ML20138N035
| ML20138N035 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/31/1985 |
| From: | Mcdonald R ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.05, TASK-TM GL-85-12, NUDOCS 8511050027 | |
| Download: ML20138N035 (12) | |
Text
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Malling Address Alatrm a Power Company 600 North 18th Street Post Office Box 2641 Birm;ngham. Alabama 35291 Telephone 205 783-6090 R. P. Mcdonald Senior Vice President Flintridge Building Alabama Power October 31, 1985 Docket Nos. 50-348 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Automatic Trip of Reactor Coolant Pumps - Generic Letter 85-12 and NUREG-0737, Item II.K.3.5 Gentl emen:
Generic Letter 85-12 requested information required by the NRC to complete their plant specific reviews of the NUREG-0737, Action Item II.K.3.5.
Attached is Alabama Power Company's response to this request.
If there are any questions, please advise.
Yours very truly, I)... N 7Y R. P. Mcdonald RPM /J AR:ddb-043 Attachment cc: Mr. L. B. Long Or. J. N. Grace Mr. E. A. P,eeves Mr. W. H. Bradford Su2!!B8EBn8lhe
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4 ATTACHMENT Response to NRC Questions Cbncerning Automatic Trip of Reactor Coolant Pumps (RCP)
NRC-Question A.1:
Identify the instrumentation to be used to deter.aine the RCP trip setpoint, including the degree of redundancy of each parameter signal needed for the criterion chosen.
APCo Response A.1:
l Two subcooling margin monitors along with their associated
' instrumentation inputs are used to determine the RCP trip setpoint.
The instrumentation inputs utilize measurement of Reactor Coolant System (RCS) pressure and temperature and include the following channels for each subcooling monitor.
a.
RCS pressure (i) two channels of wide range RCS pressure (ii) one channel of pressurizer pressure b.
RCS temperature (1) two channels of wide range RCS hot leg temperature (ii) two channels of wide range RCS cold leg temperature (iii) eight channels of core exit thermocouples (TC)
NRC Question A.2:
Identify the instrument uncertainties-for both normal and adverse contai nment 'condi tions. Describe the bases for the selection of the adverse containment parameters. Address, as appropriate, local conditicas such as fluid jets or pipe whip which might influence the instrumentation reliability.
APCo Response A.2:
The Unit 1 subcooling monitor uncertainties at 400 psig are 22.5 *F for normal containment and 87.3 'F for adverse containment. The Unit 2 subcooling monitor uncertainty at 400 psig is 22.5 *F for normal containment and at 585 psig is 226 *F for adverse containment. The difference in the uncertainty for Unit 2 and Unit 1 adverse containment is dre to use of Rosemount transmitters for wide range RCS pressure in Uni'.1 in lieu of Barton Lot 2 transmitters. Modifications are sr.neduled for the next refueling outage for Unit 2 to replace the darton Lot 2 wide range RCS pressure transmitters with Rosemount transmitters at which time the uncertainties for ecch unit will be i
l Attachesnt Response to NRC Questions Concerning Automatic Trip of Rtactor Coolant Pumps (RCP)
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' APCo' Response A.2:
(continued) identical. These uncertainties consider all error components from the.
sensarlto the subcooling monitor display and are based on a statistical cor.oination of error components for the individual instrument channels (pressure and temperature measurement), as well as for the subcooling monitor hardware and software. Transmitter uncertainties for adverse.
- containment conditions are based on manufacturers' qualification test -
data in which the transmitters have been tested under conditions (temperature, pressure, radiation) which envelope those calculated for -
.the intended plant conditions.
- The-transition point at which the adverse containment uncertainties are -
employed. in the detennination of subcooling margin uncertainty is a' function of containment radiation and temperature or pressure. A switchover radiation of 105 Rad /hr is used and is discussed in the WOG ERG' Revision 1 Executive Volume. As discussed in this document, radiation exposure tests have shown that sensors generally exhibit insignificant' post-accident errors for integrated radiation doses of 106 Rads. Hence; use of a containment radiation level of 105 Rads /hr to initiate switchover to adverse containment parameters ensures a considerable period of time until a 106 Rads total dose is received. A switchover containment temperature of 180 *F is currently used for Unit 1.
This is a conservative value based on the operational characteristics of transmitters. Based on the results of a recently completed evaluation.of the Rosemount transmitters, a new Unit 1 switchover criterion of 4 psig will be used after appropriate procedures are revised. This new criterion is conservative and will ensure consistency between Units 1 and'2. Barton Lot 2 wide range RCS pressure transmitters are currently utilized for Unit 2 with a switch-over containment pressure of 4 psig. Upon 1.nsta11ation of the Rosemount transmitters in Unit 2, the switchover criterion will reaain 4 psig.
All of the instrumentation channels identified in resoonse to NRC Question A.1, with the exception of the eight channels of core exit TCs, were reviewed for the effect of adverse location conditions.
The core exit TCs will be replaced in accordance with Alabama Power Company's commitments related to R.G.1.97.
Fluid jets from or pipe whip of the primary side RCS piping could potentially impact this instrumentation; however, all impacts were determined to be in the t
conservative direction.
Although both chennels of the subcooling monitor could be affected by adverse local conditions, the probability of this happening is low due to the physical separation of the instrumentation and impact would be in the conservative direction. The most likely resalt of an adverse local condition would be a LOCA situation in which RCP trip is desired.
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l Attachment L
Response to NRC Questions Concerning Automatic Trip of Reactor Coolant Pumps (RCP) t Page 3 i
If the adverse local conditions occur during a Steam Generator Tube Rupture (SGTR) or non-LOCA event, RCP trip is acceptable as it is part of the dc:;ign basis for Farley Nuclear Plant.
NRC Question A.3:
In addressing the selection of the criterion, consideration to uncertainties associated with the WOG supplied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not representative of the generic data groups.
If a licensee determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-specific procedure be developed. For example, use of the NRC-required inadequate-core-cooling instrumentation may be useful to indicate the need for RCP trip. Licensees should take credit, for all equipment (instrumentation) available to the operators for which the licensee has sufficient confidence that it will be operable during the expected conditions.
APCo Response A.3:
The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses. Both SGTR and non-LOCA event were simulated in these analyses. LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and non-LOCA analyses. The code has been validated against the January 198/. SGTR event at the Ginna plant. The results of this validation show that LOFTRAN can accurately predict RCS pressure, RCS temperature and secondary pressures accurately especially in the first ten minutes of the transient which is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism, assuming no changes in the initial plant conditions (i.e., full power, pressurizer level, all appropriate SI and AFW pumps run), in the computer program results, are due to either models or inputs to LOFTRAN. The following are considered to have the most impact on the determination of the RCP trip criteria:
1.
Break flow 2.
Safety Injection (SI) flow 3.
Decay heat 4.
Auxiliary Feedwater (AFW) flow
s -
Attachment Response 'to NRC Questions Concerning Automatic Trip of Reactor Coolant Pumps (RCP)
Ph'e 4 I
- The following provides an evaluation of the uncertainties associated with each of these items.-
l i
To conservatively simulate a doubl2 ended tube rupture in safety analyses, the break flow model used in LOFTRAN includes a substantial amount of conservatism-(i.e., predicts higher break flow than actually expected). Westinghouse has performed analyses and developed a more realistic break flow model that has been validated against the Ginna
.l.
SGTR-data. - The break flow model used in the WOG analyses has been shown to be approximately 30% conservative when the effect of the higher predicted break flow is compared to the more realistic model.
The consequence of the higher predicted break flow is a lower than expected predicted minimum pressure.
i The SI flow inputs used were derived from best estimate calculations, assuming all SI trains (one pump per train) operating. An evaluation of the~ calculational methodology shows that these inputs have a maximum i
uncertainty. of +10%.
- The decay heat modes used in the WOG analyses was based on the 1971 ANS 5.1 standard. When compared with the more recent 1979 ANS 5.1 decay heat inputs, the values used in the WOG analyses are higher by about 5%.
T. determine the effect of the uncertainty due to the decay heat mode!. A sensitivity stu@ was conducted for SGTR. The results of this stu@ show t' tat a 20% decrease in decay heat resulted in only a 1%
-decrease in RCS pressure for the first 10. minutes of the transient. -
Since RCS temperature.is controlled by the steam dump,.it is not affected by the decay heat model ' uncertainty.
The AFW flow rate input used in the WOG analyses are best estimate values, assuming that all auxiliary feed pumps are running, minimum pump start delay and no throttling. To evaluate the uncertainties with P
AFW flow rate, a sensitivity study was performed. For a two loop plant, a 64% increase in AFW flow resulted in only an 8% decrease in minimum RCS pressure, a 3%' decrease in minimum RCS.subcooling, and an 4
8% decrease in minimum pressure differential. For a 3 loop plant l
study, a 27% increase in AFW flow resulted in only a 3% decrease in minimum RCS pressure, a 2% decrease in minimum RCS subcooling, and a 2%
decrease in pressure differential.
4 The effects of all these uncertainties in models and input parameters were evaluated and it has been concluded that the contributions from the break flow conservatism and the SI uncertainty dominate, The calculated overall uncertainty in the W0G analyses (which include both uncertainties in the computer program results and uncertainties i
resulting from Farley specific features not representative of the i
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Attachment Response to NRC Questions Concerning Automatic Trip of Reactor Coolant Pumps (RCP)
Page 5 generic data groups) as a result of these considerations is -3 'F to
+20 'F for the RCS subcooling RCP trip setpoint. Due to the minimal effects from the decay heat model and AFW input, these results include only the effects of the uncertainties due to the break flow model and SI flow inputs.
NRC Question B.1:
Assure that containment isolation, including inadvertent isolation, will not cause problems if it occurs for non-LOCA transients and accidents.
a.
Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LOCA situation is confirmed to prevent seal damage or failure.
b.
Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
APCo Response B.1:
It should first be noted that the Farley units do not lose seal
' injection from the Chemical Volume Control System (CVCS) to the RCPs when containment isolation occurs. Component Cooling Water (CCW) flow to the pump seals and to the pump motor, however, is losc when Phase B isolation occurs.
The values of the Farley RCP trip paraceter (i.e., RCS subcooling) for both normal and adverse containment conditions are such that, upon installation of the Rosemount transmitters in Unit 2, acceptable discrimination capability will exist to prevent RCP trip for SGTR and non-LOCA events. This capability provides sufficient margin so that RCP continued operation for SGTR and non-LOCA events can be assured.
In the case that the operator does trip the pumps for one of these events, he is not required to restart them since RCP operation for non-LOCA and SGTR events is not mandatory.
If water services needed for RCP operation are terminated due to a Phase B containment isolation, seal injection to the seals will not be terminated, but CCW flow to both the RCP motor and seals will be lost.
Under these circumstances, the Westinghouse Owners Group (WOG) position is that the RCPs should be stopped. Even though flow continues to the seals because of seal injection and seal integrity will be assured, the motor may be damaged if operation continues. Also, since Phase B containment isolation is activated at a containment pressure of 27 psig, in all probability this was caused by a LOCA or secondary break for which loss of the RCPs will not adversely affect plant safety.
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. Attachment-L Response to NRC Questions Concerning Automatic Trip of Reactor Coolant Pumps (RCP)
H Page 6 L
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'If the Phase B containment isolation was inadvertent and not indicative of a high containment pressure, and if a non-LOCA or SGTR event is in progress, the RCPs should be. tripped. - As with the actual Phase B isolation, loss of the RCPs will not adversely affect plant safety..
Consequently, if RCP. operation is still desired, restoration of water services may be performed at the discretion of the operator and as I
other plant recovery actions dictate.
l If CCW is lost for reasons other than Phase B containment-isolation, p
the RCPs should be tripped.
NRC Question B.2:
Identify the components' required to trip the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when determined to l
be necessary, will. occur.
If necessary, as a result of the location of any critical component, include.the effects of adverse containment i
conditions on RCP trip reliability. Describe the bases for the adverse containment parameters selected.
APCo Response B.2:
The key components of the manual RCP trip circuitry are listed below.
Unit 1 Breaker SWGR Main Control Termination Remote
' Control RCP Cell Board Cabinet Handswitch Power
-1A DA04 nihil Q1H2BL005-A N1841 N1R41 NGMCB2500C-AB HS2112A-N LOO 1G-N 1B DB03 nihil Q1H2BLO25-B N1B41 N1R41 NGMCB2500C-AB HS21128-N LOO 1H-N 1C
-DC03 nihil Q1H2BL005-A N1841 N1R41 NGMCB2500C-AB HS2112C-N LOO 1G-N l
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Attachment Response'to NRC Questions Concerning Automatic
-Trip of Reactor Coolant Pumps (RCP)
Page 7 Unit 2 Breaker-SWGR Main Control
- Termination Remote Control RCP Cell Board Cabinet Handswitch Power 2A!
LDA04 N2H11 Q2H2BL005-A N2B41 N2R41 NGMCB2500C-AB HS2112A-N LOO 1H-N 28 DB03-N2H11 Q2H2BLO25-B N2B41 =
N2R41 NGMCB2500C-AB HS21128-N LOO 1G-N 2C DC03
_N2H11 Q2H2BL005-A N2B41 N2R41 NGMCB2500C-AB HS2112C-N
. LOO 1H-N All components and cables required to trip the RCP's using the handswitches in the Control Room are outside containment and not exposed to adverse containment conditions. RCP trip reliability is therefore not affected by adverse containment conditions and is assured.
NRC Question C.1:
Describe the operator training program for RCP trip.
Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running.
APCo Response C.1:
The FNP criteria for tripping of the RCPs are based on the WOG Emergency Response Guidelines (ERGS).
In general, the RCPs are not-tripped unless previous analysis has indicated that further RCP operation would exacerbate the accident situation.
RCP trip criteria have been developed and incorporated into the ERGS to provide for RCP trip when required for SBLOCAs and to minimize the probability of RCP trip when not required. The RCP trip criteria consist of two fundamental parts:
1)
Successful operation' of the safety injection system 2)
Selected plant parameters reaching critical setpoints.
s
' Attachment k
Response to NRC Questions Concerning Automatic
_. Trip of Reactor Coolant Pumps (RCP)
. t-Page 8
-In the Optimal' Recovery Guidelines, the RCPs are not tripped unless
'the above two-part criteria are satisfied.- At least one high pressure SI pump must.be in _ operation and capable of delivering flow to the RCS
- before the RCPs are tripped.
If this condition is not met, the RCPs should.not be. tripped regardless of whether or not;the plant parameters
- indicate that a setpoint reqJiring RCP trip has been -eached. Analysis
'has shown that if the SI system is not in operation, the RCPs.can be operated to provide core heat _ removal. For SBLOCAs with the high-head safety injection pumps not in operation, the RCPs. continue to provide core heat iemoval via the. break and the steam generators. With the RCPs' running, the RCS can be depressurized to the point where the accumulators and ~the low-head safety injection pumps can ensure core heat removal before symptoms of inadequate core cooling are exhibited.-
1 A' parameter and a corresponding setpoint for operator action has been established to ensure that the RCPs ~are tripped early in a SBLOCA transient.. However,. the use of an overly conservative parameter and setpoint could also result in RCP trip for a SGTR and other non-LOCAs for which it is desirable to keep the RCPs running. It is desirable L keep the RCPs running during these transients to:
- 1) maintain normal pressure control using pressurizer spray and thereby avoid opening of the pressurizer PORVs, 2) prevent the formation of a stagnant water
-volume in the upper head region which may flash and form a steam bubble during subsequent cooldown and depressurization, 3) minimize potential
.c. '.
pressurized thermal shock challenges, and 4) minimize operator actions such as tripping the RCPs and then restarting them 1ater. Although it is. beneficial to keep the RCPs running during a SGTR or non-LOCA event, tripping the. RCPs would not violate any safety criteria since the design of plant safety. systems and the FSAR analysis for these accidents are based on concarrent loss of offsite power, and therefore on RCP trip.
RCP trip at FNP is based on RCS subcooling. This allows longer RCP operation. Subcooling is also a parameter which is readily available to 'the operator.. RCP operation following a SBLOCA does not lead to excessive RCS liquid inventory loss through the break until the time is reached when tripping the RCPs would cause the break to uncover y
immediately. The break cannot be uncovered until a significant amount
,4' of-voiding has occurred in the RCS.
Since it is expected that voiding will' occur first at the core exit, it is not necessary to trip the RCPs as long as subcooling;is maintained in the RCS hot legs. RCS subcooling based on either the wide range hot leg RTDs or the core exit
.TCs, as. appropriate, can be used for this purpose.
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f Attachment I
Response to NRC Questions Concerning Automatic
~
Trip of Reactor Coolant Pumps (RCP)
Page 9 p.
L
'The operator training program for RCP trip criteria has included the
' following:
l-
- 1) Licensed Operator Retraining prior to implementation of ERG I-procedures included:
AP-74 Administrative guideline for ERG development including emphasis on mechanics of style.
EEPs 0,1, Half de classroom discussions followed by
-2, 3
half day simulator sessions.
RCP trip Discussion included SBLOCA analysis Criteria referencing ERG Generic Issue RCP Trip / Restart, WCAP-9584 and WCAP-9600.
- 2) Licensed Simulator Retraining for 1985 includes an approximately two' hour review lecture concerning RCP trip criteria and the SBLOCA accident.
3)
Initial License Training. teaches all the major Emergency Response
- Procedures'(ERPs) and their bases (e.g., RCP trip criteria).
4)
Initial License Training provides classroom and simulator sessions on all major ERPs'where RCP trip criteria is applicable.
'NRC Question C.2:
Identify those procedures which include RCP trip related operations:
(a) RCP trip using WOG alternate criteria
.(b) RCP restart L(c) Decay heat removal by natural circulation
- ( d). Primary system _ void removal (e) ;Use of steam generators with and without RCPs operating (f) RCPl trip for other. reasons
Attachment Response to NRC Questions Concerning Automatic Trip of Reactor Coolant Pumps (RCP)
Page 10 APCo Response C.2:
Identified below are those procedures which include the RCP trip related operations heading each list:
a.
RCP trip using WOG alternate criteria (1)
EEP-0 (ii)
EEP-1 (iii) EEP-3 (iv)
ECP-2.1 (v)
ESP-0.0 (vi)
ESP-0.1 (vii) ESP-0.2 (viii) ESP-0.4 b.
RCP restart-
-(i)
ESP-0.1 (ii)
ESP-1.1 (v)
ESP-1.2 (vi)
EEP-3 (vii) ECP-2.1 (viii) ECP-3.1 (ix),
ECP-3.2 (x)
ECP-3.3 (xi)
FRP-C.1 (xii) FRP-P.1 (xiii) FRP-I.3 c.
Decay heat removal by natural circulation (1)
ESP-0.1 (ii)
ESP-1.1 (v)
ESP-1.2 (vi)
EEP-3 (vii) ECP-0.1 (viii) ECP-2.1 (ix)
ECP-3.1 (x)
ECP-3.2
. ~ ~ '
(xi)
ECP-3.3 d.
Primary system void removal (i)
FRP-I.3 u____._________
Attachment Response to NRC Questions Concerning Automatic Trip of Reactor. Coolant Pumps (RCP)
Page 11-APCo Response C.2:
(continued) e.
Use of steam generators with and without RCPs operating Steam generators are used 'throughout the Farley Procedures set.
f.
RCP trip.for other reasons
[1] Spray ' valve failure to open (i)
EEP-0 (ii)
ESP-0.1 (iii)
ESP-1.2 (iv)
.ECP-3.1 (v)
FRP-I.1
[2] Phase B actuation (1)
EEP-0 (ii)
FRP-Z.1
[3] Lack of proper support conditions (1)
ESP-1,2 (ii)
ESP-3.1 (iii)
ESP-3.2 (iv).
' ESP-3.3 (v)
ECP-1.1
'(vi)
ECP-2.1-(vii)-
ECP-3.1 (viii) ECP-3.2
~(ix)
ECP-3.3
~[4] Other reasons (i)
FRP-C.1 (ii)
FRP-C.2 (iii)
FRP-H.1 2