ML19269E386
| ML19269E386 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/21/1979 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 3--3-A-3, 3-0-3-A-3, NUDOCS 7906280093 | |
| Download: ML19269E386 (38) | |
Text
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Florida P.o. e.. e.._r w
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- d. O May 21, 1979 3
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cs to Mr. James P. O'Reilly, Director U.S. Nucicar Regulatory Commission Office of Inspection and Enforcement 101 Marietta Street, Suite 3100 Atlanta, Ca 30303
Subject:
Docket No. 50-302 Operating License No. DPR-72 I.E. Bulletin 79-05B
Dear Mr. O'Reilly:
Enclosed is our response to Items 5 and 7 of 1.E. Bulletin 79-05B dated April 21, 1979.
Our response describes a generic design for implementation of safety grade reactor trips into the Reactor Protection System at Crystal River Unit 3 for loss of main feedwater and turbine trip.
The detailed design and procurement of this modification will require 12 months following NRC approval of the generic design.
This modification would be installed at the first ref ueling outage or outage of suf ficient duration following this 12 month period.
Item 7 of I.E. Bulletin 79-05B required the submittal of those technical specifications which must be modified as a result of our response to this bulletin.
However, as per our discussions with Mr. Hugh Dance of your staf f, Florida Power Corporation will be submitting changes to the CR #3 technical specifications f ollowing issuance of the SER f or CR #3.
2170 168 7906280093,
G n r i Office 320i inirt tounn street sovin. P O Bon 14042 St Petersbu g Fiorida 33733, 813 - 666 5 t 51 r
a Mr. James P. O'Reilly Page 2 May 21, 1979 Should you require further discussion concerning this submittal, please contact this office.
Very truly yours, FLORIDAPOWERCORPOP.AI40N 1
W, f. BrauTat W.
P. Stewart Manager, Nuclear Operations WPSemhM08 D6 cc:
U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspections Washington, D.C. 20555 File:
3-0-3-a-3 0
169
NEW SAFETY-GRADE REACTOR TRIPS FOR RPS-1 2!70 170
This report describes the implementation of safety-grade reactor trips into the RPS-I for loss of main feedwater and turbine trip.
Loss of Main Feedwater Trip - Control oil pressure switches on both main feedwater pumps will input an open indication to the RPS on feedwater pump trip.
Contact buffers in the RPS will sense the contact inputs and initiate an RPS trip when both pumps have tripped.
This trip will be bypassed below a predetermined flux level, typically 20% FP.
Reference Figure 1.
Turbine Trip - Contact outputs from the main turbine electro-hydraulic control unit will input an open indication to the RPS on turbine trip.
Contact buffers in the RPS will sense the contact inputs and initiate an RPS trip when a turbine trip is indica,ed. This trip will be bypassed below a t
predetermined flux level, typically 20% FP.
Reference Figure 2.
B&W Pressure switches for both trips will be supplied by the customer.
will supply all RPS cabinet mounted equipment. lists the cabinet mounted equipment and gives the trip response time. also gives the contact buffer isolation voltage and the customer requirements for the contact inputs.
Figure 1 is a simplifiad drawing of the main feedwater pump trip.
Figure 2 is a simplified drawing of the turbine trip.
Drawing 51079DGB-1 shows the generic logic for the new trips.
Drawing 51079MLG-1 is a legend for the generic logic drawing.
2170 1/1
CABINET MOUNTED EQUIPMENT FOR ADDITION OF RPS TRIPS ON LOSS OF MAIN FEEDb'ATER AND TURBINE TRIP 3 Contact Buffers 2 Bistables Per Channel 2 Auxiliary Relays j
Modules will be installed in a pre-wired mounting case and tested as a unit prior to shipment. The mounting case is to be installed in an empty row of each RPS channel and connections made to the RPS wiring.
Trip response time of the RPS cabinet mounted equipment will be 1 150 ms.
Isolation of the contact buffer module is 600 volts with the contact input lines not grounded.
Customer contact input requirements:
Continuous 90 ma, P-P Surge 250 ma, P-P Voltage 118 VAC Closed contact indicates pump running Open contact indicates pump tripped 2170 172
TYPICAL RPS CHANNEL EXISTING TRIP STRING c-
__ q ADDED EQUIPMENT g
FOR LOSS OF MFW TRIP l
N MFW PUMP A l
1 l
o TRIPPED l
~
T
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CONTACTS CONTACT l
m BUFFER g
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_L l
0 TRIPPE0 I'
CONTACT CONTACTS BUFFER l
_q
-l 1
I 1
l l
FLUX 2
i j
l BISTABLE L_
T 2/4 TRIP LOGIC RPS TRIP DN LOSS OF MAIN FEE 0 WATER (SIMPLIFIED)
Figure 1
~
if TRIP TO CROCS
~
TYPICAL RPS CHANNEL ExlSTING TRIP STRING c'
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ADDED EQUIPMENT FOR k
1 N
TURBINE T
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Flux I
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T 2/4 TRIP RPS TRIP DN TURBINE TRIP '
LOGIC (SIMPLIFIED)
Figiire 2
~
TRIP TO CROCS
ANTICIPATORY TRIP FUNCTIONS FOR 177 FA PLANTS 2i70 175
4 INDEX
1.0 INTRODUCTION
2.0 ASSESSMENT
OF POSSIBLE ANTICIPATORY TRIPS 3.0 FUNCTIONAL ANALYSIS
4.0 CONCLUSION
S AND
SUMMARY
APPENDIX A: LOSS OF ONE FEEDPUMP ANALYSIS 2170 1,6 7
9,-
1.0 INTRODUCTION
For the purposes of this report, an anticipatory trip is defined as a trip function that would sense the start of a loss of OTSG heat sink and actuate much earlier than presently installed reactor trip signals.
Possible anticipatory trip signals indicative of changes in OTSG heat removal are:
turbine trip, loss of main feedwater, low steam generator 1cvel.
This report evaluates the ef fectiveness of anticipatory trips compared to the existing high RC pressure trip for a LOFW.
Qualitative and elimination of the quantitative arguments are presented which support level trips in the steam generators from final design considerations of anticipatory trips.
Functional response is presented in terms of a parametric study of time to trip.
Thus, irrespective of the plant specific trip signals and actuation time, the hardware design can proceed with greater flexibility. That is, by presenting system parameters, such as pressurizer fill time, as a tunction of time to trip, then if one plant's turbine trip signal occurs 2.1 secs after initiation of the event and another plant's trip signal occurs at 2.5 secs, this study will still be applicable to both.
ECemhM08 D6 2170 177
9 Page 2 Some of the results presented in this report have already been submitted to the NRC in Reference 1*, the balance of the information will be submitted by May 21, 1979.
The analyses are performed with the revised setpoints, i.e., high RC pressure trip at 2300 psig and PORV setpoint at 2450 psig.
It is shown that anticipatory trips provide additional margin between the peak RC pressure af ter the reactor trip and the PORV setpoint, but provide littic additional with margin in the longer term repressurization to the PORV setpoint continued delay of auxiliary feedwater initiation.
2.0 ASSESSMENT
OF POSSIBLE ANTICIPATORY TRIPS In accordance with Bulletin 79-058, item 5, an evaluation for design basis for anticipatory trips on turbine trip, loss of main feedwater, and low steam generator level has been completed.
Ref. 1:
B&W teport " Evaluation of Transient Behavior and Small Reactor Coolant System Breakers in the 177 Fuel Assembly Plant", dated May 7, 1979.
2170 178
S**"" *
=<
Page 3 The evaluation showed low steam generator level not to be anticipatory and, therefore, it has not been recommended as an anticipatory trip Figure 2-1 shows the OTSG startup level f rom site data and function.
the CADDS ca.Aulated OTSG mass inventory as functions of time following the TM1-2' event. The time of reactor trip on high RC that a steam pressure is noted on the figure and clearly demonstrates generator level trip would not have been anticipatory for a level setpoint that would not interfere with normal operations and The initial rapid f all in OTSG level occurs as the turbine maneuvers.
of the stop valves close, momentarily stopping steam flow out The mass inventory increases during this period due to generators.
the loss of flow f riction a P.
I,y the time the reactor trip occurs, at 8 secondr., steam flow is re-established through the bypass system, flow fr!.ction aP re-establishes the level and both mass and measured level
.ta rt to decrease uniformly.
An OTSG level trip set to trip on the initial drop shown in Figure 2-1 would need to be set restrictively high f or normal plant maneuvers and/or lower power levels.
Further level information (in terms of mass inventory) is given in the figures for the analysis in Section 3.0.
The results for those cases also indicate that the steam generator low level trip fu,nction would not be sufficiently fast to be considered anticipatory.
ECSemhM08 D6 2170 179
. r,.
F igur e. 2 -1 LOFW (TMI-2 EVENT) 2.0 160 TRIP DN HIGH RC 1.6 PRESSURE
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--- -- TOT AL S/0 MASS, CADOS START-UP LEVEL, TMI-2 SITE DATA 2170 180
Page 4 Anticipatory trips for loss of feedwater and turbine trip can be designed to trip the reactor in a more expedient manner than the high RC pressure trip f or some overheating transients.
An anticipatory trip will provide more margin to PORV setpoint during the initial overpressurization resulting f rom loss of feedwater and/or turbine trip.
These trips will provide slightly more time to PORV setpoint and pressurizer fill for delayed auxiliary feedwater initiation conditions.
3.0 FUNCTIONAL ANALYSIS A series of LOFW evaluations was performed at 100% lull power (2772 MWt) with a reactor trip assumed on an anticipatory signal.
With the new high RC pressure setpoint of 2300 psig, a reactor trip would be expected at about 8 seconds af ter the LOFW.
The anticipatory trip study considered reactor trips with 0.4 sec, 2.5 sec, and 5 see delays f rom time zero.
These studies also included sensitivities to AFW failure and reactor coolant pump coastdown.
The anticipatory trip study modeled a generic 177 FA plant, and is considered applicable to raised or lowered loop designs.
A feedwater coastdown similar to that estimated to have occurred at the March 28th THI-2 event was used to generate sepsrate heat demands for each CADDS ECSemhM08 D6 2i70 181
.=,
Page 5 analysis. The heat demands will change as the reactor trip time is delayed, because the additional heat input will boil of f the fixed steam generator inventory at dif f erent rates.
For the cases where" AFW flow was modeled, 1000 gpm was assumed, starting at 40 seconds. With proper steam generator level and pressure control, the system parameters will begin to stabilize at 195-290 seconds, depending on trip delay time and RCP operation; see Tabic 3-1 and Figure 3-12.
The PORV will not be actuated, nor would the pressurizer fill or empty.
With the assumption of no AFW, the PORV will be actuated about three minutes into the event, as a result of system swell; the pressurizer fills as 10-12 minutes (see Tabic 3-1).
A delay of reactor trip of 2-3 seconds is seen to reduce PORV time to actuate by about one minute, and pressurizer fill by about 2 minutes.
For PORV setpoints other than 2450 psig, the times will vary and can be determined f rom Figures 3-3, 3-4, 3-8 and 3-10.
In each of these cases, the mass addition and cooling effect of expected make-up system operation is not rudeled.
One make-up pump running will add about 10 inches per minute to pressurizer level, and
~1/2% heat de ma nd.
It should be noted that the May 7 report used a ECSemhM08 D6 2170 182
- r Page 6 heat demand which reproduced the TMI-2 LOFW event; it has been reported by the operator that two make-up pumps were running f rom 13 see into the event, creating a higher heat demand than the assume.
This difference is anticipatory trip studies of the report shown in Figure 3-l'2.
The steam generator heat demands, reactor power, RC system pressure, pressurizer level, and RC inlet / outlet temperatures are given in Figures 3-7 through 3-11 for the trip on high RC pressure (t=8 see)
The ef f ects of delayed auxiliary feedwater initiation are also case.
shwon on the high RC pressure trip curves.
2170 183 ECHemhM08 D6
TABLE 3-1 LOFW EVENT (LOFW at T=0 sec)
TIME OF REACTOR REACTOR AUXILIARY PORV PRESSURIZER S/G Li. VEL CD J)J0L TRIP (" DELAY")
C00LAST PUMPS FEEDWATER OPERATES FULL (400")
(P
=1025 psig) stm 195 sec 0.4 Run at 40 sec 225 sec 2.5 Run at 40 see 275 sec 5.0 Run at 40 see 0.4 Run None 235 sec 790 see 2.5 Run None 180 sec 685 see 5.0 Run None 140 sec 575 see 255 0.4 Coastdown at 40 see 0.4 Coastdown None 190 sec 700 sec LOFW EVENT - TIME =0 sec REACTOR TRIP AT 2300 PSIG TIME OF TRIP RCP IJV PORV PRESS. FULL S/C LEVEL CONT.
260 sec 8.0 Run at 40 sec 8.0 Run None 175 sec 620 see 2170 184
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e LOSS OF FEEDIATER AT T = 0 SEC NO AUXILIARY TEEDIATER t
i b' l
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,i 800 PRES $URl2ER flLLS j
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9 0.4 1.0 7.0 3.0 4.0 5.0 6.0 Tip: to reactor trip ("telay"). seconos 2170 196
4.0 CONCLUSIMS AND
SUMMARY
I spectrum of delay times, representing anticipatory trips, has been anal; ad for the loss of feedwater transient. The spectrum included trips at time zero, with a 0.4 second instrument delay, up to high RC pressure trip at time 8.0 secoads.
Since a high RC pressure trip occurs. cry soon after a leis of heat sink (overpressurization) transient from 100% FP, only turbine trip and direct loss of feedwater detection trips would be considered Anticipator;>.
For all *. rips considered, including high RC pressure, the PORV is not actuated when normal system operations occur.
The pressure rise in the primary side is less for the anticipatory trips providing additional margin to PORV lift.
If auxiliary feedwater is significantly delayed, then an anticipatory trip will, at best, provide about 1 minute additional time to PORV lift and about 3 minutes additional time to filling of the precsurizer.
These results can be seen in Table 4-1 which shows the segocnce of events for a LOFW transient with trip on high RC pressure (2300 psig) and trip at time zero.
2170 197
TABLE 4-1.
LOFU-SEQUENCE OF EVENTS COMPARISON 40-s 120-s TRIP AT ZERO, EVENT AFW DELAY NO AF'a' NO ADJ Loss of feedwat.er iaitiated 0
0 0
0 (trip occurs)
(0.4 delay)
High-pressure trip (2300 psig) 8 8
8 a
PORV opens (2450 psig) a a
175 235 Peak RCS pressure 10 10 175 235 Pressurizer full a
a 620 790
- Does not occur for these cases 2170 198
e APPENDIX A LOSS OF ONE FEEDWATER PUMP A special analysis was performed at the B&W Owner's Group request.
This analysis considered the loss of one main feedwater pump with the plant operating at 100% FP, RC pumps running, no power runback or auxiliary feedwater initiation and a RC high pressure trip setpoint at 2300 psig.
The base parameters for this study are the same as those used in the " realistic" analysis presented in Section 4.2 of the May 7, 1979, B&W Repott for 177 FA plants.
The objectives of this study were two-fold:
- 2) Determine if the OTSG level would be a viabic anticipatory trip, i.e., how rapidly does the steam generator inventory decrease in relation to the time a high RC pressure trip would occur.
RC system pressure and pressurizer level as functions of time are shown in Figu'es A-2 and A-3, respectively.
Reactor trip occurs on high pressure (2300 psig) in 15.8 seconds af ter the loss of one main feedwater pump.
Figure A 4 shows the steam generator mass as a function of time and only 30% of the mass is boiled off by the time the reactor trip occurs. This is insuf ficient inventory decrease to cause a level trip in an anticipatory mode.
Figure A-2 shows that no PORV actuation results from this transient.
2170 199
~
Figure A-1 FEEDWATER COAST 00WN TO 50'J l
I i
i 100 90 -
80 -
70 -
e 60 g
~
50
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40 30 20 10 N
2170 200 0
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20 40 60 80 100 Time, seconds
Figure A-2
.FEEDWATER C0AST00WN TG 505 2400 2300 -
E a
a; 2a U
oc 2200 w_
2l00 I
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5 10 15 20 25 Time, seconds 2170 201
Figure A-3 FEEONATER C0ASTDOWN TO 50',
240 3
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52 C
2 Z
220 5
02 210 200 I
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Figure A-4 LOSS OF ONE FEEDPUMP-RAMP TO 50:; IN 10 SEC TOTAL S/G MASS, CADDS i
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