pump outlet temperature closely to maintain RCS temperature below 190*F.
8.7.2 Raise RCS level as dictated in Step 8.6.2 of this procedure.
8.8 BORIC ACIO Fl.0W PATHS TO THE CORE: When fuel is in the reactor there shall be at least one flow path to the core for boron injection by the CVCS.
O 8.8.1 Durine Coid s8eteown coee4tions. the Preferred fiow Pat 8 to the core is:
BASTS through Boric Acid Transfer Pumps through *-113A and *-1138 (Console) to Charging Pump (s) suction.
Charging pump (s) discharge through their nonnal charging path via HCV *-121 (Console) and
- -310A or *-310B (Console).
If *-113A or *-113B is unavailable, use MOV *-350 (Console) to supply boric acid to Charging Pump (s) suction.
If *-113A is available by *-113B and MOV *-350 are unavailable, use *-113A and
- -356 (local) to supply boric acid to Charging Pump (s) suction, 8.8.2 The preferred alternate flow path to the core is:
RWST through LCV *-115B (VPB) to Charging Pump (s) suction. Charging Pump (s) discharge through their nonnal charging path via HCV *-121 (Console) and *- 310 A or *-3108.
If LCV *-115B is inoperable V *-358 may be used to supply boric acid to the charging pump (sf suction.
8.8.3 A secondary alternate flow path to the core is:
BASTS through Boric Acid Transfer Pumps via *-113A and *-113B to the charging pump (s) suction.
Charging Pump (s) discharge through one or more RCP seal water injection lines.
If *-113A and *-113B is unavailable, use MOV-350 to supply baric acid to the charging pump (s) suction.
If *-113A is available and *-
113B and MOV-350 are unavailable, use *-113A and *-356 (local) to supply boric acid to charging pumps suction.
12/23/87 ADMINISTRATIVE PROCEDURE 0103.32, PAGE 8 REACTOR COLO SHUT 00WN CONDITIONS O
8.8.4 An acceptable Alternate flow path to the core THAT WILL WORK ONLY WHEN THE RCS IS DEPRESSURIZE0 AND ORAINED TO MID N0ZZLE IS:
RWST through *-864A (VPB) and *-8648 (VPB) through *-887 (local) and the alternate low head Si flow path, *-872 (VPB).
8.8.5 Oue to plant configurations and maintenance activities, the above flow paths may not be available.
In this case, an acceptable alternate flow path will need to be identified, and a temporary procedure written identifying the requi red operator actions to establish a boron injection flow path.
An Operator Aid shall be posted to reflect the flow path and then cancelled when the normal flow paths are restored to operable status.
8.9 If less than two Coolant Loops are operable, take immediate corrective action to return at least two Coolant Loops to operation.
8.10 At least one ICW header and basket strainer shall be operable and lined up to supply the CCW Heat Exchangers.
The A ICW Pump shall be operable when the A RHR Loop is required to be operable.
The B or C ICW Pump shall be operable when the B RHR Loop is required to be operable.
At least one ICW Pump shall be capable of being supplied from an operable Emergency Diesel Generator.
8.11 The CCW System has the following operability constraints:
8.11.1 The A or C CCW Pump and one CCW Heat Exchanger shall be operable when the A RHR Loop is required to be operable.
8.11.2 The B CCW Pump and one CCN heat exchanger shall be operable when the B RHR Loop is required to be operable.
8.11.3 Two CCW heat exchangers are required to be operable when 2 RHR Loops are required to be operable.
8.11.4 At least one CCW Pump shall be capable of being supplied from an operable Emergency Diesel Generator.
8.11.5 Two CCW pumps shall be in operation to support the operation of two RHR pumps and heat exchangers. One CCW pump should be in operation for single RHR pump and heat exchanger operation.
NOTE:
The motivation for reducing the CCW System to single pump operation is to keep the system flow through each CCW heat exchanger to less than 6840 gpm (while operating)above the CCW System l ow pressure pump start setpoint.
This administrative flow rate limit for each CCW heat exchanger is a vendor recommendation as delineated in 3/4-0P-030.
O
- **
- 31*. *. v^ '."w
- wa w ? ~.7. e' *. *..* I,
t* *. *... v.a. 9
. 3..a l.'.a. r.=-
a. s* *. a*"v R
- 6.' u..f * * * *v*
"*M a
1.1 4
...s
.n C.....o, 2. 5. n!
.y
.....s
(
- n: ;).:si:a1 and 7olume Control Sys:em provides con:rol of the Rase:or 04:1-ant Syst:n bar:n inven:ory. There are :hree sources of bors:ad '.a:er avai'-
able i::.n'es:1:n through three differen: pa:hs (1)
- he teri: acid transf er punps can deliver the horic acid :ank
- n::nks:o:he:hargingpunps.
(2)
The :harging pumps can :ake al:erna:e su::i:n fr:m :he refueling <a:::
s: rage :ank.
(3)
Tha safe:y injec:1:n pu=ps can :aka :het sue:Lon from the 7: fueling va:er s:orage :ank and injec the boren injecti:n tank con:en:s.
- he quan:1:y :f beri: acid in s::: age fr:n ei: hor :he bert: acid :anks or
- ne refueling a:er s:Orage :ank is suffi:1ent f::.: eld shu:i:en a: any tine during : re life.
One:hannel:jhea: :: acing is suffi:1en: : -ain:ain :he Specifiad I
- =;c t s:ure.i=1:.
s,
' See ref erence (11) on Page 33.4-1 7eference TSAR - See:1:n 8.2 A
d 03.5-1
'. enc =ents 73 5 72
...p.
-e
- * ~
.t**
3.
Ieereenev Centair.mnt Tilterine Sveton p
Two of three filter uni::s have capacity to meet the talA
(
ar.alysis.
4.
Com onent Cooli:2 Syster.
One pu=p and two beat exchssgers neet the requirements of the 12A analysis. (10)
.5.
Intake Coolin veter syste CONTRo**n e n,,...,
One pu=p meets the require:e=. )n.3ant ysis. (6) a.
4 ts of'
.a L..
References:
o (I) TSAR 6.2.2
.(2)
TSAR 14.2.5*
~,
(3) TSAR 14.3.2 (4)
TSAR 14.1.9 (5) 7SAR 6.2.3
-(6) ISAR 14.3.4 (7)
TSAR 6.3 /
(S) 75AR 14.3.5 (9)
TSAR 6.4 (10)
ISAR 9.3 (11) The requirement f or use of the SIT tanks for Mitigation of se Main Steam I.ine Break acciden; has been removed follow'.ng installation of the Model u.T Steam Generators. The required supporting analyses tan be found in U-81 '502), dated 11/3C 11 The temperature rec.uirement above it.5' T is no longer apelicable. Therefore, the heat tracing require =en: is :ot necessary.
There is to !cret Concentration Require =ent in the 317.
33.4-1
.bendments 78 172 1
NRC EXAM QUESTION REVIEW
)
QUESTION 7.13:
List six things you are required to monitor during refueling.
RESPONSE
This is an open-ended question and as such we request that there should be other acceptable answers such as those highlighted in the attached reference material.
REFERENCE:
3-OP-038.1, Preparation for Refueling Activities, pgs. 9-13 3-OP-040.2, Refueling Core Shuffle, pgs. 7-13 O
- R 1 -i'jt :01.28e88 Page ti
,__7 e
m +a..e ne
=w :. : c, 3.O P.038.1 PreparaUon for Refueling Activities 7/23 87 (V3 5.2 Requirements Prior to Core Alterations (Excluding Fuel.\\lovement)
INIT Date< Time Started:
5.2.1 Initial Conditions 1.
The unit is in Slode 6. Refueling.
5.2.2 Procedure Steps g..=....==.==..=.....=.-.=..=..=..==..=.=.=..=..=...-.3 NOTES I
I This section provodes the Technscal Specification requ,rements to be completed fr.r any core alteratton encept fuelmovement.
I l
e if different types of core alterations are beIng performed on succession and l
Technical Specofocatton equipment operabilsty requorements have been I
maintarned. thos section need only be completed for the onettal start of the core I
g altera tsons.
j-6.==.==.==.-..=.==.-.=..==.==..==..=.-..=.==..=.==.==..==.=..=.J 1.
Within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of core alterations, perform the following:
pd a.
Establish containment integrity in accordance with 3 OSP 051.12, Refueling Containment Penetration Alignment. [ Commitment - Step 2.3.1]
(1)
Record date/ time 3 0SP-051.12 is completed.
/
b.
Perform Containment Air Particulate Channels R 311 and R-312 functional test in accordance with 3 OSP 067.1, Process Radiation 31onitoring Operability Test.
3-OSP-067.1 are completed. pplicable sections of Record date/ time the a (1)
/
l c.
Perform the Auxiliary Hoist Load Test in accordance with 3-OSP 038.3. Auxiliary Holst Operability. (Slake N A if l
Auxiliary Hoist will not be used for the Core Alterations.
1 1
(1)
Record date/ time 3-OSP-038.3 is completed.
/
i O
,...+.v.a
2.,,:..^.
.se:...
- e
,9 e:c.* w e 3.O P 038.1 Preparation for Refueling Activities 7!23/87 OV INIT 5 2.2 iCont'd) 2.
Within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> prior to the start of core alterations. perform the following:
a.
Complete 3-OSP 0 59.1. Source Range Nuclear Instrumentation Analog OperauonalTest (1)
Record the date, time 3 OSP 059.1 is completed.
3.
Within i hour prior to the start of core alterations perform the following:
a.
Establish and check the direct communications between the Control Room and the Refueling Canal area.
(1)
Record the date/ time direct communications checked.
i 4.
Prior to the start of core alterations, and during core alterations, commence performance of Attachment 3, Core Alterations
.\\linimum Equipment Checklist.
C AUTlON All core alterations shall be directly supervosed by a Sensor Reactor Operator. on the containment who has no other concurrent responsubsisty durong the core alterations (Commotment Step 2 3 2) 5.
Prior to the start of core alterations, verif-Senior Reactor Operator is stationed in the containment to directly supervise
-~~
the core alterations.
Date. Time Completed:
PERFOR.\\1ED BY (Print)
INITIALS REVIEWED BY:
Plant Supervisor Nuclear or SRO Dengnee
...e....
1,.
- w., : >.
3 012038.1 Preparation for Refueling Actisities 7,23:87
[]
I 5.3 Req uirements Prior to Refueling Core Shuffle INIT Date Time Started:
5.3.1 Initial Conditions 1.
The unit is in Mode 6. Refueling.
5.3.2 Procedure Steps 1.
Complete At'.achment 2, Refueling Equipment Inventory Checklist.
2.
Verify that a satisfactory channel check has been completed for the following area radiation monitors in accordance with 3 OSP-201.1, RCO Daily Logs o2 the monitor has been replaced by a temporary portable monitor equipped with an alarm:
a R 2, Unit 3 Containment Oper Floor b.
R 7, Unit 3 Spent Fuel Bldg Canal Area c.
R 19, Unit 3 Spent Fuel Pit Exhaust Duct d.
R 21 Unit 3 Spent Fuel Bldg North Wall 3.
Verify the Unit 3 S pent Fuel Pit SPING High Range Noble Gas Monitor has been c etermined to be operable by the Chemistry Tests Checks, personnel in accordance with OP.0204.2, P Department and Operating Evolutions.
4.
Perform 3-OSP 034.1, Spent Fuel Pit inlet and Exhaust Damper Operability Test.
5.
Within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of fuel movement, perform the following Technical Specification requirements:
Establish containment integrity in accordance with a.
3-OSP 051.12, Refueling Containment Penetration Alignment. [ Commitment - Step 2.3.1]
(1)
Record date/ time 3 OSP-051.12 is completed.
/
O
-_---..--.__y.
s ee. e.s
' u ea..e am
- m..+ we 3 OP 038.1 l' reparation for Refueling Activities 7/23:87 mb INIT 5.3.2 5 (Cont'd) b.
Perform the manipulator crane load and automatic cutoff test in accordance with 3 OSP 038.2, Manipulator Crane Operability Test.
(1)
Record date/ time 3 OSP 038.2 is completed, c.
Perform the Auxiliary Hoist load test in accordance with 3 OSP 038.3, Auxiliary Holst OperabilityTest.
(1)
Record dateltime 3 OSP-038.3 is completed.
d.
Perform the Containment Air Particulate Channels R 311 -
and R 312 functional test in accordance with 3 OSP 067.1, Process Radiation Monitoring Operability Test.
(Commitment Step 2.3.1]
(1)
Record dateltime the applicable section of 3 OSP d67.1 are completed.
6.
Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the start of fuel movement, perform the following Techmcal Specification requirement:
a.
Complete 3 OSP 059.1, Source Range Nuclear Instrumentation Analog OperationalTest.
(1)
Record date/ time 3 OSP 059.1 is completed.
/
7.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of fuel movement, perform the following Technical Specification requirement:
a.
Verify the refueling canal water level is greater than or.
equal to 56 feet,10 inchesc (1)
Record date/ time refueling canal water level is verified.
/
8.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the start of fuel movement perform the following Technical Specification requirement:
I a.
Establish and check the direct communications between the.
Control Room and the applicable refueling stations.
(1)
Record the date: time direct communications checked.
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o,q, Aco 3 4 :ce 3 O P 038.1 Preparation for Refueling Activities 7'23 87 b,m INIT 5 3.2 (Cont'd i 9.
Prior to the start of fuel movement, verify with the HPSS that Health Physics coverage is available.
- 10. Prior to the start of fuel movement in the reactor vessel, verify the reactor has been suberitical for at least 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />s:
date time reactor was suberitical.
(1)
Record c
(2)
Record date time reactor verified to be suberitical greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
/
O o te' rime co=9 eted:
i PERFORMED BY (Print)
INITIALS REVIEWED BY:
Plant Supervisor. Nuclear or SRO Dessgnee
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3 018040.2 Refueling Core Shufne 8<4;87 0
4.0 l'R E C A UTIO NS/LI.\\llT ATIO N S 4.1 Sliniinum equipment operability requirements, requiring immediate suspension of refueling operations, are listed in Attachment 3. Refueling Core Shufne Slinimum Equipment Checklist.
4.2 All personal items. glasses, pencils, personnel monitors, etc., shall be tied or '
taped to prevent them from falling into the refueling cavity. Each person should inventory personal items he takes into the containment. Do not take any unnecessary items into the containment.
4.3 All applicable radiation protection precautions and procedures shall be observed.
4.4 If. at any time, the Plant Supervisor Nuclear suspects that continued refueling will involve undue risk to personnel or equi,pment, or will compromise the Technical SpeciGcations or license provisions. operations shall be suspended pending resolutions.
4.5 Tools and equipment which are withdrawn from the refueling water should be.
monitored for radiation.
4.6 A minimum level of 56 feet 10 inches shall be maintained in the Spent Fuel Q
Pit at all times.
4.7 The manipulator crane operator shall not unlatch from fuel assemblies whien have been installed in the core until directed to do so by the individual maintaining and evaluating the inverse count rate ratio plot.
4.8 Access to the refueling work stations shall be restricted to members of the refueling team and observers authorized by the Nuclear Watch Engineer supervising fuel movement.
4.9 Prior to lifting any refueling equipment with a hoist. visually inspect the associated hoist take up drums for correct take up sequence.
4.10 >!anning refueling stations:
4.10.1 A!! core alterations shall be directly supervised by a licensed SRO stationed in the containment.(Commitment Step 2.3.11
- 4. l b.2 All personnel designated to operate refueling equipment must complete the applienble rH teling equipment qualineations prior to c
operating the r tueling equipment.
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- 1 0l'.010.2 Refueling Core Shufne 8 87 r-t v
4.10.3 A Nuclear Watch Engineer (NWE) shall be in the containment to supervise fuel movement.
1.
If a problem arises in the Spent Fuel Pit that stoos fuei movement in the containment. the NWE should go to the Scent Fuel Pit to help correct the problem.
NOTE i
I Steos 4.10.4. 410 S. and a 10 6 may ce performed by a qualof:ed vendor.
The manip'ulator crane shall be manned by a designated primary or 4.10.4 alternate. censed Reactor Control Operator iRCO).
1.
The operators are designated by the Refueling Outage Coordinator and the Plant Supervisor - Nuclear.
2.
Prior to operating the Manipulator Crane. non-designated RCO's will be trained until they are sufficiently qualified as determined by the Plant Supervisor - Nuclear and the Refueling Outage O
Coordinator.
G 4.10.5 An operator shal1 be presant to visually verify fuel insertion into the core in accordan,ce with Enclosure 3. Observation of Fuel Assembly Loaded into the Reactor Vessel.
4.10.6 The Spent Fuel Pit Bridge Crane shall be manned by a licensed.>r non-licensed operator.
4.10.7 Reactor Engineering personnel or a licensed operator shall be stationed in the Con ~ol Room to maintain the inverse Count Rate Ratio Plot in accordance with Attachment 1. Inverse Count Rate -
Monitoring During Refueling.
4.10.8 A licensed or non-licensed operator shall maintain communications in the Control Room with the other refueling stations.
4.10.9 The same non licensed operators should man their assigned refueling station for both 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> refueling shifts during the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> refueline schedule to provide continuity.
l 1.
This may be relaxed if other operating evolutions make.:
impossib{e to man a station with the same personnel.
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l 4.11 Lifting of the fuel assemblies by means of tools or adapters attun "
l anemoly shat! be performed with :ne assemnly in the vertical pom n
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- ).01).040 9 Refueling Core Shufne Sr4 87 gO 4.12 Records of all fuel movements shall be kept in the Control Room. Local records shall be maintained of fuel movements on the operating deck of the containment and in the Spent Fuel Pit area.
4.13 The Control Room copy of the fuel movement record serves as the permanent history of fuel transfer and core loading.
4.14 Any inconsistencies in the loading sequence or in the recording of the fuel movements shall be cause to immediately cease fuel movements and notify the Plant Supervisor Nuclear and the Reactor Engineering member on site.
4.15 Changes to the Fuel Handling Data Sheet (Form 5712) have no adverse affects on nuclear safety, nor constitute a change to the intent of the Fuel Handling Data Sheets as long as the final core configuration for the cycle has not been changed.
4.15.1 Changes to a Fuel Handling Data Sheet which are a change ofintent, are made in accordance with AP.0109.3. On The Spot Changes To Procedures.
4.15.2 Changes to a Fuel Handling Data Sheet which are not a change of intent, are made in accordance with Attachment 2, Fuel Handling Data Sheet Changes.
bs) 4.16 Each shift change shall be accompanied by a turnover that fully describes the
~'
status of the equipment at that refueling station including any variations in equipment parameters (e.g. indexing. tape measurements). The oncoming operator shall then evaluate the equipment status and verify that the equipment and lighting are functionally acceptable before initiating any fuel movement.
4.17 The Inverse Count Rate Ratio (ICRR) shall be monitored during fuel movement in the core. If the ICRR falls below.4. the shuffle shall be stopped.
the situation analyzed and the Operations Superintendent. Nuclear notified prior to proceeding.
4.18 Do not move a fuel assembly into the transfer canni from the Spent Fuel Pcol until the lifting frame is in the full up position and verified.
4.19 Prior to any horizontal moves of fuel assemblies, verify the fuel assembly is in the full up position by visual verification.
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V 4.20 Any special equipment or tools used in the reactor cavity or refueling canal area shall be returned to their proper storage place after use to prevent them from interfering with the manipulator crane movement.
4.21 Any small hand tools used to perform maintenance in the reactor cavity or refueling canal areas shall be removed from the cavity or refueling canal after use.
4.22 The fuel loading status board located in. the Control Room shall be kept current for each fuel assembly or insert movement.
4.23 In the event of damage to a fuel assembly, refer to 3 ONOP 033.3. Accidents Involving New or Spent Fuel.
4.24 If communication is lost between any refueling station and the Control Room, no fuel movement at that station shall be initiated until communication is restored.
4.24.1 Fuel must be placed in a safe storage location and unlatched from any refueling equipment until communications is restored.
4.25 Attachment 3, Refueling Core Shuffle Minimum Equipment Checklist shall (UR have been completed prior to movement of fuel and each 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift thereafter, except for the RHR item which shall be done every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.26 All core alterations shall be directly supervised by a licensed Senior Reactor Operator who has no other concurrent responsibilities during this operation.
4.27 During movement of fuel assemblies within the reactor vessel, a minimum cavity level of 56 feet 10 inches shall be maintained and at least two drop lights in the reactor vessel shall be operable.
4.27.1 Use caution when moving lights near the vessel hot leg nozzles as RHR suction could pulllights into the hot legs.
4.23 If lighting failures occur that hinder any visual monitoring specificall:.
required of the operator at that refueling station, the failures shall be corrected prior to initiating any fuel movement at that location.
4.29 Any problems with lighting or operation of any equipment, directly or indirectly involved with the movement of fuel, shall be brought to the attention of the Plant Supervisor-Nuclear for evaluation and resolution.
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4.30 Heavy loads, defined as weight in excess of 2000 pounds, shall be prohibited from travel over irradiated fuel assemblies in the spent fuel pool.
4.31 A copy of Enclosure 3, will be used to remind containment operators of specific observations required during fuel movement.
4.32 The below listed fuel racks do not have their supports directly under the last row of storage cells on the edge of the rack. To arevent these racks from tilting, do not load or unload any fuel storage rac < so that it only has fuel stored along its edge listed in the following table:
ROW AND COLUMN IN SFP LAST ROW AND COLUMN RACKS WHICH MAY OF SFP RACKS WHICH MAY CAUSE RACK TILT BE TILTED 53 53 K 1 V 1 V
53 A -- V 41 A 53 W JZ
.45 W -- JZ A! J 75 AA -- LL 68 AA LL LL 76 AA -- LL 85 AA Fuel assemblies may be in the rows which might cause rack tilt only if there is at least an equal number of assemblies in the same rack two or move rows in from the edge of the rack.
4.33 Fuel may be moved into Region 2 of the Spent Fuel Racks only ifit meets the burnup and enrichment limits specified in the Technical Specifications, Table 3.17-1. Otherwise,it must be moved into Region 1 of the Spent Fuel Racks.
4.34 All personnel engaged in fuel handling activities shall comply with OP-16000.1, Limitations and Precautions for Handling Fuel Assemblies.
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5.0 ST A RTU P NO R31 Al. O PE R ATION 5.1 Refueling Core Shufne INIT Date Time Started:
5.1.1 Initial Conditions 1.
The. unit is in.\\ lode 6. Refueling.
5.1.2 Procedure Steps NOTES I
l Fuel Handling Data Sheet (Form 5712) os shown on Enclosure 1.
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I The fuel Handlong Data Sheet Temporary Procedure format is shown on e.
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1.
Prior to commencing fuel movement verify the Fuel Handling Data Sheets (Form 5712) have been reviewed and approved in dg accordance with AP 0109.6. Temporary Procedures.
a.
Record the Temporary Procedure number.
TP.
2.
Prior to commencing fuel movement, verify 3-OP-038.9.
Refueling Activities Checkoff List is complete through the QC Hold Point prior to Section 3 of Attachment 1.
RE ACTOR SUPERVISOR VERIFICATION POINT Versfy the fuel being moved onto Region 2 of the spent fuel racks has been verofred to meet the burnup and entschment requirements specofted on Technical Specificatoon Tuble 3 17 1.
Vero fie.d by Reactor Supervisor Signature oate 3.
Prior to commencing fuel movement, commence performance of. Refueling Core Shufne.ilinimum Equipment Checklist.
4.
Commence fuel movement and perform :he following:
11aintain the inverse count rate rate ' monitoring in a.
accordance with Attachment 1. Inverse Count Rate
~.\\lonitoring During Refueling.
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.1 01'040.2 Refueling Core Shufne 8 i 87 gV INIT 5.1.2. 4 (Con t'd) b.
Observe fuel assembly loading into the reactor vessel in accordance with Enclosure 3. Observation of Fuel Assembiv Loaded into Reactor Vessel.
c.
Reactor Engineering shall maintain the Fuel Status Record in accordance with AP.0149.1. Special Suelear.\\laterial-Accountability.-
d.
If necessary make changes to the Fuel Handling Data Sheet temporary procedure in accordance with Attachment 2. Fuel Handling Data Sheet Changes.
e.
Document any fuel handling movement delays in accordance with Attachment 4. Fuel Handling.\\1ovement Delays.
5.
After fuel movement has been completed, perform the following:
a.
Perform a core map with an approved copy of the post refueling core pattern in accordance with OP 16900.13. Core Sle.pping Following Core Loading.
I O b.
Stop, performance of Attachment 3. Refueling Core Shuffle i V Slinimum Equipment Checklist.
c.
Notify the on-shift Health Physics Shift Supervisor to perform a radiation survey around the exterior of the Spent Fuel Building to ensure that the radiation levels have not i
increased greater than Imrihr above background due to the j
increased spent fuel storage.
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Date< Time Completed:
PERFORSIED BY (Print)
INITIALS REVIEWED BY:
Plan t S uperut w.e. Nuclear ~ r SRO De~tyn..
O
NRC EXAM QUESTION REVIEW
\\
QUESTION:7.18 the Natural Circulation Cooldown procedure, EOP-ES-0.2, requires the operator to "verify cold shutdown RCS boron concentration" after boration.
NOTE:
Excerpts of EOP-ES-0.2 are enclosed for reference, b) State the indication utilized tojudge the ultimated shutdown condition.
RESPONSE
We request that you accept "monitoring critical safety function status trees" as an additional answer. The reas.n for this request is that the status trees are being monitored at this time and one of the itums monitored is nuclear instrumentation.
REFERENCE:
3 EOP-F-0, Critical Safety Function Status Trees, pg. 5 O
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3.E O P. F.0 CRITICAL SAFETY FUNCTION STATUS TREES 1/7/87 O
ENCLOSURE F 0.1 SUBCRITICALITY iR G,O TO
=.s i GOTO C000000000 0
aa s i NO
- POWER RANGE '
Y LESS TH AN 5%
0 00 y
G..o TO,
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INTERMEDIATE.
NO RAAGE SUR ORE INTERMEDIATE NO NEGATIVE RANGESUR TH AN -0 2 DPM YES
~
'G SAT NO
$OURCE; RANGE ENERGIZED YES GO TO O
y sn.s :
SOURCE NO RANGE SUR ZE40 0R NE G AT!vE itG A MM A.
YES R
= RED PATH VE ACS3 O
= ORANGE PATH c3,
'G SAT Y
= YELLOW PATH G
= GREEN PATH O
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NRC EXAM QUESTION REVIEW r
QUESTION:
8.04 Unit 3 and Unit 4 are operating at 100% power when Boric Acid Transfer Pump (BATP) 3A fails. BATP 4A had failed the previous day, and expected time of repairs on both pumps is in excess two weeks. Assuming all other components operable, select the statement below which most correctly describes the actions required by Technical Specifications.
NOTE; Technical Specifications are enclosed for reference.
a.
Unit 3 or 4 must be placed in hot shutdown, b.
Units 3 and 4 must be placed in hot shutdown.
c.
Unit 3 or 4 must be placed in cold shutdown.
d.
Units 3 and 4 must be placed in cold shutdown.
RESPONSE
O We request that the answer be changed to "c" for the following reason:
You need 2 pumps for single unit operation and 3 pumps for dual unit operation. With 2 pumps out of service for more than 2 weeks (exceeds the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit) we have only 2 operable pumps which meet the requirement for single unit operation.
REFERENCE:
Technical specifications, section 3.6, pgs. 3.6-1 and 3.6-2 l
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Applies Oc :he operatic 41 s:4:us o f :h e Che=1:21 and 7elu=e C::::o1 Systa=.
To defi.:s thosa :: di:10:a of :he Che=1:a1 and 7elu=s Ob*ee:1ee:
5 Con::cl Sys:as necessarf :s ::.sure saf a : tac::: opera:. n.
Seecifica:1cn: a. 'Jben fuel f.s 1: the ::ac:c: : hare shall be a: least out flev path to the core for bo:sn i=j ec:Lon.
- b. A reac:c: shahl not be made critical unless the folloving Chemical and Volume Con::31 Sys:an condittor.s a:s set:
1.
740 associa:ed charging pu=ps shall be perable.
2.
'"40 borte acid ::assis pumps shall be operable.
~
h
~he boric acid :acks is serrica shall :stais a :::21 3.
of a: leas: 3,080 gatens of a, 20,000 :: *0,500 ;p:
- selu:1:n a: 4 :a:pera:ure of 4:
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b*9..e Sys:a= ;1;1:g, it.:::; cks and valres shall be. :pe: :~a to :he extar.: :f as:ab".istas; ::e f:. V pa:h !::= ::e botte acid :ank.s, and :a fiev pa:h f = :he :sfueling va::: storage :ask, :s the 3.aac::: Occian: Sys:a=.
3.
740 cha==ais f.ea: ::ae:.:s small be :pera ' a f:: ::e f1:v pa:h fr== ::a bcit: aci: :2:G.s.
a 6.
- he p:1 ar-r va:a: s :::St ta k :::: aims ::: 1:ss.ta:
20,000 ga ns:s =f va:ar.
- . T e see: 4 ::ac::: shall ::: be ade ::1: :ai unless the l
follevist :::di:12:s a:e :st:
M.ene :ent Nos. ~~ i 57 3.5-1
I 2.
3REZ boric -acid transter pumpe shall be operablo.'
The boric acid :anks in sa:vice shall contain a :::a1 3.
! at least 6160 gallens of a 20,000 :s 22,500 p;s boren solu:1cm at a camparature of at ' e as t 145 T.
Systes piping, tuterlocks and valves shall be opstable 4
to the ex:ent of establishi=g ::a flev path f cm the bort: acid tanks, and one flev path f:ca the refueling va:ar s:orage tank, to each Reactor Coolant System.
5.
!*JO cht,anels of heat :: acing shall be operable f er :he flev path from the beric acid :ank.s.
6.
The prina:7 vater ses; age tank contains not '.ess than 30,000 gal. :s of water.
d.
During power cperatics, the requirenants of 3.6.b a:d :
be nay be =odified to allow one of the foll:vi:g ::= pone::s ::
If the system is not restored to =eet :he-requirements of 3.6b and c vi:his :ha :1.se perted specified, e
- he reac:::(s) shall be placed :n :.e hot shu:dev?. ::n-4:: ion.. if the : equi:ements of 3.6.b and c are not satisfied vi:hin an acdt::enal e3 hours. :he reae:::(s}
shall be placea in the :cid shu:::vn ::nca:i:n.
Specifica:::n 3.0.1 applies :o 3.6.d.
1.
One of the :vo perable chars as pu:ps :ay be removed f;ce service pr=vided tha: 1: is restored :s perable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
One boric acid ::ansf er pump =ay be out of service provided that t: is restored :s operable seseus vi:hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
One h4=nel of hea: :: acing 47 be :u: of service f::
- k neurs.
' Only two moric acid transfer :umes need be ::ers:le :ur :
'; nit 2 L:w ::wer :hysic s :es t'9g 'er Ofci e 3.
~9is :e '::
shall no.: exceed 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> of testing.
Amendment Mo. !!
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P. o Box 14000. JUNO 8E ACH. Pt. 33408 o420 s.%.
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L-86-348 Office of Nuclear Reoctor Regulation At tention:
Mr. Thomas M. Novak, Acting Director Division of Pressurized Water Reoctor Licensing A U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear Mr. Novoks Re: Turkey Point Units 3 & 4 Docket Nos. 50-250 & 50-251 Emergency Diesel Generator Technical Specifications This will confirm our recent discussions with you and your stoff regarding the limiting condition for operation (LCO) opplicable to the operation of the d
emergency diesel generators at the Turkey Point plant. The NRC hos requested, pending submittal and opproval of revised technical specifications for Turkey i
Point in occordance with our Performance Enhancement Program (Project 10),
that on LCO of fixed duration be established by odministrative control. 'The.
current Turkey Point Plant Technical. Specification 3.7.2.b providns for notification to the NRC if a diesel generator outoge is to be seven (7) days or more, but does not require a unit shutdown.
Accordingly, FPL has established interim odministrative controls implementing a seven (7) day LCO in modes I, 2,3 and 4 for the Turkey Point plant emergency diesel generators :in oddition to the current Technical Specification 3.7.2.b requirement for notification. It is understood that exceeding this odministrative LCO would necessitate plocing both units in cold shutdown.
This would not preclude o request for emergency or exigent relief, if conditions warrant.
The basis for our interim LCO is attached.
Very truly yours,
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NRC EXAM QUESTION REVIEW U.-~ QUESTION:
8.08 Concerning the AC electrical operating system requirements, answer each of the following statements TRUE or FALSE.
NOTE:
Technical Specifications are enclosed for reference.
d)
Power operation may e-antinue for a maximum of 7 days if one diesel is out ofservice.
RESPONSE
We request that "TRUE" be accepted as a cearrect answer also IF_the assumption was stated addressing the compensatory action letter which applies to Technical Specification 3.7.2.b.
REFERENCE:
O C.O. Woody letter to Mr. Novak (NRC)
Technical Specifications, Section 3, pgs. 3.7-1 and 3.7-2 O
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>>>' i > := :h av a ': >= : 4 :7 =f t e::=4:41 :c.< r for the opera; ion of auxiliaries.
Obfective:
To define these condi: ions of electrical ;ower availaelity necessary (1) to provice for safe reactor operation, and (2) to provide for the continuing availacility of engineered safety features.
Soecification:
1.
Either reactor shall not be started ft:ra a colo shutdown without:
a.
The associated 23S K'l-4150 vol: start-u; transformer in service.
b.
4150-volt busses A anc 3 of :he associa:e:
unit, and eitner tus A or 3 of :ne secer.d uni,0, energi:ed.
c.
THREI cut of FOLR 480-volt lead centers and 480-vol
.c or control centers A, 3 or C, and 0 of the associated uni; energizec.
d.
TWO diesel generators operable with on site supply of 40,000 gallons of 'uel availabl e.
e.
Four batteries and associ.atec OC sys es are c;erable wi:n FOUR out of SIX battery chargers operable.
2.
During power Operation or restarting fr:m het shutdow the following comoonents may be inoperable:
4.
ONE start-up transfomer my be out of i
service provided both diesel generators are operable.
Die NRC shall be nc f fied
[
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sad te advised Of ;lans to Pts *ere 'ne :Pansfomer
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3.
P wer operatica rey ::ntinue iflui di 0
the rentining diesel generator i O
ts 41 cally and its associated engineered sa ec features are operable, and start-up transformer is oper(able.2) ettner more the NRC shall be notified. d If the-CNC battery may be out of service for a c.
period of twenty f ur hours.
4.
Specification 3.0.1 app 11es to 3.7.2 e
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NRC EXAM QUESTION REVIEW QUESTION:
8.11 State five (5) methods of detecting RCS leakage into containment which would satisfy Technical Specifications.
RESPONSE
We request that you accept ' water inventory change" as an acceptable alternate to D.
REFERENCE:
Technical Specifications, pg. B-3.1-4 O
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- Ill-5/jt:01/28/88 Page 3
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- ns.:ere: :: :e Of a.:ce criance as : - ay.cicue a ::nc'.:!:n is teve!cping *Ma: u:uic cac :: gt:ss
'.eakage. Or:ss '.eakage mu t te prevented :: minim;ce any remote ; css. su:y of
'tiease Of activity :: and from :ne site. Leakage preventi:n Crs: Of all ;r:tects me ;u:.ic and a!sc :: prevents ; :en:!a! ::ntam!. a:!:n Of ~ ;ne equi;..en.
Fremp aintenar.ce anc repair :eads :: :.;reved eda:dity, vnics :s an :verad
- era:ing ::jeenve.
D.us any indication of :eakage; for examples uncalanced water :nven: ries, raciati:n menit:r reading increues, teric acid crfstal to:csits..nsuia::en dampness; snau be considered to :e the result of a feak and snad require
.mmeciate attentien with prompt evalua:!cn required.
iction snail te ;rompt as i: is ;cssib!e mat a smad ;eak may ;r:;aga:e arc
- ec:me a majer :eak. De 'ae: :nat a !eak of 5 g;m, at :ne -aximv.- si;: wee i
react:r c:elant activity, f !t:ased as airbcene material uitncut.. :cy: Or f\\
cleance, would not exceed 10 CFR 20 !imits snau not permit relaxa:!:n of me V
recuirement nat acticn be prompt anc =csitive.
When a real er imagined :eak is detected. :ne ?! ant Su erviser vtU 'mmecia:eiy
'.nitiate a cetailed investigati:n as :o scurce and cause after first c ifying :Me
?! ant Superintendent or his designated a!:ernate. Evaluatien wiu be race :y me
?! ant Su erintendent, who will call upon Production Oe:artment ;ucervisces, sucn u :ne Regional Superir4tendent and me Su erintendent of Generating Stations, as.ecessary for c:nsultation. This prececure is an estariisned and
~
preven :ne in Oceration of feasil fuel fired units unen :caks devefcp, u !! : rings
- tear the,'udgement of experienced persons.
Vhen : e 'eak Mas been ;dentified, me plant management wiU determine :y a safety eva!uation wnetner Operation may centinue. Leskage source (ex. valve stem, pump shaft seaD snau be c:nsidered.
Make up capaciU;y and ;ctential increased demanc snail also be one of :ne evajuation faciers.
s
_ v B3.1 4 Amencment Nos. 95 and 39 e m
NRC EXAM QUESTION REVIEW k._)
QUESTION:
8.19 The following events occurred while Unit 3 is operating at 100% power:
(Date: January 26,1988) 1:00 AAI: Accumulator "A" pressure drops below 600 psig (circumstances such that corrective action will not be complete for one week.)
1:10 AAI: Commenced bringing Unit 3 to Hot Standby.
4:00 AAI: Unit 3 is in Hot Standby.
4:00 PAI: Commenced bringing plant to Hot Shutdown.
6:00 PAI: Unit 3 is in Hot Shutdown.
Answer the following questions. Consider each case separately.
NOTE:
Technical Specifications are enclosed for reference, a)
Were the time limits for any LCOs exceeded by the operators?
b)
By what date/ time must Unit 3 be in cold shutdown?
c)
If Unit 3 were initially in Hot Standby when Accumulator "A" low pressure occurred at 1:00 AAI, by what date/ time must Unit 3 be in cold shutdown?
b If a similar Accumulator"B"pressure drop occurred one hour after Accumulator O
d)
"A" pressure drop occurred, by what date/ time must Unit 3 be in cold shutdown.
RESPONSE
We concur with the answers to A, B and C, but we request that answer d) be changed to Jan. 27/3:00 PAI for the following reason:
Technical specification 3.0.1 applies only after failure of the second accumulator and you have one hour from that time before taking action. Assuming that a shutdown has not commenced, you have until 3:00 PSI on Jan. 27th to be in cold shutdown. (See attached time line)
REFERENCE:
Technical Specifications, section 3 and Bases, pgs. 3.0-1 and B-3.0-1
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T P
B3.0 B ASE5 - LIMITING CONDITIONS FOR OPERATION
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The soecifications of this section provide the general requirements apolicable to each of the Limiting Conditions for Operation within Section 3.
In the event of a disagreement between the requirements stated in these Technical Soecifications and those stated in an aoolicable Federal Regulation or Act, the requirements stated in the applicable Federal Regulation or Act shall take precedence and shall be met.
B3.0.1 The soecification delineates the measures to be taken for those circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a soecification.
For example, Specification 3.4.2.a requires two containment spray pumps to be OPERABLE and provides explicit ACTION requirements if one spray pump is inoperable. Under the requirements of Specification 3.0.1, if both the required containment spray oumos are inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that provided for operation in subsequent lower OPERATION MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S)in which the inoperability is discovered but the times provided for achieving a mode reduction are not applicable if the inoperability is discovered in a mode lower than the applicable mode. For example, if one containment spray pump was discovered to be inoperable while in ST ARTUP, the ACTION Statement would allow up to 109 hours0.00126 days <br />0.0303 hours <br />1.802249e-4 weeks <br />4.14745e-5 months <br /> to O
achieve COLD SHUTOOWN.
If HOT ST ANDBY is attuned in li hours kJ rather than the allowed 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />,93 hours would still be available before the plant would be required to be in COLO SHUTOOWN. However, if this system was discovered to be inoperable while in HOT STANOBY, the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> orovided to achieve HOT STANDBY would not be additive to the time available to achieve COLD SHUTDO'VN so that the total allowable time is reduced f rom 109 hours0.00126 days <br />0.0303 hours <br />1.802249e-4 weeks <br />4.14745e-5 months <br /> to 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.
B3.0.2 This soecification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Ooeration and associated ACTION requirements.
83.0.3 This specification defines t5e applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is orovided to delineate soecifically when each soecification is applicable.
B3.0.4 This specification provides that entry into an OPERATION AL MODE or other specified applicability condition must be made with: (1) the full complement of required systems, equioment, or components OPERABLE and (2) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out-of-service provisions contained in the ACTION statements.
The intent of this provision is to ensure that f acility operation is not initiated with either required equipment or systems inoperable or other soecified limits being exceeded.
O Exceotions to this provision have been ocovided f or a !!mited number of t,g specifications when startuo with inoperable equipment would not affect plant safety. These exceptions are stated ir. the ACTION statements of t5e appropriate soecifications.
B3.0 1 Amendment Nos.114 and 108
1 4 4 j
3.0 LIMITING CONDITIONS FOR OPERATION - APPLICABILITY 3.0.1 When a Limiting Condition for Ooeration is not met, except as provided in the associated ACTION requirements, within I hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
a)
At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,'
b)
At least HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and*
c)
At least COLO SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
There corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual specifications.
This specification is not applicable in \\iODE $ or 6.
3.0.2 Non-compliance with a soecification shall exist when the requirements of the Limiting Condition for Ooeration and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to exoiration of the specified time intervals, completion of the ACTION requirements is not required.
/l 3.0.3 Compliance with the Limiting Conditions for Operation contained in the V
succeeding specifications is required during the OPERATIONAL \\iODES or other conditions soecified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements sha!!
De met.
3.0.4 Entry into an OPERATIONAL \\iODE or other soecified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL \\iODES as required to comply with ACTION requirements.
Exceptions to these requirements are stated in the individual specifications.
3.0.5 For purooses of determining if a component is coerable for LCO considerations, the comoonent need not be considered inoperable due to inoperability of its normal or emergency power suoply if all of its redundant components are operable with their normal or emergency power supplies operable.
' NOTE: Until full conversion to STS, when a LCO action statement requires a unit to be placed in HOT SHtJTOOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, refer to Table l.1 and place the unit on the required status to meet the HOT STANDBY NiODE.
O 1.0-1 Amendment Nos.114 and 108
FEB 03 '98 16:24 001 FPL-JHL 628 P.02 P. O. Son 14@0, NNo BE ACH, F L 33405 0420
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f=AL FEBRUARY 3 MB L-68-55 Dr. J. Nelson Grace Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta ut.,
N.W.,
suite 2900 Atlanta, GA 30323 Dear Dr. Gracel Ret Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Senior Reactor Oeerator Exan Comments Florida Power & Light Company has reviewed the Senior Reactor Operator Upgrade examination presented to Turkey Point operators on January 26, 1988.
Our comments on questions in the examination were submitted for NRC review and consideration prior to grading the examinations in our letter L-88-49 dated February 1,
1988.
The attachment to this letter contains a revised response to Question 8.04.
Should you or your staff have any questions on this information, please contact us.
Very truly yours, yv
- x. O.
vs Vice President C
dy E ecu COW /PLP/gp Attachment cc:
Document Control Desk, USFRC Mr. J. A. Arildsen, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant PLP/001.80L GD,nl9% G >
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n e n esena asa w n
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FEB 03 '98 16829 001 FPL-JNL 628 P.03
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NRC EXAM QUESTION REVIEW OUSSTION:
8.04 Unit 3 and Unit 4 are operating at 100% power when Boric Acid Transfer Pump (BATP) 3A fails.
BATP 4A had failed the previous day, and expected time of repairs on both pumps is in excess two weeks.
Assuuing all other components operable, select the statement below which most correctly describes the actions required by Technical Specifications.
NOTE:
Technical Specifications are enclosed for reference.
I a.
Unit 3 or 4 must be placed in hot shutdown.
l b.
Units 3 and 4 must se placed in hot shutdown.
c.
Unit 3 or 4 must be placed in cold shutdown.
d.
Units 3 and 4 must be placed in cold shutdown.
JtESPONSE:
l.,
We rec[uest that the answer be changed to "c"
for the follow;.ng reason:
You need 2 pumps for single unit operation and 3 pumps for dual unit operation.
With 2 pumps out of service for more than 2 weeks (exceeds the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit) we have only 2 operable pumps which do not aset the requirement for dual unit operation.
During single unit operatton, if the BATP Technical specification requirements cannot be met, the affected unit would eventually be placed in cold shutdown.
Turkey Point interprets this requirement to also be applicable to dual unit operation in that if the Technical Specification requirements for dual unit operation are exceeded, one unit will eventually be placed in cold shutdown.
JtEFERENCE Technical Specifications, Section 3.6, pgs. 3.6-1 and 3.6-2.
PLP/ 001. SOL
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