ML20236G326
| ML20236G326 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/23/1987 |
| From: | FLORIDA POWER CORP., GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT |
| To: | |
| Shared Package | |
| ML20236G313 | List: |
| References | |
| NUDOCS 8711020498 | |
| Download: ML20236G326 (72) | |
Text
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' LOADING EVALUATION FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 10-23-87 l
- B711020498 871026 PDR ADOCK 05000302 P PDR i
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TABLE OF CONTENTS j I. EXECUTIVE
SUMMARY
II. MECHANICAL METHODOLOGY III. ELECTRICAL METHODOLOGY )
IV. RESULTS 1
V. TABLES VI. APPENDICES l
, A. Scenario Writeups I
B. Modification Descriptions l
C. Safety Evaluations of Modifications i
i
I. EXECUTIVE SUMM ARY This report presents an evaluation of the present loading on the Emergency Diesel Generators at Crystal River Unit 3 to assess their capability to handle load in the event of a Design Basis Accident requiring engineer safeguards actuation coincident with a Loss of Offsite Power.
The following criteria were established for this evaluation.
- 1. The loading calculation would be performed on Emergency Diesel Generator 3A due to its requirement to supply power to the Motor Driven Emergency Feedwater Pump. There is no corresponding load on Emergency Diesel Generator 3B as the Redundant Emergency Feedwater Pump is Turbine Driven and this assures the heaviest loaded Diesel would be reviewed.
- 2. The " swing" 480V Engineered Safeguards Motor Control Center 3AB supplying power to the Reactor Building Cooler 3C will be aligned to the "B" 480V Engineered Safeguards Bus.
- 3. Actual Power Factors for the connected loads would be applied to more realistically model the system. l I
Per the above criteria a load table was constructed based on nameplate rating for the large motors and fans connected to Emergency Diesel Generator 3 A. The results of this approach indicated an automatic diesel loading of 3449KW with 372KW of potentially manual applied loads which exceeded the manufacturers rating of 2750KW continuous,3000KW for 2000 l
hours and 3300KW for 30 minutes.
To remove some of the conservatism inherent in the nameplate values actual brake horsepower (bhp), requirements of the pumps were calculated based on full flow conditions of the various safety systems. These values were then translated to motor load requirements using nameplate motor efficiencies and converted to Kilowatt based loads. The results of this approach indicated an automatic diesel loading of 3379KW with 372KW of potentially Gilbert Commonwealth I-1 l
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, ,e manual applied loads Again, due to the z inherent conservatism of this '
approach manufacturers ratings were exceeded.
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the final analysis the dieselloading for each of the foikowing Design Basis
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' -A'e cidents were evaluated 'to more accurately model the expected station emergency loadings.
- t. Loss of Coolant Accident (LOCA)
A. Large Break -
" 1 -- B. Intermediate Break C. Small Break
~ 2. .- ' Steam Line Break Accident (SLB)
A. Inside Containment
-l
. B. Outside Containment
- 3. ' Feedwater Line Break -Inside Containment
- 4. Steam Generator Tube Rupture j 1
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The above Design Basis Accidents were evaluated as these scenarios would challenge the EngineeredSafeguards Actuation system. Single failures considered were: i l
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- 1. Loss of The Emergency Diesel Generator 3B
. 2. Loss'of the turbine driven emergency feedwater pump EFP-2. "
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Gilberttommonwealth I-2 e.. -
The loss of EFP-2 was reviewed due to the sensitivity of the calculation'to increased flow requirements of the motor driven pump to assure the limiting case was assessed.
' System flow requirements were es+ablished for each pump for each of the above scenarios. These flow requirements were converted to pump bhp via test curves unique to 'the individual pumps or calculated based on system parameters and finally to motor horsepower and Kw loadings via the motor nameplate efficiencies.
To impr'ove long term dieselloadings and provide additional benefits to plant operation and reliability the following design enhancements were identified:
A.'- Inhibit start or provide tripping of -the. motor driven emergency
. feedwater pump in the event low pressure injection' flow is established coincident with a Loss Of Offsite Power (LOOP). This condition would-only occur for a large or intermediate break LOCA in which LPI flow has been established. Coincident with this modification, the present automatic shedding of the motor driven emergency feedwater pump after 30 minutes of operation in the 30 minute rating of the diesel generator will be removed.
B. Realign one of the alternate motor operated valves supplying steam to the turbine driven emergency feedwater pump from Battery 3A thus providing redundancy. Presently both valves are supplied from Battery 3B. Due to recent modifications to the emergency feedwater .
system control- this provides load sharing capability ' for the .two emergency feedwater pumps thus improving plant reliability while' reducing the load on the A diesel generator.
C. Provide automatic load shedding of the heat tracing from the Emergency Diesel Generator 3A loading. This load is non-safety I related and not required for accident mitigation. I l
Based on incorporation of these design changes the worst case diesel loading j for the above scenarlos is 2991RW.
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This results in the Emergency Diesel. Generator 3A' operating 'within its :
2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 3000KW.' Thereby assuring its availability for accident.
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- mitigation.
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- 1 in the following sections' specific discussions of the methodologies used in the '
mechanical' and electrical calculations to support these conclusions .are
. presented. . Also scenario descriptions, summary loadings, results .and -
detailed . descriptions of the design modifications Lwith theiri safety evaluations are provided.
1, 1
i Gdtwrt Commenmalth I-4
f II. MECHANICAL METHODOLOGY A. Aporoach i
The accident parameters are addressed in this report only for the purpose of defining the ES loading regnirements for the emergency diesel generator. No attempt has been made to justify the core cooling or protective function capability of the pump flows used to develop the equipment horsepower requirements for these scenario analyses.
Pump flows were used that conservatively enveloped the ac.aal- ,
requirements identified in the accident analyses during the various l time intervals.
Accident scenarios were developed to define the equipment power requirements for their specific operating modes during Design Basis Accidents. The existing accident analyses for the plant design basis !
events as defined in the FSAR were utilized where available and for those accidents that were updated beyond the original licensing basis, B&W topical report information was utilized. These existing analyses defined the containment, primary and secondary system parameters resulting from these postulated accidents and the sequence of the response of the engineered safeguards (ES) systems to mitigate the accidents. In some scenarios adjustments were made to account for the differences in the resulting parameters because of the assumed '
coincident occurrence of the loss of offsite power and the accident.
The scenario analyses were based on-reactor, reactor coolant pump, main feedwater pump, and turbine trip coincident with accident detection to maximize dieselloading.
The Design Basis Accidents covered by scenario analyses in this report are those that have been identified in the existing FSAR accident analyses as resulting in automatic actuation of the engineered safeguards actuation (ESAS) systems. Therefore the fellowing Gilbert Commonwealth 11- 1 ,
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L accident analyses for those events are included in this report as they I~ result in automatic actuation of the ESAS:
- 1. LOCA - Small, Intermediate, and Large Size Breaks l
Steam Line Break Accident 2.
- 3. Steam Generator Tube Failure Accident i
- 4. Feedwater Line Break Accident - Inside the Reactor Building . ~
l The following accidents do not result in automatic actuation of the ESAS thus requiring reduced dieselloading and are not presented in this report.
- 1. Fuel Handling Accident
- 2. Rod Ejection Accident
- 3. Waste Gas Tank Rupture Accident i
- 4. Loss of Feedwater
- 5. Feedwater Line Break Accident - Outside the Reactor Building G. Uncompensated Operating Reactivity Changes
- 7. Start-up accident
- 8. Rod Withdrawal Accident at Rated Power Operation
- 9. Moderator Dilution Accident
- 10. Cold Water Accident
- 11. Loss-of-Coolant Flow !
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- 12. Stuck-Out, Stuck-In, or Dropped Control Rod Accident l
- 13. Loss of Electric Power 1
- 14. Maximum flypothetical Accident (MIIA)
The scenario analyses presented in this report cover the major equipment which have variable operating modes and characteristics depending on the specific accident and the accident's resulting effect j
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r on- the primary,' secondary, and containment system parameters relative to pressure, temperature and flow. The following systems and-their major equipment are included:
- l. High Pressure injection (HPI)- MUP-1A and MUP-1B
- 2. Low Pressure Injection (LPI)- DHP-1 A and DHP-1B 4 3. Reactor Building Spray System (BS)- BSP-1 A & BSP-1B
' 4. ^ Emergency Feedwater (EFW)- EFP-1 and EFP-2
' Equipment, in the above systems and other systems which function but whose operating modes and , characteristics are independent of the accident and.its results, are the following:
- 1. Reactor Building Emergency Cooling Units
'2.. ' Decay Heat Sea Water Pumps
- 3. ' Decay Heat Closed Cycle Cooling Water Pumps
- 4. Miscellaneous valves
- 5. Inverters
'6. Battery Chargers
- 7. Control Complex and Emergency Lighting Additionally there is equipment whose operating characteristics are independent of the accident and its results but are dependent on the number of redundant components available (assumed single failure case) to operate within the common system. The following pumps are in systems which have different operating characteristics (flow and.
head) when there are two pumps available and operating in parallel and ,
sharing the total system flow than when there is only one pump i available to provide the design flow required:
' 1. Emergency Nuclear Services. Sea Water Pumps - RWP-2A and RW P-2B l
Gelbertcommonwealth 11- 3 l, ,
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- 2. Emergency Nuclear Services Closed Cycle Cooling Water Pumps -
SWP-1 A and SWP-1B
'These pumps have reduced power requirements when both are available because of the shared flow within the common system piping.
There are also systems that have . equipment that 'is not required immediately following the accident to mitigate the accidents severity but is required to operate at a later time to provide essential functions ' I within the plant. These equipment power loads are manually appl _ led to the diesel generator at a time which is dependent on the initial operating conditions of the systems or- at a ' procedurally identified time.
The following equipment is required to be. loaded onto the -diesel generator within 10 ' minutes:
- 1. Control Complex Emergency Duty Supply Fans - AHF-18A and AHF-18B
- 2. Control Complex Return Air Fans - AHF-19A and AHF-19B The following equipment would be manually loaded as required at the ;
discretion -of the operator at -a time when diesel ~ capacity is either available or can be made available by selective shedding of loads nonessential to the accident scenario:
- 1. Spent Fuel Coolant Pumps
- 2. Chilled Water Supply Pumps
- 3. Control Complex Water Chillers The scenario analyses cover the EDG operating time period until p discretionary action by the operator can be exercised. This was i h considered a sufficient time period in which the characteristics of the )
equipment could be defined for the mitigation period of an accident l where automatic equipment actuation would occur. Beyond this time f
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was considered the recovery period of the accident where the operator can use discretion to determine what equipment is necessary and take manual actuation to initiate or terminate same (procedural guidance will exist for load shedding).
The specific scenario for the Design Basis Accidents are included in the appendices to this report. The accident dependent flow requirements for pumps are discussed in detail in each scenario. The pump horsepower was determined using the pump vendor test performance curves or calculated based on system scenario parameters and are individually listed in Table 1 and Table 2. The pump horsepowers for the pumps which have characteristics dependent on the assumed single failure are also listed individually in the Tables. All the other loads that are independent of the accident scenarios have been summed together and are listed as " Remaining Loads" in the Tables. The horsepower for the equipment which is manually loaded within 10 minutes has been included as a separate listing.
For the purposes of defining the maximum loading for the diesel generator, two single failure cases were evaluated. The first was the
'B' EDG Failure Case. This case was based on the loss of the 'B' train of the 250/125 vde system with the resultant loss of the emergency diesel generator 3B, EDG-3B, due to loss of field flashing and control.
This results in the ES equipment required for accident mitigation to be loaded onto emergency diesel generator 3A, EDG-3A, including the motor driven emergency feedwater (EFW) pump, EFP-1. Since the redundant EFW pump, EPP-2 is turbine driven, EDG-3B does not include an equivalent load. The other ES equipment horsepower provide nearly identical load requirements for EDG-3 A and EDG-38.
Therefore, of the two diesel generators, EDG-3A has the greatest potential load requirement. The 'B' EDG Failure Case insures that the worst case will be reviewed.
l The second single failure case evaluated was the EFP-2 Failure Case.
As the first single failure case was the loss of EDG-38, the turbine driven EFW pump, EFP-2, was assumed to be available and pumping in Gilbert Commonweeth l I1-5 1
parallel with the motor driven pump EFP-1. As a result EFP-1 pump horsepower requirement is reduced due to shared load.Since EFP-1 is '
a.significant load on EDG-3A a second -single failure of the, turbine -
' driven' pump was evaluated. :This case maximized ths horsepower requirement of EFP-1. 'The EFP-2 Failure Case was base'd on both:
-EDG3A and 3B to be available with the result that both train A and B of. the ES equipment would be. operating. In the. evaluated scenario analyses this _ failure ' case affected the operation of the Emergency Nuclear. Services Sea . Water Pumps, RWP-2A and RWP-28, and the .j Emergency Nuclear Services Closed Cycle Cooling-Water Pumps (SWP-1A 'and SWP-1B).' These systems have two redundant pumps operating -l In parallel to supply water through a common piping system to the
~ components requiring cooling water. With the two pumps operating the Individual h'orsepower of ~each pump is reduced. In some scenarios the flow requirements of the ES pumps are reduced as the required flow is shared byl the two. In other scenarios, the ES pump flows are not -
affected due to the fact that the pumps operate independently and the:
flows are not combined in a load sharing operability mode.
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6 II.~ Mechanical Methodology 1: .
B. .: Scenarlo Assumptions in each scenario analyses, the Design Basis Accident (DBA) is assumed to be coincident with loss'of off-site power (LOOP). As a result of LOOP,- the reactor, the . reactor coolant pumps, the main feedwater pumps and the turbine are tripped. For some of the scenarios this assumption moves the initiation of these events to an earlier time in the accident sequence than
[ was originally assumed for the design basis events accident analyses. Wh'ere j, this sequence time shift is significant adjustments are made in the accident scenarios.
In the EFP-2 Failure Case, both the emergency diesel generators 3A and 3B, EDG-3A and 38, are assumed to be available and both the 'A' and 'B' trains of
'the ES equipment lare assumed operating. It is assumed that until operator action ~ls taken to shutdown any of the equipment on either EDG3-A or 3B
'both trains of equipment are operating. The EFP-2 Failure Case has a-significant' impact in the operating characteristics of the pumps in the Emergency Nuclear Services Sca Water System and the Emergency Nuclear Services Closed Cycle Cooling Water System. Each of these systems has common piping and it is assumed that each of the two pumps would share the i total system flow based on the system resistance curves generated for this mode of operation. .
l The Emergency Feedwater System is assumed to be-in operation for each of the scenarios. The following is a description of the control system and its )
assumed mode of operation during the scenarios: l EFIC CONTROLS Emergency Feedwater Initiation and Control (EFIC) is initiated by (1) loss of -
both main feedwater pumps, (2) loss of all RC pumps, (3) both HPI actuation 1 trains, (4) low Steam Generator Pressure, or, (5) low Steam Generator level. J Any of the initiating signals will cause actuation and control of Emergency l
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! Feedwater (EFW). Only low Steam Generator pressure causes Main Steam !
l Isolation and Main Feedwater Isolation.
The EFIC System is depicted in FSAR Fig. 7-26. is comprised of four initiating channels (A, B, C and D) which feed two actuation channels (A and B). Actuation Channel A being assigned to'the motor driven pump EFP-1 and a control valve to each Steam Generator. Actuation Channel B being !
- assigned to the turbine driven pump EFP-2 'and a control valve to each Steam Generator. Initiation Channel D initiates' signals to the two block valves from EFP-1 while Initiation Channel C initiates signals to the two block valves from EFP-2 whenever the EFIC logic determines the. particular
, conditions exists.
The EFIC 'A' and. 'B' Actuation channels also - automatically control at two predetermined OTSG setpoints;,1) thirty-six inch for -any RC pump running and 2) 65% of operating level if all RC pumps are off to insure natural recirculation capability. A fill rate is imposed by the control system-if the 65% setpoint-is chosen. The fill rate of between 2 and 8 inches per minute is a function of the Steam Generator pressure, allowing maximum fill rate at maximum pressure.
The power level at the time of the accident determines the Steam Generator inventory and consequently the time necessary to reach the fill setpoint. i ASV-204 and ASV-5 are parallel valves, powered by the 'B' power source through a single control switch and actuated "open" by EFIC 'B' Actuation
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Channel. The "open" condition of the valves admits steam to the turbine driven pump. The separating of ASV-204 from ASV-5 and providing ASV-204 with an 'A' power source with a EFIC 'A' actuation channel signal will provide redundant means of admitting steam for EFP-2. Hence a complete loss of the 'B' power source does not negate operation of the EFP-2.
For the scenarios of this report with a loss of 'B' battery and 'B' diesel, the EFIC 'A' Channel will open ASV-204 and allow full f. low from EFP-2 since l - EFIC 'B' control is lost. This occurs immediately upon loss of power while j l- the 'A' powered motor driven pump is being load sequenced after the 'A' l
Gdbert Common *ealth 11- 8 a __ = _ __. _ l
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l y l diesel is running. The valve opens and the turbine runs up to speed in approximately 15 seconds.
- The EFP-2 pump will continue to.run with'out 'B' channel' flow control until
- the Steam Generator Overfill setpoint is reached. At that time EFIC 'C' [
Initiation channel outputs a CLOSE signal to the motor driven block valves l for EFP-2. During this period of operation the EFIC 'A' channel will have controlled EFP-1 flow depending on fill rate and finally? placing the EFP-1 pump .into re-circulation upon reaching Steam Generator setpoint at 65E
-The EFIC 'C' Initiate signal will alternate initiate CLOSE-OPEN signals according to' the OVERFILL and OVERFIIL RESET setpoints if any form of. 1 Steam Generator boll-off is occurring. As the loss of 'B' battery did.not result in the worst case loading on EDG-3 A it was not analized further. 1
- For the 'B' EDG Failure Case EFP-1 and EFP-2 would be available and each ]
would share the flow requirements dictated by EFIC. j It is assumed 'that the EFW flow rate to remove the core decay heat by.
steaming in the once through steam generators (OTSG) is 470' gpm, unless .;
specific information was available from previous accident analyses. This represents the EFW flow required to remove the decay heat core power level of 3% at approximately 4 minutes after reactor shutdown by relieving .
- saturated steam at the safety valve setpoint of 1050 psig. j I
It is understood that after a period of time, once the 65% SG level is reached, the EFW flow will be reduced to match steaming rates of 470 gpm or less. However, to conservatively determine diesel loadings, the maximum i EFP-1 flow is assumed for the entire length of each scenario. ,
- The maximum fill rate excluding. steaming controlled by EFIC during the I recovery of the steam generator level is 330 gpm per steam generator. This corresponds to a fill rate of 8 inches per minute (the maximum rate at the maximum steam generator pressure.) The steam generator requires approximately 42 gallons to raise the level one inch. l The EFW pump re-circulation line will pass 200 gpm during all modes of l L pump operation based on actual flow test results. With two pump parallel L
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operation it is assumed the recirculation rate of 200 gpm will be shared by _
each pump.
To obtain conservative fic v estimates for the pumps the maximum flow values permitted by flow control devices in the system were assumed. The LPI pumps, DHP-1A and DHP-1B are limited to a maximum flow of 3150 gpm (Ref. FSAR Section 6.1.2.1.2). The R.B. Spray pumps, BSP-1A AND BSP-1B, are limited to a mnximum flow of 1600 gpm (Ref. FSAR Section 6.2.2.1).
There are no flow limiting devices in the Emergency Nuclear Services Closed Cycle Cooling Water System, the Emergency Nuclear Services Sea Water System, the Decay heat Sea Water System, and the Decay Heat Closed Cycle Cooling Water System. For the closed cycle cooling water systems the flows are assumed to be 10% greater than their design values. For the sea water systems the pumps are assumed to operate at flows which corresponds to the ;
maximum pump horsepower taken from the test curve. The HPI pumps, MUP-1 A and MUP-1B, are assumed to operate at a flow of 600 gpm which corresponds to the maximum horsepower taken from the pump test curve unless other flow conditior: were noted in the scenarios It is assumed appropriate operator action may be taken after 10 minutes following the initiation of ES if a procedure exists for such an action and after 30 minutes if no procedure is in place.
Gelbert Commonwealth 11- 1 0
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!!!. ELECTRICAL METHODOLOGY The following defines the methodology used in identifying the diesel generator loads and electrical load values used in determining the overall diesel generator running load and voltage dip values.
A. ASSUMPTIONS l
- 1. Losses in electrical power' cables are considered to have no significant impact on the overall diesel' loading, and therefore are neglected. This assumption is based on the results of analyzing 4 the worst case 4kV and 480V loads and determining worst case cable losses of 0.5kW and 2.OkW respectively. Due to the limited number of loads at the 4kV level (8) and 480V level (1), other than motor control center loads which are larger size cables and
-produce much smaller cable losses, this assumption is considered valid.
- 2. The power factor of non-motor loads is conservatively selected as unity. The non-motor loads include emergency lighting, battery i; chargers, inverters, and miscellaneous power distribution panels. j
- 3. A power factor of 0.5 is assumed for the motor operated valves.
A review of sample motor data sheets for valve motor operators )
indicated power factors between 0.4 and 0.6. Therefore 0.5 is I considered appropriate and is applied to the total kVA of the valve motor operators.
- 4. Motor full load efficiencies were used in calculating the kW input l
for the large horsepower pumps, as in most scenarios the motors j are operating at or near full load. In the scenarios where the l Makeup and Purification Pump, Decay Heat Pump and/or the I
Emergency Feedwater Pump will be operating in the recirculation mode, reduction in efficiency is considered negligible. This ma ean,n.m 111 - 1
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assumption is based on review of typical squirrel cage induction !
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l=, motor efficiency versus load curves. ,
- 5. Small motor loads (less than 100 hp) are considered to operate at -
nameplate horsepower ratings. !
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B. APPROACH ,
The first step in performing the emergency diesel generator 3A 1
(EDG-3A) loading summary analysis was to identify the loads that are energized for emergency diesel generator operation during a loss-of-offsite-power condition coincident with an ES actuation. The I total diesel loading includes (1) the loads not tripped on an undervoltage signal and subsequently re-energized as block 1 loads ;
when the diesel comes on line, (2) loads that are tripped on an j undervoltage signal and automatically re-energized in sequence as blocks 2, 3, 4 and 5, and (3) manual loads that by procedure must be i manually connected prior to the operator taking administrative action ,
to shed any unnecessary loads.
The individualloads were analyzed to determine the amount of kW and kVA cach load contributed to the diesel loading. The kW values were used to determine the overall diesel generator running load as the diesel engine rating is based on kW. The kVA values were used to determine the voltage dip and recovery time experienced by each of the block loads.
In determining the kW and kVA values of the loads, the loads were r grouped into four basic categories:
Motor loads of 100 horsepower or larger. l Motor loads less than 100 horsepower. '
Non-Motor loads.
Motor Operated Valves.
Gdbert Commonwealth III-2
- 1. Motor loads of 100 hp or larger: l These are pump motors, which comprise the major portion of the diesel load and they have been analyzed for 4 specific loading cases.
1 Case 1 - Motor kW based on nameplate horsepower rating, i
i Case 2 - Motor kW based on full flow brake horsepower. ;
Case 3 - Motor kW based on the pump flow rate required for the accident scenario resulting in the worst case diesel loading with a single failure of the "B" channel diesel i
generator..
Case 4 - Motor kW based on the pump flow rate required for the accident scenario resulting in the worst caso diesel !
loading with a single failure of the turbine driven j emergency feedwater pump. I d
For each of the above cases the motor kW was determined as follows:
Motor kW = Pump Brake Horsepower X 0.74G Motor Full Load Efficiency The motor running kVA and starting kVA for each motor was calculated using the motor rated volts and motor full load amps and locked rotor amps. The motor running and starting kVA's were only calculated for Case 1 as this case resulted in the most conservative loadings on the diesel generator and thus the worst case voltage dips and recovery times.
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- 2. Motor loads less than 100 hp:
For motor loads of less than 100 horsepower the motor kW was determined from the motor nameplate ratings and full load efficiencies. These values of kW were then utilized in each of the four cases of the diesel loading analysis. The motor running and starting kVA's were calculated in the same manner as for the large horsepower motors of Item 1 above.
- 3. Non-Motor Loads:
Non-Motor loads include the battery chargers, inverters, emergency lighting, and miscellaneous motor control center and distribution panel loads. The load kVA on the battery chargers and inverters was determined by summing the individual loads on respective 250/125VDC and 120VAC vital distribution panels.
The required AC input to the chargers and inverters was determined by calculating the efficiency of each and applying it to the total charger and Inverter loads. Other ncn-motor loads were determined by summing the individual loads based on review of the applicable distribution panel drawings and vendor documents.
- 4. Motor Operated Valves:
The total load of ;he valve motor operators was determined by summing the fullload amps (FLA) and locked rotor amps (LRA) of the valves that are actuated under an ES condition. The kW load l produced by the valve motor operators was determined by applying a 0.5 power factor to the total valve FLA. The running and starting kVA was calculated using rated volts and the total valve FLA and LR A values.
After the load kW values were established in Items 1 thru 4 above, the loads in each of the 5 loading blocks were totaled, and the total diesel loading was determined by summing the 5 block load totals. This was Gdbort CommonweMth III-4
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performed for each of the four loading Cases (Ref. Item 1 above).
From this total load the diesel running load was determined by subtracting the momentary loads. Momentary loads include valve motor operators and lube oil pumps that serve only as a backup to a mechanical oil pump and are automatically tripped by an oil pressure signal. These momentary loads are included in the block loading because they must be considered in the voltage dip and recovery time analysis, but since they are energized for only a short duration they are deleted from the total diesel kW load. Finally, the loss through the 4.16kV/480V transformer is added to the t'otal diesel load to determine the total kW running load that is automatically sequenced'onto diesel generator EDG-1 A.
Manually Connected Loads:
Two of the manually connected loads (Control Complex Emergency !
Duty Supply Fans and the Control Complex Return Air Fan) are committed to be connected to the diesel generator within 10 minutes of the loss-of-offsite-power condition due to control room habitability concerns. Therefore, these particular fan loads are added to the total diesel sequenced load. The other manually applied loads are not included in the diesel loading summary as these loads are not considered necessary until such time when the operator can shed other loads which are no longer required.
Voltage Dio and Recovery Time Analysis:
The voltage dip values and recovery times resulting from the block loading were obtained from load / voltage curves received from the diesel manufacturer. Voltage dip was porportional' to the summation the starting kVA of each load block while the voltage recovery time was proportional to the summation of the respective starting and cumulative running kVA of each load block. Voltage dips were also calculated per the following equation supplied by the diesel manufacturer as a design check:
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% Voltage Dip = 100 / [1 + GRKVA
- 100 / (SKVA
- Cx"d)]-
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'Where GRKVA: Generator Rated kVA i 1
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.. . -t SKVA . = Inrush Reactive Power
]
= Locked Rotor Amps *, Volts
- 1.73/100 I
- \
Cx"d '=Corrected Sub-Translent Reactance .l
= (x'd - x"d)
- 2/3 + x"d where L X'd' ' = Transient Reactance of generator in percent' c < X"d = Sub-Transient Reactance of generator in percent .
Results in'each case were comparable, l i
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l- . From th'e KW total loads depicted on Tables 1 and 2, it is concluded that with
- theLimplenientation Af the design changes described in = Appendix B of this
>ftI' , _ - . y, reports)hhe total diesel generator' KW running load is. below .the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> .
4
.+
ff
, maximum rating of 3000KW for each of the postulated accident scenarios.
j T1 . [ .
d From Table 3 it is concluded that the voltage dips produced by.each of the
- block loads are accepta'ble' levels, and will not cause de-energization"of any - !
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' safety related electrical components. !
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- j. , TABLE 3 r:
BLOCK LOAD VOLTAGE DIP
SUMMARY
L LOAD STAltTING % VOLT. % VOLT.
'y BLOCK kVA FitOM CUltVE CALCULATED j- 1 7436 68.1 73.8 2 4325 80.1 82.9 3 4329 80.0 82.8 4 4035 81.5 83.8 5 3914 81.9 84.2 Above voltage dip values are on 4160V Base.
APPENDIX A - SCENARIOS A.1 LOSS of COOLANT ACCIDENTS 1.0 GENERAL The loss of coolant accidents considered are discussed in the FSAR Section 14.2.2.5. The worst case loading of the emergency diesel generators t>
l was established considering ESF systems response for large line break LOCAs (14.14 ft2 to .5 ft 2), intermediate line break LOCAs(.5 ft2 to .01 ft 2) and smallline break LOCAs(less than .01 ft2).
The consequences of a LOCA are mitigated by the actuation of the ESAS which monitors reactor coolant pressure and reactor building pressure.
2.0 LARGE BLOCK LOCA The LOCA analysis (Reference BAW-10103A, Rev. 3) indicates that for a large break LOCA, the HPI/LPI actuation setpoint of 1500 psig in the RCS and the reactor building spray setpoint of 30 psig in the reactor building are reached in the 10 second time delay required for the diesel to start and be connected to the busses.
Therefore, at the time the emergency diesel generator is ready to be loaded, ESAS actuation signals on low RC pressure of <1500 psig or <500 psig and high RB pressure of >d p::!g or >30 psig may be assumed to be present to start the IIP!, LPI, and BS pumps. The reactor building cooling fans are automatically transferred to ES operation.
The LPI to EFP-1 interlock will prevent starting of the EFP-1 for the large break LOCAs. Since the primary system voids and decouples from the secondary system,EFW flow is not required for accident mitigation.
t Gelbert Commonwedith A.1-1 l
m ..
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i 2.1 "B" EDG Failure Case r
. ~ . .
,j
. Postulating' the failure of:the ; emergency diesel generator 3B leaves the '
emergency diesel generator 3A to' provide- AC power to the 'ESF features ]
r'equired to mitigate the consequences of a large LOCA.
The HPI pump (MUP-1 A) is actuated'by.the Engineered Safeguard Actuation' '
System (ESAS) as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The HPI pump will operate at its maximum flow of-600 gpm and continue to operate until operato action is taken to terminate :
it.
l The LPI pump (DHP-1A) is also initiated by ESAS. Since the RCS pressure -
decreases rapidly below its maximum discharge pressure, the LPI pump will p operate at a flow of 3250 gpm,.which includes 100 gpm pump recirculation 'l flow,'and continue at.this flow throughout this analysis.
RB Spray will be initiated by the 30 psig RB setpoint. The RB Spray pump (DSP-1A) will be operating at a flow of 1600 gpm and will continue at this i
flow until operator action is taken to terminate it. j
>1 Emergency Feedwater .(EFW) will be initiated -on loss of reactor coolant pumps or ESAS and the turbine driven pump (EFP-2) will be operating,. if i steam is available, and controlled'by EFIC. EFP will be prevented from j
, starting because of the LPI flow logic since LPI' flow is established almost l immediately in this analysis. Since the LPI to EFP-1 interlock, will be j actuated the loading will be reduced..
- 2.2 EFP-2 Failure Case Postulating the loss of the Turbine Driven pump, (EFP-2) allows the
' assumption that two emergency diesels generators are available. The LPI to j EFP-1 interlock reduces the load on EDG-3 A. 1 Both HPI pumps (MUP-1A and MUP-1B) are actuated by ESAS as a result of i 1
RCS Pressure less than 1500 psig or RB pressure greater than 4 psig. The l canc .m.n..m
)
A.1-2
.5-HP1 pumps will operate at their maximum flow of 600 gpm.and continue to operate until operator action is taken to terminate them.
The LPI pumps (DHP-1A and DHP-18)'are also initiated by ESAS. Since the RCS pressure decreases rapidly below their maximum discharge pressure, the LPI pumps will operate at a flow of 3250 gpm, which includes 100 gpm recirculation flow. They both continue at this flow throughout this analysis.
RB Spray will be initiated by the 30 psig RB setpoint. -The RB Spray pumps (BSP-1A and BSP-1B) will be operating at a flow of 1600 gpm and continue until operator action is taken to terminate them.
EFW-will be initiated on loss of reactor coolant pumps or ESAS. Because of the LPI flow logic, EFP-1 will be prevented from starting since LPI flow is established almost immediately in this analysis. EFP-2 was assumed to fall to start in this analysis.
3.0 INTERMEDIATE BREAK As the break size decreases to the smaller break sizes (i.e from .5 ft2 to .01 ft 2), the ESAS actuation signals are delayed. Since a RC pressure of less than 1500 psig is generated at the beginning of.the event and LPI flow will be !
delayed, EFP-1 and EFP-2 will be started by EFIC. EFP-1 will be tripped later upon LPI flow initiation.
Because the ESAS loading sequencers which control the HPI and the LPI pumps are actuated by either a low 1500 psi RC pressure or a RB pressure of 4 psig, these pumps will always be the first loads to be started. However, the LPI pump operates on recirculation until the RC pressure is reduced to approximately 185 psig where it can deliver 1000 gpm. To maximize diesel loading, the LPI flow was conservatively assumed to be 3250 gpm for the entire event.
For intermediate break LOCAs, the RCS depressurization can be extended, resulting in diesel loadings which are indicative of small and large break LOCAs. During the initial depressurization, following ESAS actuation by Gdbert. Commonwealth A.1-3
\;
i RCS pressure less than 1500 psig, the primary and secondary systems remain coupled and EFW flow is required to maintain OTSG levels. For this time period, the diesel loading is identical to the small break LOCA. Once the RCS pressure drops below 185 psi, LPI flow to the RCS is realized, EFP-1 is tripped and the diesel loading becomes identical to the large break LOCA loading. The LPI flow chosen to trip the EFP-1 ensures the EFP-1 load is shed prior to a dramatic increase in LPI pump loads. The scenario descriptions for the "B" EDG failure and the EFP-2 failure cases describe the resulting end state for the event, with the HPI, LPI pumps running with high flows.
The intermediate break cases also result in extended times to RB spray flow initiation. Since the timing varies with break size, it is conservatively assumed the BS pump is at full flow for the entire event.
3.1 "B" EDG Failure Case Postulating the failure of the emergency diesel generator 3B leave the emergency diesel generator 3A to provide AC power to the ES features required to mitigate the consequences of an intermediate break LOCA.
The HPI pump (MUP-1 A) is actuated on ESAS as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The HPI pump will operate at its maximum flow of 600 gpm and continue to operate until operator action is taken to terminate it.
The LPI pump (DHP-1 A) is also initiated by ESAS. Since the RCS pressure does not decrease immediately below DilP-1 A's maximum discharge pressure, the pump will initially be at recirculation flow of 100 gpm and then increase the flow as the RCS pressure decreases until a flow of 3250 gpm is attained. It will continue at this flow throughout the analysis.
RB Spray will be initiated by the 30 psig RB setpoint. The RB Spray pump (BSP-1A) will be operating at a flow of 1600 gpm and will continue at this flow until operator action is taken to terminate it.
GilberrCommonwealth A.1-4
\ - _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
EFW will be initiated on loss of reactor coolant pumps or ESAS. Both EFW pumps (EFP-1 and EFP-2) will start initially and be controlled by EFIC to share the flow requirements. Once LPI flow begins EFP-1 will be tripped by the' LPI flow logie. EFP-2 will continue to operate based on steam availability.
3.2 EFP-2 Failure Case h
)
Postulating the loss of the Turbine Driven pump (EFP-2) allows the assumptions that two emergency diesel generators are available. The LPI flow may be delayed resulting in the operation of the EFP-2 to fill the OTSG's.
Both IIPI pumps (MUP-1A and MUP-18) are actuated by ESAS as a result of _
RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The IIP 1 pumps will operate at their maximum flow of 600 gpm and continue to operate until operator action is taken to terminate them.
The LPI pumps (DHP-1A and DHP-1B) are also initiated by ESAS. Since the RCS pressure does not decrease immediately below the pumps maximum discharge pressure, the pump will initially be at recirculation flow of 100 gpm and then increase their flow as the RCS Pressure decreases until a flow of 3250 gpm per pump is attained. They will continue at this flow throughout this analysis.
RB Spray will be initiated by the 30 psig RB setpoint. Both BS pumps will be operating at A flow of 1600 gpm and continue until operator action is taken to terminate them.
EFW will be initiated on loss of reactor coolant pumps or ESAS. EFP-1 will start initially and be controlled by EFIC to fill the OTSGs and provide the water source for OTSG steaming. Once LPI flow begins EFP-1 will be tripped by the LPI flow logic. EFP-2 is assumed to fall to start in this analysis.
Gdbert' Commonwealth A.1-5 l
4.1- Small Break LOCA j i
Small breaks of less than .01 ft2 results in the actuation of HPI without a l reduction of RC pressure sufficient for LPI flow to the RCS to be initiated.
The small break LOCA coincident with a loss of offsite power starts with the initiation of the ESAS loading sequence. This results in the start of the HPI and LPI pumps. The reactor building cooling fans are automatically transferred to ES operation. The LPI remains in the recirculation mode until tripped by the operator. The EFIC initiates the start of the EFP-1 and EFP-2. The BS pump is not initiated since the RB pressure does not reach 30 psig. This is ensured by two RB fan coolers operating within 10 minutes of the accident.
4.1 "B" EDG Failure Case Postulating the failure of the diesel generator 3B leaves the emergency j diesel' generator 3A to provide AC power to the ES features required to mitigate the consequences of a small break LOCA.
The HPI pump (MUP-1 A) is actuated on ESAS as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The HPI pump will maintain a pressure of approximately 1250 psig for a break size of .01 ft2 based on a B&W analysis (Ref. B&W Document No. 51-1158-449-00). The HPI pump was assumed to operate at its maximum flow of 600 gpm throughout this analysis. This is a conservative assumption since this maximum flow is not reached untillower RCS pressures are attained.
The LPI pump (DHP-1 A) is also initiated by ESAS. Since the RCS pressure is at 1250 psig for this analysis scenario, the RCS pressure is greater than the maximum discharge pressure of the LPI pump. The LPI pump operates on recirculation at a flow of 100 gpm throughout this analysis.
I f The RB pressure does not reach the 30 psig setpoint for this scenario analysis based on the evaluation performed by B&W (Ref. B&W Document No. 51-t A.1-6
-___--_-----__j
I i 7
1158-449-00). The RB Spray pump will therefore not be operating for this analysis. 1 I
EFW will be initiated on loss of reactor coolant pumps or ESAS. EFIC will control the EFW flow from EFP-1 and EFP-2 to fill the OTSG and provide the water source for OTSG steaming for core decay heat removal. ' This has been determined by B&W analysis to require a flow from both EFP-1 and j EFP-2 of approximately.1100 gpm which includes 200 gpm pump l recirculation flow (Ref. B&W Document No. 51-1158-449-00). EFP-1 is -l assumed to contribute approximately 550 gpm of this flow requirement throughout this analysis.
4.2 EFP-2 Failure Case Postulating the loss of the Turbine Driven Pump (EFP-2) allows the assumption that two emergency diesel generators are available. The LPI flow trip will not be actuated resulting in the operation of the EFP-1 to fill the SG.
Two different sizes of small break LOCA scenario analyses were performed for this failure case. The first was for the .01 ft2 size break which was analyzed for the 'B' EDG Failure Case. The other was for a .02 ft2 break size. The first represents a 500 gpm RCS leak. The second represents a 1000 gpm RCS leak which is approximately the maximum makeup capacity .
of the two HPI pumps operating in this failure case. f i
.l-
.01 f t2 Break Size l Both HPI pumps (MUP-1A and MUP-1B) will be actuated on ESAS as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The j two HPI pumps will maintain a pressure greater than 1250 psig for a break ,
size of .01 f t2 based on a B&W analysis (Ref. B&W Document No. l 51-1158-449-00). The HPI pumps were conservatively assumed to operate at m .ncommoa...in l A.1-7
1 i
a flow of.600 gpm based on this RCS pressure. This flow will continue until- l operator action is taken to depressurize the RCS.
l The LPI pumps (DHP-1 A and DHP-1B) are also initiated on ESAS. Since the RCS pressure is approximately 1250 psig for this analysis scenario, the RCS pressure is greater than the maximum discharge pressure of the LPI pumps. !
The LPI pumps operate on recirculation at a flow of 100 gpm throughout this analysis.
The RB pressure does not reach the 30 psig setpoint for this analysis scenario based on the evaluation performed by the B&W. (Ref. B&W Document No. 51-1158-449-00) The RB Spray pumps (BSP-1 A and BSP-18) will not be operating for this analysis. l EFW will be initiated on loss of reactor coolant pumps or ESAS. EFIC will control the EFW flow from EFP-1 to fill the OTSGs and provide the water '
source for OTSG steaming for core decay heat removal. This has been determined by B&W analysis (Ref. B&W Document No. 51-1158-449-00) to require an EFW flow of approximately 1100 gpm which includes 200 gpm pump recirculation flow. EFP-1 (the only pump available for this failure case) will operate at a maximum flow of 1070 gpm based on system hydraulle.
analysis (performed by G/C) with the OTSG pressure at its relief valve set pressure. Since the pump capacity does not meet the EFIC control demand, it is assumed the OTSG fill rate will be reduced below the maximum and the refill period will be extended. ,
.02 ft2 Break Size Both HP! pumps (MUP-1A and MUP-18) will be actuated on ESAS as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. The two HPI pumps will maintain a pressure greater than 1250 psig for a break size of .02 f t2 based on a B&W analysis (Ref. B& W Document No. 51-1158-449-00). The HPI pumps were conservatively assumed to i
GeNrt Commonwealth
! A.1-8 '
1.1 + )
'l
- operate at'.a flow of 600'gpm based on this.RCS pressure. This flow will continue until operator action is taken to depressurize.the RCS.'
The LPI pumps (DIIP-1 A and DHP-1B) are also initiated on ESAS. Since the
'.RCS pressu're is approximately 1250 psig for this analysis' scenario, the RCS pressure is greater than the maximum discharge pressure of the LPI pumps.
The LPI pumps operate on recirculation at a flow of 100 gpm throughout this analysis.
The RB pressure doe's not reach the 30 psig setpoint for this analysis scenario.
based on the evaluation performed by the B&W. (Ref. B&W Document No.
51-1158-449-00) The . RB Spray pumps (BSP-1 A and BSP-1B)~ will not be operating for this analysis.
' EFW will be initiated on loss of reactor coolant pumps or ESAS. EFIC will control the EFW flow from EFP-1 to fill the OTSGs and provide the water
-source for OTSG -steaming for core decay heat removal.. This has been
.'determin'ed by B&W analysis (Ref. B&W Document' No. 51-1158-449-00) : to require an EFW flow of approximately 1100 gpm which includes 200 gpm pump recirculation' flow. EFP-1 (the only pump available for this failure case) will operate at a maximum flow of 1070 gpm based on system hydraulle analysis (performed by G/C) with the OTSG pressure ~ at its relief valve set' pressure. Since the pump' capacity does not meet the EFIC control demand, it is assumed the OTSG fill rate will be reduced below the maximum and the refill period will be extended. -
L Gilbert Commonweth A.1-9
= _ _ _ _ = _ _ - _ _ _ _ _ - - _- . - . . _ - _ _ _ . __ - - __ _ _ __
LARGE BREAK LOCA
'B'EDG Failure Case 600 HPI Pump Flow F
1 3250 LPI Pump Flow m O
W R
a R. B. Spray Pump Flow ,
1600 t
e S
' I (gpm) 0 >
EP-2 Failure Case Same as above for HPI, LPI, RB Spray, EFP-1 5 10 15 20 25 30 35 40 Time ,
GiftMert Commonwealth A.1 10
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p . . .o INTERMEDIATE BREAK LOCA -) ,
l:: . 'B'EDO Failure Case - ': .
- 4
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.l 600 > i
{
LPI Pump Flow 1 3250 i
- i 100' i
- . s 1600' >
F a
- 1. ' Shared Flow 0 EFP-1 Flow .
i,
-a st- J
-e S~
I@S) - EFP-2 Failure Case Same as above for HPI, LPI, RD Spray . I Full Flow- j i EFP-1 .
10 20 ~30 40 i Time Time (Minutes)
I i
Gilbert Commonwealth A.1- 11 ci_ _ _ _ _ _ _ _ _ ___.. _ _ _ _ _ _ - - - - .__ -l
t I i
.,[ f 5 J \ ^
p ,.
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- i .
SMALL BREAK LOCA
'B'EDG Failure Case -
ri.
600; HPI Pump Flow .
I 100 LPI Pump Rectre. Flow
' +
RB Spray Pump Flow 0
1;
. o:
- w .
. 550 EFP-1 Flow .
-R )
-a
- t<
e-
.S EFP-2 Failure Case. .
Same as above for HPI, LPI,RB Spray 1070 EFP1 ,
.J i
l
.l i
l I
l
,5 'l 10 15 20 25 30 35 40 Time (Minutes) 1
-)
Gilbert Commonwealth A.1-12
_____________q
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k
-y _l 1
APPENDIX A - SCENARIOS A.2> STE AM LINE BREAR ACCIDENT (SLBA).
The basic accident dynamics for the steam line break accident (SLBA) are j discussed !n FSAR Section 13.4.2.2.1. Since loss of off. site power is assumed 1
.i concurrent with' the SLBA,' the turbine, reactor, and : main feedwater pumps 1 are assumed to be tripped at time zero of this accide'nt. .Because of the response of engineered safeguards features to breaks inside the' Reactor --
Building (RB), two evaluations were performed: One for breaks inside the RB and one for breaks outside the RB. The scenarios for these analyses.are presented below for the two assumed failure cases. (failure of the .i emergency diesel ' generator .38 to start and failure of the turbine driven .
emergency feedwater pump to start.) i 1
A.. Inside the RB ' ,
- 1. 'B' EDG Failure This scenario is similar to Case I described in the FSAR section l 14.2.2.1.3 in that the main feedwater block valves or the main feedwater pump suction valves will be closed. However, the !
FSAR analysis assumes continued operation of -the main feedwater pumps and does not reflect the c' eduction in feedwater flow due to pump trip and coast down time, with resulting reduction-in energy input into the RB consistent with the assumption.of loss of offsite power concurrent with the SLBA. I Therefore, the RB pressure resulting from the SLBA in this 'B' EDG failure case scenario was determined to be below 28 psig .
(Reference, B&W Document No. 51-1158-449-00).
The HPI pump (MUP-1 A) is actuated by ' the Engineere'd Safeguards Actuation System (ESAS) as a result of RCS pressure less than 1500 psig or RB pressure greater than 4 psig. After the initial transient to fill the RCS due to shrinkage, which requires an HPI pump flow of 600 gpm, the HPl pump will operate at a GdbertCOmmonwedith
'A.2-1
flow of 295 gpm. This flow was determined by the pumps head capacity curve for a discharge pressure equal to the RCS safety.
relief valve pressure set. point. The flow will continue until operator action is taken to depressurize the RCS.
The LPI pump (DHP-1 A) is also initiated ESAS. Since the RCS pressure does not fall below the pressure for when LPI flow to the RCS is initiated (Ref. B&W Document No. 51-1158-449-00). The LPI pump will be on recirculation at a flow of 100 gpm.
RB Spray will not be initiated because the RB pressure will not reach the 30 psig set point. The maximum pressure based on a B&W evaluation for this case scenario is below 28 psig (Ref. B&W Document No. 51-1158-449-00) and therefore, the RB Spray pump (BSP-1A) will not be operating.
Emergency Feedwater (EFW) will be initiated on loss of reactor coolant pumps, low steam generator (OTSG) pressure / level, or ESAS. EFIC will. initiate isolation of both OTSG's and control EFW flow to the unaffected OTSG at the maximum fill rate of 8" per minute (330 gpm). Since steaming is occurring at the same time, the flow to the unaffected OTSG would be 470 gpm. The ,
EFW pump flow from both EFP-1 & EFP-2 is 1070 gpm which includes the 200 gpm recirculation flow. EFP-1 is assumed to contribute approximately 500 gpm flow during the period to refill the OTSG and then 335 gpm thereafter to maintain levels, match steaming rates and a recirculation rate of 100 gpm.
- 2. EFP-2 Failure Case This scenario is similar to Case I described in the FSAR section 14.2.2.1.3 in that the main feedwater block valves or the main feedwater pump suction valves will be closed. However, the FSAR analysis assumes continued operation of the main feedwater pumps and does not reflect the reduction in feedwater flow due to pump trip and coast down time with resulting ca.non,moman A.2-2 L___ ___-.
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reduction in . energy input into the RB consistent with the
- assumption of loss'of offsite power concurrent with the SLBA.- 1
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. Therefore '.the '. RB pressure 'resulting from the SLBA in this 'B'~
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EDG failure' case scenario was determined to be below 28 psig (Ref. B&W Document No. 51-1158-449-00).
The HPI pump -(MUP-l A) is actuated by the Engineered Safeguards Actuation System (ESAS) as a result of RCS pressure l
- less than 1500 psig or RB pressure greater than 4 psig. After the l initial transient to fill the RCS due to' shrinkage, which requires an HPI pump flow of 600 gpm, the HPI pump will operate at a flow of 295 gpm. This flow:was determined by the pumps head I
capacity curve for a discharge pressure equal to the RCS safety. .;
. relief valve pressure set point. The flow .will continue until !
operator action is taken to depressurize the RCS.
The LPI pump (DHP-1A) is also initiated ESAS. - Since the RCS l
.i pressure'does not fall below the pressure for when LPI flow to the RCS'is initiated (Ref. B&W Document No. 51-1158-449-00). The
- LPI pump will be on recirculation at a flow of 100 gpm.
RB Spray will not be initiated because the RB pressure will not.
reach the 30 psig. set point. .The maximum pressure based on a B&W evaluation for this case scenario is below 28 psig (Ref. B&W Document No. 51-1158-449-00) and therefore, the RB Spray pump '
(BSP-1 A) will not be operating. '
Emergency Feedwater (EFW) is initiated on loss of reactor l coolant pumps, low steam generator (OTSG) pressure / level, or ESAS. EFIC will initiate isolation of both OTSG's and control f EFW flow to the unaffected OTSG at the maximum fill rate of l 8" per minute (330 gpm). Since steaming and EFW recirculation are occurring at the same time, an additional flow of 670 gpm is j required to maintain the maximum fill rate and recirculation requirements. However the maximum fill rate to an OTSG (for a single OTSG available scenario) is controlled by EFIC at 740 gpm.
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Therefore, the total flow from EFP-1 (the only pump available in this failure case) will be 940' gpm including the 200 gpm recirculation flow. EFP-1 would continue at this flow until the OTSG is refilled.
1 i
B. Outside the RB i
1.. 'B' EDG Failure Case l 1
This case analysis is similar to Case III described in the FSAR. l Section 14.2.2.1.3 in that the steam line break is located outside j the RB is isolable and does not affect RB pressure.
The HPI pump is actuated by the ESAS as a result of RCS pressure .less than 1500 psig. The operation of the HPI pump in .
this case is similar to the SLBA 'B' EDG failure case scenario for f inside the RB. The operation of the LPI pump in this case is similar to the SLBA 'B' EDG failure case scenario for inside the RB. EFW will be initiated on loss of reactor coolant pumps, low j OTSG pressure / level or ESAS. EFIC will initiate isolation of both i OTSG's and control flow to both OTSG's at the maximum fill rate of 8" per minute (330 ~gpm to each OTSG). Since steaming is occurring at the same time, an additional EFW flow of 470 gpm is .
1 required to maintain the maximum fill rate control setting. The j total flow from both EFP-1 and EFP-2 is 1330 gpm which includes 200 gpm pump recirculation flow. E F P-1 is conservatively ,
assumed to supply 665 gpm flow during the period to refill the OTSG.
Since RB pressure is unaffected by this scenario, RB Spray pump, 1
BSP-1A is not operating.
- 2. EFP-2 Failure Case 1
I This case is similar to Case III described in the FSAR and is identical in its transient parameters to the 'B' EDG failure case GilbertCommonwealth
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since the- failure of EFP-2 does result in' different events. or
, responses as presented in.the FSAR.
- The -HPI pump is actuated by the ESAS as a result of RCS' pressure 'less than 1500 psig. The operation of the HPI pump in this' case is similar to th'e SLBA 'B' EDG failure case; scenario-for I
-inside the RB. The operation of the LPI pump.in this case is .
similar to the SLBA 'B' EDG failure' case scenario for inside the RB. . EFW will be initiated on loss of reactor coolant pumps, low OTSG pressure / level or ESAS. EFIC will initiate isolation of both' OTSG's and control flow to both OTSG's at the maximum rate of
~ 8" per minute'(330. gpm to each OTSG).- Since steaming .is'.
occurring at th6 same time, an additional EFW flow of 470 gpm is required.to maintain the maximum fill rate control setting. The total. flow from EFP-1 (the only pump available in this failure case) which includes 200. gpm pump -recirculation flow is
'1070;gpm since this is its maximum flow at . OTSG pressure corresponding to the relief valve set pressure. .Since the. pump capacity does 'not meet the EFIC control demand, it is assumed ethe OTSG fill rate will be reduced and the refill period will be extended.
Since- R.B. pressure is unaffected by this scenario, R.B. Spray pumps, BSP-1A and BSP-18 are not operating.
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APPENDIX A - SCENARIOS A.3 FEEDWATER LINE BREAK ACCIDENT (FLBA)INSIDE THE RB A feedwater line break results in a loss of the primary heat sink, primary system l heat up, increased pressurizer level and pressure, and reactor trip on high RCS pressure. For a FLBA inside the RB, ES is actuated by the 4 psig high RB pressure setpoint. For a FLBA outside the RB there is no ES actuation as the ES actuation setpoints are not reached. FSAR Section 14.2.2.9.3 describes the worst case for feedwater line break since EFW flow is assumed not to be initiated until 15 minutes after the break. This analysis and Table 14-64 are used as the basis for identifying the operating modes of the systems required to mitigate the accident and the sequence of events. The scenarios for this analysis are presented below for the two assumed failure cases. 4-
- 1. 'B' EDG Failure Case This analysis is different from the FSAR analysis in that the EFIC logic will isolate the feedwater earlier and the feedwater pump trip and coastdown
~
time is reduced due to the assumption of ..a 'I/O O P - coincident with the FLBA. This results in a reduction in energy input in'o the RB. Therefore the RB pressure resulting from the FLBA in this scenario was determined to be -
below 28 psig. (Ref. B&W Document No. 51-1158-449-00)
ES actuation is initiated on the 4 psig RB high pressure setpoint. HPI pump MUP-1A is initiated by this ES actuation and continues to operate throughout the accident. The pump will operate at a flow of 295 gpm. This flow was determined by the pump head-capacity curve for a discharge pressure equal to the RCS safety relief valve pressure setpoint. This flow will continue until operator action is taken to depressurize the RCS.
LPI is initiated on ES actuation. Since the RCS pressure remains at the pressure of the safety relief valve setpoint throughout this scenario, the RCS pressure is maintained at a greater pressure than the maximum discharge pressure of the LPI pump, DHP-1 A. Therefore, there will be no flow into the RCS and DHP-1 A will be on recirculation at a flow of 100 gpm.
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RB Spray will not be initiated because the RB pressure does not reach the ]
30 psig ES setpoint. Therefore the RB Spray pump, BSP-1A will not be l operating.
EFW will be initiated on either low OTSG pressure / level'or ES actuation. I EFIC will initiate isolation of both OTSG's and control EFW flow to the j unaffected OTSG initially at the maximum fill rate of 330 gpm. Since !
steaming and EFW recirculation are occurring at the same time an additional flow of 670 gpm is required to maintain the maximum fill and recirculation rate requirements. The flow from each EFP-1 and EFP-2 is assumed to be 500 gpm during the period to refill the OTSG.
- 2. EFP-2 Failure Case The RB pressure resulting from the FLBA for this failure case is the same as for the 'B EDG Failure Case, i.e., less than 28 psig.
ES actuation is initiated on the 4 psig RB high pressure setpoint. Both HPI pumps, MUP-1A and MUP-1B are initiated by ES actuation and continue to operate throughout the accident. The pumps will each operate at a flow of 295 gpm. This flow was determined by the pump head-capacity curve for a discharge pressure equal to the RCS safety relief valve pressure setpoint. <
The pumps will continue at this flow rate until operator action is taken to depressurize the RCS.
LPI is initiated on ES actuation. Since the RCS pressure remains at the pressure of the safety relief valve setpoint throughout this scenario, the RCS pressure is maintained at a greater pressure than the maximum discharge pressure of both LPI pumps, DHP-1 A and DHP-1B. Therefore, there will be no flow into the RCS and both DHP-1A and DHP-1B will be on recirculation flow of 100 gpm each.
RB Spray will not be initiated because the RB pressure does not reach the 30 psig ES setpoint. Therefore the RB Spray pumps, BSP-1A and BSP-1B will not be operating.
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- EFW' Is' initiated. on loss of RC pumps,11ow OTSG pressure / level ' or ,ES.-
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to the unaffected OTSG initially at the maximum fill rate'of 330 gpri. Since-
- steaming and EFW recirculation are occurring at the same time an additional 1; I
flow ~of 670 gpm is required to ' maintain the maximum fill and recirculation rate requirements. However, the maximum fill rate to'n OTSG (for a single I
OTSG solely available scenario) the flow rate is controlled by EFIC at 740 ;
i GPM. Therefore, the total flow from EFP-1 (the only pump available in this ]
- failure case) will be 940 gpm including' recirculation. EFP-1 is assumed to . l continue at this rate until the OTSG is refilled. -:
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s A.i' STEAM GANERAjOR TRB,JJAILURE ACQJDENT 3
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T e basic accident dynamics ar describad b PbkK Sectlon 14.2.2.2. Since loss of a
offif te poser is asst 6ned, coincident with the steam genaratod tube failure' -
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o ;ident, tqe reactor, thp RC pumps, and the turbine are assumed to be tripped at' time, zero/of this scen uio analysis. The V.D pressure is not affected by this
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Nelhett since the RCS leak is contained within the OTIG.
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- 1. 'B' EDG FaQqLe, ,
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The prirrdry and secondary coolant system pgameters and sequence of' events fo'r this scenario analysis are shdilar to the steam generator tube <. ;
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4< failure accident described in FSAR 3cet'on 14.2.2.2 and FSAR Tables 14-29 !
4/ and 14-30. Ilowever the FSAR analys5 does not account for the reactor trip L
coincitari S:ith the accident because hf ' ghe loss of off-site power for this .
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1< scensrlo[ nalysis.
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s3 The HPl pump (MUP-1 A) is actuated S[the Engineered Safeguards Actuation I System (ESAS) as a result of RC[bresstIre below 1500 psig. The HPI pump is 1 4 conservatively assumed to operate at '.ls maximum flow of 600 gpm to )
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I i g'f compensate for leakage through the, mptucqd tube and RCS shrinkage. This j I j maximum flow will continue until oprator action is taken to reduce RCS
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LPI is initiated by FSAS. The RCS pressup will remain above the maximum -
( , , discharge pressure f of the LPI pump based on' the FSAR description. 1
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RCS and DHP-1A will be on a recirculation at a flow of 100 gpm. '
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tMs' accident. Therefore, the RB Spray pump, BSP-1 A will not be operating.
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EFW will be initiated on loss of RCapumys,' low OTSG pressure / level or ES
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actt'.ation. j EFIC will control EF# flow to both OTSG's initially at the i maximdm fill rate of 8" per minuto (330 gpm). .The affected steam generator is asstsmed to be filling or steaming from the RCS leakage. Since steaming f
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t 4 L(470 gpm) is occurring at the same time, the total flow to both OTSG's would p < > be a total'of 800 gpm.1The total flow' from both EFP-1 and EFP-2 is 1000 gpm.l The total flow includes 200 gpm pump recirculation flow and assumes
~ the RCS leakage satisfied the affected OTSG's maximum fill rate. EFP-1 Is
- ass'umed to contribute 500 gpm flow during the period to refill the OTSG.
2.- t EFP-2' Failure Case
- The primary' and secondary coolant system' parameters and sequence Lof events 1for this scenario analysis are similar, to the steam generator tube.
failure accident described in FSAR'Section 14.2.2.2 ~and FSAR Tables 14-29 and 14-30. However the FSAR analysis does not account for the reactor trip ,
coincident'with' the accident because of the loss of off-site power. for this '
scenario analysis. '
~ HP! pumps, MUP-1 A and MUP-1B,.will be actuated by ES as a result of RCS y pressure below -1500 psig and will operate at an initial total flow of 850 gpm (Ref. B&W Document No.'51-1158-449-00). The flow was assumed to be
- shared equally between the two pumps. MUP-1A flow is therefore 425 gpm.
LPI is initiated by ESAS. The RCS pressure will not decrease below 1000
'psig throughout this analysis. The RCS pressure is always greater than the maximum discharge pressure of ~ both LPI pumps, DHP .1A and DHP-18.
Therefore, there will be no flow into the RCS and both DHP-1A and DHP-1B.
will be on recirculation flow 'of 100 gpm'each. .
RB Spray will not be initiated because the RB pressure does not change for this accident. : Therefore, the RB Spray pumps,' BSP-1A and BSP-1B will not be operating.
EFW will be initiated on either loss of reactor coolant pumps, or ESAS. EFIC -
will control EFW flow to both OTSG's at the maximum fill rate of 8" per minute'(330 gpm). The affected steam generator is assumed to be filling or steaming from the . RCS . leakage. Since steaming (470 gpm) and pump q recirculation (200 gpm) are occurring at the'same time, an additional flow of 670 'gpm is required to maintain the maximum fill and recirculation requirements. The totalrequiredflow from EFP-1 (the only pump available in Gubert Commonwealth A.4-2
- . . - un_ i ___ _ ____ !
this failure case) would be 1000 gpm. EFP-1 would continue at this flow until the OTSG is refilled.
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.I-4 STEAM GENERATOR TUBE FAILURE ACCIDENT ;'B' EDG FAILURE CASE 600 m HPI PUMP FLOW.
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APPENDIX B - DESCRIPTION OF MODIFICATIONS B.1 Modify EFP-1 Auto Trio Circuit M AR T87-10-03-01 and EDG-1 A Monitorinst Timer M AR T87-10-19-01 This modification involves several changes to the existing circuitry and alarms.
First, the auto trip of the Motor Driven Emergency Feedwater Pump EFP-1 after
! 30 minutes and its associated timers and alarms are being removed. Next the
! control of EFP-1 is being interlocked with the Low Pressure Injection (LPI) flow.
l An finally, new mechanical rotary type timers and asso isted alarms are being added to inform the operator when the Emergency Diesel Generator 3A (EDG-3A) has entered its 30-minute rating in excess of 3000 KW. Subsequent alarms have also been provided for 5,24, and 29 minutes into the diesel generator 30-minute rating.
The first modification involves removing the existing control and timing circuitry which trips EFP-1 after 30 minutes of operation following a 4160 volt bus undervoltage and an ES actuation.
The second modification described as follows is being installed as " temporary" for fuel cycle 6 only.
The existing EFP-1 trip logic is being replaced by an AND logic based on coincidence of LPI flow and EDG-3A contact closure. Closure of the EDG-3A breaker arms the circuit through an auxiliary contact, and a selector switch located in the control room on the PSA Panel. If LPI flow goes above 400 gpm a bistable contact closes which actuates a relay and blocks closing of the EFP-1 power source breaker thus tripping EFP-1. A contact from the trip relay provides seal-in of the relay. A Remote Shutdown transfer relay normally closed contact is wired in series with the selector switch and relay coil to prevent inadvertent relay actuation in the event of a control room Appendix R fire scenario.
This logic may be bypassed, from the main control room, by the selector switch on the PSA Panel which prevents the trip relay from being energized. With the selector switch in the BYPASS position the trip logic circuit is defeated and EFP-1 can be started manually or automatically if commanded by EFIC. The BYPASS Gdbert (Commonwealth B.1-1
position also allows normal procedural testing of the DH pumps during plant power operation without disabling the EFP-1 circuit.
In addition, annunciator / events recorder alarms have been included for the LPI Flow Trip and LPI Flow Trip Bypass. <
The " Bypass", " Normal" and "EFP-1 Trip Relay" indicating lights are mounted, as safety related, in the main control board PSA Section. The trip relay is mounted as safety related in relay rack RR3A. The bistable for LPI flow is mounted in the ,
! Remote Shutdown Auxiliary Equipment Cabinet.
Single failure for EFW is met by the redundant 'B' EFW train. Single failure of the instrument string comprised of flow element, transmitter, I/E converter and bistable device is not met; however, the failure is no worse than loss of the 'A' diesel, the 'A' Battery, or the EFP-1 pump failure to start. 'The redundant 'B' EFW train is available and no LPI Flow Trip modifications are being made to its circuitry.
The range of the LPI flow instrumentation is 0 to 5000 GPM. The flow set point of 400 GPM assures that the RCS pressure has decreased to below 185 psig without creating unnecessary trips caused by lesser flows. The flow element is not in the LPI recirculation line and therefore does include this flow in its measurement.
The final modification involves the installation of an elapsed time indicator (ETI),
mechanical rotary timers, and associated alarms for recording the amount of time the EDG-1A, has been operating above 3000kW. The ETI and timers will monitor the cumulative time that EDG-3A operates above 3000 KW, and starts providing alarms anytime EDG-3A is loaded above 3000 KW for more than 10 seconds. A direct current alarm module receives a 4-20 MA signal from the watt transducer monitoring the output of EDG-3A. This alarm module has a 10 second alarm response time delay to preclude alarming during motor in-rush. When the alarm module contact closes an annunciator / events recorder alarm alerts the main control room operator that EDG-3 A is now operating in its 30-minute rating in excess of 3000 KW. Should EDG-3A continue to operate in its 30-minute rating similar alarms would be initiated at 5 minutes,24 minutes, and 29 minutes. Should EDG-3A's loading be reduced to below 3000KW, the amount of time operated in the 30-minute rating will be maintained by the mechanical rotary timers such that aaws comma.u,,n B.1-2
if EDG-3A again goes into its 30-minute rating the timers will continue to time and alarm as described above. A non-resettable digital ETI is mounted on the front of the SSF Section of the main control board to give the operator the cumulative time the EDG-3 A has operated in its 30-minute rating.
.The mechanical rotary timers, alarm module, reset pushbuttons, terminal strips, isolation fuses, and a circuit breaker will be housed in a box and mounted in the back of.the main control board. The box and the isolation fuses are safety related and the rest of the components, including the ETI, are non-safety related. The non-safety related components will be mounted seismically in accordance with anti-fall down criteria.
l l
l ome._._... ;
B.1-3
.-__--____a
. i. j ' . .j t
APPENDIX B - DESCRIPTION OF MODIFICATIONS B.2 - ' ASV-5/204 Power Seoaration M AR T87-10-09-01 This"m' modification provides for the separation' of motor ' operated valves .(MOV)'
ASV-5 and ASV-204. MOV ASV-5 will retain the existing 250/125 volt de ES B '
channel power feed, while MOV ASV-204 will be powered from the 250/125 volt de l t
.ES A channel power source. - MOV ASV-204' will be controlled automatically by an auxillary contact from the 'A' channel EFIC logic or manually by a new control- i
. switch mounted in the PSA Section' of the main control board. Status lights. .l
' indicating valve position will be located with the new control switch. ;
I.
I-
- Channel. However,10CFR50 Appendix R does not apply as ASV-5 will remain on the ES B Power and meet: the present Appendix R commitments. Therefore, control and indication 'will not be included at the itemote Shutdown Panel and-ASV-204'is not required to comply with 10CFR50 Appendix R separation criteria.
. Existing annunciator / events' recorder alarms will be mod'ified to accommodate
~
MOV ASV-204 from the new power source. Additional time delay relays are being added to relay rack RR3A, and powered from the ES A Channel battery, to allow the ASV-204 alarm logic > for pump EFP-2, "Falled to Start", " Auto Start", and ~ l
" Turbine Steam Supply Not Ready" to operate in conjunction with . ASV-5 alarm l logic. .
l 1
Gilbert < Commonwealth B.2-1 4
p .
4 ~
APPENDIX B - DESCRIPTION OF MODIFICATIONS I .
B.31 EDG-1 A Emergency Load Shedding - Heat Tracing M AR T87-10-03-01:
l'
-Thisimodification provides for the automatic load shedding of the following heat -
tracing transformers from their respective power sources. Currently these-transformers are fed directly from the circuit breakers in the MCC's.
Heat Tracing Transformer Power Source-HTTR-1A 480V ES'MCC 3 A1, Unit 11AR HTTR-2A 480V ES MCC 3A2, Unit 6AL HTTR-3 A 480V ES MCC 3 A1, Unit 11AL HTTR-4A 480V ES MCC 3A2, Unit GAR HTTR-5A -480V ES MCC 3A2, Unit SAR This.modication provides for the automatic load shedding of the heat tracing load from the Emergency Diesel Generator 3A following a 480 V ES Bus undervoltage condition coincident with an ES actuation.Re.energization of the affected heat tracing will be manual operation by the control room operator.
Status lights are mounted on the main control board PSA Section to indicate the energized /deenergized condition of the wall-mounted contactors. Also mounted- l on'the PSA Section of the main control board is a . test pushbutton for periodic testing of the wall-mounted contactors.
cThe contactors providing the isolation are seismically mounted in two boxes local to the respective MCC power source. The contactors for heat tracing transformers HTTR-1 A and HTTR-3A are mounted in a box local to ES MCC 3Al and- the contactors for heat tracing transformers HTTR-2A, HTTR-4 A, and -
HTTR-5A are mounted in a box local to ES MCC 3A2. Along with the contactors, In' each box is a 480/120 volt transformer, fuse, terminal block, and reset f f
- pushbutton. The box, components and wiring within are considered safety related f
[.
- and are seismically and environmentally qualified for their respective mounting L l GdberMommonwealth B.3-1 m________m-_ __m_ _m
i locations. The existing power feeds downstream of the contactors are considered non-safety related.
i I
/'
Gelbert Comm0RWedith B.3-2
MODIFICATIOtl SAFETY EVALUATION l
t .
Sheet 1 Of;3 MAR NO. T87.10 -
. 04'. -01 SAFETY EVALUATION: Answer the following questions and provide specific justi!! cation (use attachment il necessary).
l.. _Is the probability of an occurrence or the consequence of an accident or malfunction, of equipment important to safety as previously evaluated in the Final Safety Analysis Report, INCREASED 7 .YES NO X Because: : ,
See Attached Sheet
- 2. Is the possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report, CREATED 7 YES __, NO X
.Because See Attached Sheet ,
- 3. Is the margin of safety, as defined in the basis for any Technical Specification, REDUCED 7 YES ' NO X Becauset-
.See Attached Sheet LICENSE REVISION REQUIRED: Final Safety Analysis Report: NO Technical Specification: YES X_ , f YES NO'X'
~ !
. NRC Authorization for Change Required: YES ~ ' 'NO I Semi-Annual Reporting to NRC Required: YES[ NO Z .j 10CFR30.39 CHECKLIST -
Does the proposed action enange the Final Safety Analysis Report or require additional description to be added to the F Final Safety Analysis Reoort?
YES ( ) NO (X )
Notify Manager, Nuclear Licensing y .
and Fuels Manaeement
, y t
15 a Change to the Technical Specifications Required? i YES( ) v NO (X )
w is any unreviewed safety question involved, i.e.,
(1) is the probability of an occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evaluated in -
the Safety Analysis Report increased? YES _ NO X_
(2) Is the possibility for an accident or malfunction c! a different type than any previously evaluated in the Safety Analysis Report created?
YES NO X (3) Is the"~marginEsafety, as defined in the basis for any Technical Specification reduced? YES __ . NO L v
Any answer YES ( ) h All answers NO (X)
Request and receive NRC Authorization for change u
Document Change including (1) Description of change (2) Written Safety Evaluation which provided basis ,
f or items (1), (2), and (3) above. I Authorization Received ( ) Description Safety Evaluation Complete h d Initiate Installation of Modification
- Required changes to Technical Specifications Prepared by
. should be processed in parallel to this checklist. ;
Name Date
-_a_ _ _ _ - --
hoNok" ANALYSIS / CALCULATION 9 corporation Crystal River Unit 3 SHEET 2 OF 3 REl/ MAR No. M A R T87-10-04-01 Date OCTOBER 19,1987 3 Project : MODIFY EFP-1 AUTO-TRIP CIRCUIT i ATTACHMENT TO MODIFICATION SAFETY EVALUATION
- 1. The proposed modification replaces a time delay whose actuation energizes an auxiliary relay (BD). The contacts of the BD relay are wired into the i START and STOP circuit of EFP-1. The initiation of the timer circuit is started when the diesel generator enters into its 30 minute 3000KW rating.
This action occurs for every accident scenario analyzed in the FSAR if the Diesel Generator 'A' is required to operate above its 3000KW load rating. A switch and indicating light is provided in the Control Room PSA Panel for warning and reset functions.
The proposed modification would utilize an AND logic of two relay contacts connected in series through a selection switch to the same relay (BD). One of the contacts exists in a diesel generator loaded auxiliary relay and the other contact would be derived from a bistable which would trip on LPI flow at a predetermined setpoint. The AND logic of this circuit would state
" diesel loaded AND LPI flow AND the select switch in the NORMAL position then energize Relay BD to trip EFP-1".
The consequence of a malfunction of Relay BD will not increase the occurrence of accidents as previously defined in the FSAR. The consequence .j of a malfunction of a single time delay relay has been decreased by the j replacement of that relay with the AND logic contact arrangement. The l malfunction of a single component of the logic circuit will not increase the l occurrence of accidents as previously defined in the FSAR. l
- 2. The EFW' System does not create accidents of a different type than any evaluated in the FSAR. A malfunction or single failure of either the 'A' Battery, 'A' Diesel, or EFP-1 has already been considered. The 'B' Battery,
'B' Diesel, or EFP-2 is the redundant system to provide EFW. However, for large LOCA scenarios, the RC pressure decreases to values allowing LPI flow. At these pressure values, heat transfer from the Primary to Secondary System cooling is not occurring, and therefore, steam is not available to drive the turbine driven pumps, i.e. no EFW is occurring. Hence, new accident scenarios are not created. For other scenarios (with an EFP-1 failure) where steam is available, the turbine driven pump provides EFW flow.
l-Design Engineer Date verification Engineer Date Supervisor, Nudear E ngineering Date Rev 7/81 912244
$}ohg" h corporation AN ALYSIS / CALCULATION Crystal River Unit 3 SHEET 3 OF 3 REl' MAR No, Date OCTOBER 19, 1987
' MAR T87-10-04-01 Project : MODIFY EFP-1 AUTO-TRIP CIRCUIT
- 3. The margin to safety, as defined in the basis for Technical Specifications for diesel generator loading above 3000KW, has not been reduced by the modification of AND logic circuit to trip EFP-1 on LPI flow coincident with diesel generator loading. Tripping of EFP-1 upon conditions of LPI flow does not reduce any margin to safety as defined in the Technical Specification.
ADDITIONAL INFORM ATION The proposed modification for tripping the EFP-1 pump upon initiation of LPI flow has been evaluated for its impact on plant safety. The modification causes an automatic trip of the EFP-1 pump upon coincident indication of LPI flow > 400 gpm and diesel "A" loading. These indications are chosen to define a large break LOCA with a coincident loss of offsite power. Under these conditions, EFW flow is not required for accident mitigation since the Primary and Secondary Systems are thermodynamically uncoupled.
LPI flow to the RCS cannot occur for RCS pressures above ~185 psi, the shutoff head of the LPI pump. Thus, events which result in primary pressures above this value will not be impacted by this modification. The only event, other than LOCA, which can result in a large Primary System depressurization is the steam line break accident. The steam line break accident, however, cannot result in pressures below ~600 psi without multiple safety grade equipment failures. Thus, consideration of events other than LOCA is not required.
For large and intermediate break size LOCAs, the Primary System pressure will eventually decrease and allow LPI flow. By this time in the scenario, the secondary cooling provided by EFW flow is not required and has stopped due to extensive Primary System voiding.
A small break LOCA will not result in a sufficient RCS depressurization to allow LPI flow. A review of the LPI injection line concluded that a break in this flowpath would not result in a small break LOCA (when EFW flow is desired) with the LPI pump feeding the break at high flows. Therefore, the only events which will result in LPI flows are large and intermediate break LOCAs, when EFW flow is not required for accident mitigation. This potential scenario was assessed to determine if the trip logic could be initiated for circumstances requiring EFW flow.
oesign E ngineer Date Verification Engineer Date Supervisor, Nuclear E ngineering oate Rev 7/81 912244
L
'. APPENDIX C - SAFETY EVALUATIONS - EDG-1A TIMER MODIFICATION SAFETY EVALUATION Sheet '1 Of 2
' MAR NO. T8 7.110 .19 . 01 1
, SAFETY EVALUATION: Answer the following questions and provide specific justification (use attachment if necessary).
't. Is the probability of an occurrence or the consequence of an accident or malfunction.
i .
'of equipment important to safety as previously evaluated in the Final Safety Analysis, Report,1NCREASED7 YES NO X
.Because -
~See Attached Sheet
- 2. Is the possibility for an accident or malfunction of a different ' type than any previously evaluated in the Final Safety Analysis Report, CREATED 7 YES NO X Because:
See Attached Sheet iP
- 3. Is the margin of safety, as defined in the basis for any Technical Specification, REDUCED? YES NO X-Because:
-See Attached Sheet
= LICENSE REVISION REQUIRED: Final Safety Analysis Report YES 1 NO Technical Specification: YES _ NOL_
i NRC Authorization for Change Required: YES 'NO Y Semi-Annual Reporting to NRC Required: YES _ NO P 10CFR$0.59 CHECKLIST Does the proposed action change the Final Safety Analysis Report or require additional description to be added to the Final Safety Analysis Report?
YES( ) NO (X )
Notify Manager, Nuclear Licensing 3r .
and Fuels Manaeement 3,
W w
is a Change to the Technical Specifications Required?
YES( ) 3r NO (X )
w is any unreviewed safety question involved, i.e.,
(1) is the probability of an occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased? YES _ NO1 ,
(2) is the possibility for an accident or malfunction of a different type
' than any previously evaluated in the Safety Analysis Report created?
YEs ~ NO X (3) is the margindsafety, as defined in the basis for any Technical Specification reduced? YES _ NO1 v
Any answer YES ( ) h All answers NO (X)
Request and receive NRC Authorization for change 1r Document Change including:
(1) Description of change (2) Written Safety Evaluation which provided basis f or items (1), (2), and (3) above.
Authorization Received ( ) Description Safety Evaluation Complete h E Initbte installation of Modification
' Required changes to Technical Specifications Prepared by should be processed in parallel to this checklist. Name Date 1
-- __. - - }
l'
$>Ih" AN ALYSIS / CALCULATION
@ Corporation o Crystal River' Unit 3 SHEET 2 OF 2 REl/ MAR No, M AR T87-10-19-01 Date ' OCTOBER 19,1987 !
l i
Project : EDG-1 A MONITORING TIMER l ATTACliMENT TO MODIFICATION SAFETY EVALUATION
- 1. A malfunction of any component in the Elapsed Time Indicator (ETI) or in the 3 three timer alarm circuits, being added to provide information whenever the diesel is .in its 30 minute 3000KW load rating, will not increase the occurrence or consequence of accidents from those defined in the FSAR.
This circuit provides no safety actuation or control functions. l FSAR Sections 7.2.4, 8.2.2.6, and 10.2.1.6 have been reviewed. j
- 2. The ETI and timer alarms provide no control or protective actuations and, therefore, will not create accidents or malfunctions of a different type than any defined in the FSAR.
FSAR Sections 7.2.4, 8.2.2.6, and 10.2.1.6 have been reviewed.
- 3. The ETI and timer circuit are non-safety and are not included in the Technical Specifications.
Technical Specification Sections 3/4.7.1 and 3/4.8.1 have been reviewed.
I l
Design Engineer Date verification E ngineer Date Supervisor, Nucleat E ngineering Date Rev 7/81 912244
y.,-_w-------- __ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - -----
f
! APPENDIX C - SAFETY EVALUATIONS - ASV-204 MODIFICATION SAFETY EVALUAT10N
' Sheet 1 Of 3 MAR NO. T8 7. 10. 09 . 01 SAFETY EVALUAT10N:
Answer the following questions and provide specific justification (use attachment il necessary).
1, 18 the probability of an occurrence or the consequence of an accident or malfunction of equipment tmportant to safety as prevsously evaluated in the Final Safety Analysis Report,1NCREASED? YES NO X Because:
See Attached Sheet 2.
Is the possibility for an accident or malfunction of a different tvoe than any previously Because:
evalusted in the Final Safety Analysis Report, CREATED? YES _ XNO 3 See Attached Sheet 3.
Is the margin of safety, as defined in the basis for any Technical Specification, REDUCED? YES NO X Becauset See Attached Sheet LICENSE REYl5 ION REQUIRED: Final Safety Analysis Reoort:
Technical Specification: '
YES NO X YES ~ NO T NRC Autnort:ation for Chance Required: YES'~ "NO Y 5emi. Annual Reporting to NRC Required: ~~
YES _ NO1 10CFR50.59 CHECKLIST Does the proposed action cnange the Final Saf ety Arutysis Report or require additional description to be added to the Final Safety Analvsis Recort?
YES( )
NO(X)
Notif y Manager, Nuclear Licensing y and Fuels Manacement ,
v
- {'
is a Change to the Technical Specifications Required?
YES( ) w I
NO(X) is any unreviewed safety question involved, i.e (1) ls the probability of an occurrence or the consequences of an accident, or rnallunction of equipment important to safety prevtously evaluated in the Safety Analysis Report increased? YES NO X (2) 15 the possibility for an accident or malfunction of a ditferent type than any previously evaluated in the Safety Analysis Report created?
YES NO X (3) is the margin of safety, as defined in the basis for any Technical Specification reduced? YES _ NO X Any answer YES ( ) l y y All answers NO (X)
Request and receive NRC Authorization for chance ._
Document Change including: i (1) Description of change (2) Written Salcty Evaluation which provided basis 1 ior items (1), (2), and (3) above. '
Authar:2ation Received ( ) l l Description Safety Evaluation Comotete Y i {
v -
Initiate installation of Modification
'Reovired changes to Technical Specifications Prepared by _
Shoul3 be procenec in parallel to this checkhst.
Name Da t e
_ . .. l-
a "
, A .'ALYSIS / CALCULATION SHEET 2 OF 3 "a corporation Crystal River Unit 3 REl/ MAR N -
MAR T87-10-09-01 Date OCTOBER 19,1987 l P'oi"t :
ASV-5/204 POWER SEPAR ATION i
l ATTACHMENT TO MODIFICATION SAFETY EVALUATION l l
i
- 1. ASV-5 and ASV-204 are motor operated valves having identical functions of supplying steam to the turbine driven Emergency Feedwater Pump (EFP-2).
Since EFP-2 is the ES "B" channel pump, ASV-5 and 204 were electrically t connected in parallel to a common 250/125 VDC ES "B" channel power and [
control source. This modification electrically separates ASV-204 from ASV- l 5 and repowers ASV-204 from 250/125 VDC ES "A" channel power. Also, i separate control room controls and separate "A" channel EFIC interlocks are '
being provided for ASV-204. Automatic control logic of' ASV-204 has not ,
changed. Therefore, the probability of an occurance or the consequences of '
an: accident or malfunction of equipment important to safety as previously evaluated in the FSAR is not increased since the logic of automatically ;
opening ASV-204 whenever the EFIC System calls for emergency feedwater 4 has not been altered. The reliability of EFP-2 has actually been increased because with this modification either "A" or "B" train power will control and operate one of the steam inlet valves to EFP-2 as opposed to both valves being "B" train powered. FSAR Sections 7.2.4, 8.2.2.6 and 10.2.1.6 have been reviewed.
- 2. The electrical separation of ASV-204 from ASV-5 does not impact the design function of either valve to supply steam to the EFP-2 turbine. Power and -
control for ASV-5 is not affected by this modification and ASV-5 retains its automatic control logic, remote manual control, local manual control and remote shutdown isolation and control. ASV-204 is being powered from the y redundant power channel, and will be provided with its own remote manual i control and with separate EFIC interlocks for automatic operation. The type of remote manual control and automatic operation of ASV-204 is the same as for ASV-5. Therefore, based on the above, the possibility for an accident or i
malfunction of a different type than any previously evaluated in the USAR is not created. FSAR Sections 7.2.4, 8.2.2.6, and 10.2.1.G have been reviewed. l Design Engineer Date Verification Engineer Date Supervisor. Nuclear Engineering Date !
Rev 7.31 _ __
}
912244 L________.__._.... _ i
i corporation ANALYSIS / CALCULATION Crystal River Unit 3 SHEET 3 OF -
3 1
Ril/ MA R No. Date A1AR T87-10-09-01 OCTOBER 19,1987 Project:
ASV-5/204 POWER SEPARATION i
-i
- 3. This modification enables the turbine driven Emergency Feedwater Pump (which is the "B" channel pump) to be operational even if a failure should occur on the "B" channel power system for which shutdown operation would ,
be via the "A" channel systems. With this capability, the turbine driven EFW !
pump is able to operate and share the EFW requirements with the "A" channel motor driven EFW pump. This will reduce the electrical load on the "A" channel diesel generator for the condition of an ES actuation coincident with a loss-of-offsite-power and failure of the "B" channel power system.
Consequently, with this modification the margin of safety, as. defined in the basis for any Technical Specification, is not reduced. It is actually enhanced because of the increased availability of the turbine driven Emergency Feedwater Pump. Technical Specification Sections 3/4.7.1 and 3/4.8.1 have been reviewed, i
i l
i i
Design Engineer Date Verification E ngineer oate Supervisor, Nuclear Engineering oate
'Rev 7,81 932244
MPEND1X C - SAFETY EVALUATIONS - HEAT TRACING MODIFICATION SAFETY EVA1.UAT10N MAR NO.Ij]- 10 . 03 - 01 Sheet 1 Of 2 SAFETY EVALUATION:
Answer the following questions and provide specific justification (use attachment 11 necessary).
1.
Is the probability of an occurrence or the consequence of an accident or malfunction of equipment important Report,1NCREASED? to safety as previously evaluated in the Final Safety Analysis, i YES NO X !
Because: '
See Attached Sheet 2.
Is the possibility for an accident or malfunction of a different tvoe than any previously Becauset evaluated in the Final Safety Analysis Report, CREATED? YES _ NO 1 See Attached Sheet 3.
Is the margin REDUCED? YESof safety, as defined in the basis for any Technical Specification, I NO X Becauset i i
See Attached Sheet LICENSE REYl510N REQUIRED: Final Safety Analysis Reoortt Technical Specification: YES NO X NRC Authorization for Chanee Required:
YES NO T Y ES ~ 'NO T Semi-Annual Reporting to NRC Required:
YES [ NO Z 10CFR50.59 CHECKLIST Dces tne proposed action change the Final Saf ety Analysts Report or require additional description to be added to the j Fin 3l Saf ety Analysis Reoort? 4 YES (X)
NO( )
Notity Marueer, Nucicar Licensing y and Fuels Manacement .
v T
" T Is a Change to the Technical Specifications Required?
YES( ) w I
NO(X) is any unreviewed saf ety question involved, i.e.,
(1) is the probability of an occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evalasted in the Safety Analysis Report increased? YES NO X (2) is the possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report created?
YES NO X (3) is the73arginHsafety, as defined in the basis for any Technical Specttication reduced? YE5_ NO l Any answer YES ( ) l v y All answers NO (X)
Request and receive NRC )
Authorization for chance _
t MP Document Change including:
(1) Description of change i
(2) Written Salety Evaluation which provided basis f or items (1), (2), and (3) above.
i Authorization Received ( )
Desertotion Safety Evaluation Comolete d
initiate installation of Modification)
- Reovired changes to Technical Specifications should t>e processed in parallel to this checklist. Prepared by _
Name Da te
_ _ . . )
S*,h corporation ANALYSIS / CALCULATION Crystal River Unit 3 SHEET 2 OF 2 Ril/ MAR N -
M AR T87-10-03-01 Date OCTOBER 5,1987 !
Project :
EGDG-1A EMERGENCY LOAD SHEDDING - HEAT TRACING 1
ATTACHMENT TO MODIFICATION SAFETY EVALUATION '
\
1.
This modification provides for the automatic load shedding of the heat tracing load from the Emergency Diesel Generator 3A (EDG-3 A) following a 480V ES bus undervoltage condition coincident with an ES actuation. The shedding of the heat tracing loads will decrease the possibility of overloading EDG-3 A. .The heat tracing loads are only required to maintain the temperature of-portions of Chemical Addition System during plant operation and are not required for accident mitigation.
FSAR Sections 6, 8, 9, and 14 were evaluated in regard to this modification.
The accidents evaluated in FSAR Section 14, " Safety Analysis", are not affected by this modification. Therefore, this modification will not increase the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety, as previously evaluated in the FSA R.
2.
The proposed modification provides the means to automatically remove non-essential load (heat tracing) from the diesel loading cycle. The modification does not degrade the performance of any safety system it is associated with, and- does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR, as heat tracing is maintained until accident initiation and not relied upon thereafter for accident mitigation. FSAR Sections 6, 8, 9, and 14 were reviewed to make this determination.
3.
The new components and control circuitry are being connected as part of the Auxiliary Power System (System MT). The only interface with the existing "MT" System is use of an existing spare normally closed undervoltage lockout relay contact. This modification does not change the ratings, function, or configuration of the Auxiliary Power System. Therefore, this modification does not affect the margin of safety or basis upon which the technical specifications are written for the Auxiliary Power System or the Doration Sys t e ms.
Technical Specification Sections 3/4.1.2 and 3/4.8.1 were -
examined.
cesign E ngineer Date Verification Engineer Date Supervisor, Nuclear Engineenng Date j Rev 7 81 912244 L
w- - -