ML20236Q301

Rev 1 to Emergency Diesel Generator Loading Evaluation

ML20236Q301
Person / Time
Site:	Crystal River
Issue date:	11/13/1987
From:	FLORIDA POWER CORP., GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT
To:
Shared Package
ML20236Q289	List: 3F1187-19, Application for Amend to License DPR-72,consisting of Rev 1 to Tech Spec Change Request 157,revising Analysis Calculations to Modify Tech Specs Re Diesel Generator Loadings.Certificate of Svc Encl ML20236Q294 ML20236Q301 ML20236Q309 ML20236Q317
References
NUDOCS 8711190301
Download: ML20236Q301 (69)

Text

{{#Wiki_filter:_ _ _ _ __ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. 1 l I 1 EMERGENCY DIESEL GENERATOR l LOADING EVALUATION l 1 l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 10-23-87 Rev.1 13-87 l 8711190301 871116 PDR ADOCK 05000302 P ppg j

l TABLE OF CONTENTS . 1.. '. EXECUTIVE

SUMMARY

II. ' MECHANICAL METHODOLOGY III. ELECTRICAL METHODOLOGY I V. ~. RESULTS i-V. TABLES . VI. APPENDICES A. Scenario Writeups B. Modification Descriptions C. Safety Evaluations of Modifications l. i' l l1 l

r-L 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 engineered 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. Per the above criteria a load table was constructed based on nameplate rating for the large motors and fans connected to Emergency Diesel Generator 3A. The results of this approach indicated an automatic diesel loading of 3449 KW with 372 KW of potentially manual applied loads which exceeded the manufacturers rating of 2750 KW continuous,3000 KW for 2000 hours and 3300 KW for 30 minutes. 4 To remove some of the conservatism inherent in the nameplate values actual i brake horsepower (bhp), requirements of the pumps were calculated based on full flow conditions of the various safety systems. These values were then j translated to motor load requirements using nameplate motor efficiencies and converted to Kilowatt based loads. The results of this approach i i indicated an automatic diesel loading of 3379 KW with 372 KW of potentially ) i Gdbert COmm0nwedith l-1 1

manual applied loads Again, due to the inherent conservatism of this approach, manufacturers ratings were exceeded. In the final analysis the diesel loading for each of the following Design Basis Accidents were evaluated to more accurately model the expected station emergency loadings. 1. Loss of Coolant Accident (LOCA) A. Large Break B. Intermediate Break C. Small Break ' 2. Steam Line Break Accident (SLB) A. Inside Containment B. Outside Containment 3. Feedwater Line Break - Inside Containment 4. Steam Generator Tube Rupture The above Design Basis Accidents were evaluated as these scenarios would challenge the Engineered Safeguards Actuation System. Single failures considered were: 1 1. Loss of The Emergency Diesel Generator 38 l l 1 2. Loss of the turbine driven emergency feedwater pump EFP-2. I Gilbert Commonwealth I l-2 l \\

,d l The loss of EFP-2 was reviewed due to the sensitivity of the calculation to j increased flow requirements of the motor driven pump, to assure the limiting case was assessed. System flow requirements were established for each pump for each of the a bove 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 improve long term diesel loadings and provide additional benefits to plant operation and reliability the following design enhancements were identified: A. Removal of 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. Also, an elapsed time indicator will be provided to display accumulated time of diesel operation above the 2000 hour rating. 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 38. Due to recent modifications to the emergency feedwater I system control this provides load sharing capability for the two emergency feedwater pumps thus improving plant reliability while l 1 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 related and not required for accident mitigation. Based on incorporation of these design changes the worst case diesel loading for the above scenarios is 3248 KW, ca. c o..., 1-3

u This results in the Emergency Diesel Generator 3A operating within its 30 minute ' rating of 3300 KW, thereby assuring its availability for accident mitigation. Prior to exceeding this rating, time frame operator action will be taken to maintain the diesel loading within its 2000 hour rating of 3000 K W. 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 with their safety evaluations are provided. i Gdbert Commonwealth I-4

1 II. MECHANICAL METHODOLOGY A. Approach The accident parameters are addressed in this report only for the l purpose of defining the ES loading requirements for the emergency i 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 actual requirements identified in the accident analyses during the various time intervals. k l Accident scenarios were developed to define the equipment power requirements for their specific operating modes during Design Basis Accident. 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 adjustmer.ts 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 diesel loading. 1 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 following accident analyses for those events are included in this report as they result in automatic actuation of the ESAS: Gilbert Commonwealth 11-1

I i 1. LOCA - Small, Intermediate, and Large Size Breaks l 2. Steam Line Break Accident 3. Steam Generator Tube Failure Accident i 4. Feedwater Line Break Accident -Inside the Reactor Building The following accidents do not result in automatic actuation of the ESAS thus requiring reduced diesel loading and are not presented in this report.. 1. Fuel Handling Accident i 2. Rod Ejection Accident 3. Waste Gas Tank Rupture Accident i 4. Loss of Feedwater 1 5. Feedwater Line Break Accident - Outside the Reactor Building l 6. Uncompensated Operating Reactivity Changes j l 7. Start-up accident i 8. Rod Withdrawal Accident at Rated Power Operation 9. Moderator Dilution Accident q 10. Cold Water Accident 11. Loss-of-Coolant Flow 4 12. Stuck-Out, Stuck-In, or Dropped Control Rod Accident 13. Loss of Electric Power 14. Maximum Hypothetical Accident (MHA) 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 on the primary, secondary, and containment system parameters Gilbert Commonwealth 11-2

r relative to pressure, temperature and flow. The following systems and their major equipment are included: I 1. High Pressure injection (HPI)- MUP-1A and MUP-1B 2. Low Pressure injection (LPI) - DHP-1 A and DHP-1B 3. Reactor Building Spray System (BS)- BSP-1A and 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 Lighting Additionally there is equipment whose operating characteristics are independent of the accident and its results but are deperdent 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 available to provide the design flow required: 1. Emergency Nuclear Services Sea Water Pumps - RWP-2A and RWP-2B Gilbert Commonwealth 11-3

l . 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 within the plant. These equipment power loads are manually applied to the diesel generator at a time which is dependent on the initial i i operating conditions of the systems or at a procedurally identified time. The following equipment is required to be loaded onto the diesel i generator within 10 minutes: J 1.~ Control Complex Emergency Duty Supply Fans - AHF-18A or AHF-188 2. Control Complex Return Air Fans - AHF-19A or AHF-198 l 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 discretionary action by the operator can be exercised. This was considered a sufficient time period in which the characteristics of the equipment could be defined for the mitigation period of an accident where automatic equipment actuation would occur. Beyond this time Gilbert Commonwealth 11-4 1

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 hor 9epower 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 torsepower 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 - t of the 250/125 vde system with the resultant loss of the emergency q diesel generator 3B, EDG-38, due to loss of field flashing and control. j 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, EFP-2 is turbine driven, EDG-3B does not include an equivalent load. The other ES equipment horsepower provide nearly identical load requiren' ents for EDG-3A 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. The second single failure case evaluated was the EFP-2 Failure Case. 1 As the first single failure case was the loss of EDG-3B, the turbine driven EFW pump, EFP-2, was assumed to be available and pumping in Gdbert' Commonwealth 11-5

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 the horsepower requirement of EFP-1. The EFP-2 Failure Case was based on both EDG-3A 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 l Emergency Nuclear Services Closed Cycle Cooling Water Pumps (SWP-I A and SWP-1B). These systems have two redundant pumps operating in parallel to supply water through a common piping system to the components requiring cooling water. With the two pumps operating the individual horsepower of each pump is reduced. In some scenarios the flow requirements of the ES pumps are reduced as the required flow is shared by 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. ) I ) l GWaert' Commonwealth 11-6

[ B. Scenario Assumptions 1 i 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 I 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. Where 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 3B, are assumed to be available and both the 'A' and 'B' trains of the ES equipment are assumed operating. It is assumed that until operator action is taken to shutdown any of the equipment on either EDG-3A 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 Sea 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 total system flow based on the system resistance curves generated for this mode of operation. 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: 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 trains, (4) low Steam Generator Pressure, or, (5) low Steam Generator level. Any of the initiating signals will cause actuation and control of Emergency Feedwater (EFW). Only low Steam Generator pressure causes Main Steam isolation and Main Feedwater Isolation. Geltwt Commonwealth 11-7 l J

'\\ The EFIC System is depleted 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 j valves from EFP-2 whenever the EFIC logie 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 i 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. 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 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 fer 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 flow from EFP-2 since EFIC 'B' control is lost. This occurs immediately upon loss of power while the 'A' powered motor driven pump is being load sequenced after the 'A' diesel is running. The valve opens and the turbine runs up to speed in approximately 15 seconds. Gdbert Commonweeth 11-8 - --- -- _ - _ - _ _j

The EFP-2 pump will continue to run without '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 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 65%. The EFIC 'C' Initiate signal will alternate initiate CLOSE-OPEN signals according to the OVERFILL and OVERFILL RESET setpoints if any form of Steam Generator boil-off is occurring. As the loss of 'B' battery did not result in the worst case loading on EDG-3A it was not analized further. For the 'B' EDG Failure Case EFP-1 and EFP-2 would be available and each would share the flow requirements dictated by EFIC. It is assumed that the EFW flow rate to remove the core decay heat by j steaming in the once through steam generators (OTSG) is 470 gpm, unless l 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. 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 EFP-1 flow is assumed for the entire length of each scenario. The maximum fill rate excluding steaming controlled by EFIC during the 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. The EFW pump re-circulation line will pass 200 gpm during all modes of pump operation based on actual flow test results. With two pump parallel operation it is assumed the recirculation rate of 200 gpm will be shared by each pump. GilbertiCommonwealth 11-9

c. I l To obtain conservative flow estimates for the pumps the maximum flow values permitted by flow control devices in the system were assumed. The LPI pumps, DHP-1 A and DHP-1B are limited to a maximum flow of 3150 gpm (Ref. FSAR Section 6.1.2.1.2). The RB Spray pumps, BSP-1A AND BSP-18, are limited to a maximum 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 watee 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-1A 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 conditions 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 e.xists for such an action and after 30 minutes if no procedure is in place. Gilbert Commonwealth Il-10

I: III. 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 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 the worst case 4kV and 480V loads and determining worst case cable losses of 0.5kW and 2.0kW 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 battery chargers, inverters, and miscellaneous power distribution panels. 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 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 for the large horsepower pumps, as in most scenarios the motors are operating at or near full load. In the scenarios where the Makeup and Purification Pump, Decay Heat Pump and/or the Emergency Feedwater Pump will be operating in the recirculation mode, reduction in efficiency is considered negligible. This Gdbert Commonweetth III-1

I assumption is based on review of typical squirrel cage induction motor efficiency versus load curves. i 5. Small motor loads (less than 100 hp) are considered to operate at nameplate horsepower ratings. B. APPROACH l The first step in performing the. emergency diesel generator.3A (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 tctal 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 undervoltage signal and automatically re-energized in sequence as blocks 2, 3, 4 and 5, and (3) manual loads that by procedure must be manually connected prior to the operator taking administrative action to shed any unnecessary loads. The individual loads were analyzed to determine the amount of kW and kVA each 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 grouped into four basic categories: Motor loads of 100 horsepower or larger. Motor loads less than 100 horsepower. i Non-Motor loads. Motor Operated Valves. 1 i Gdbert'CommonweaHh l III-2

c-

.4 l

r. Il4 l 3 ' 1. - Motor loads of'100 hp or larger: 'These are pump motors, which comprise the major portion of the diesel load and they have been analyzed' for 4 specific loading.. cases. l Case 1 - Motor kW based on nameplat'e horsepower rating. 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 f generator. ~ Case 4 - Motor kW based on the pump flow rate required for the accident scenario resulting in the worst case diesel i loading with a single failure of the turbine ' driven emergency feedwater pump. For each of.the above cases the motor kW was determined as follows: Motor kW = -- Pump Brake Horsepower X 0.746 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. 1 l 1 i ] o m-._._.. III-3

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, 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 non-motor loads were aetermined 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 the valve motor operators was determined by summing the full load amps (FLA) and locked rotor amps (LRA) of the valves that are actuated under an ES condition. The kW load 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 LRA 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 performed for each of the four loading Cases (Ref. Item 1 above). GilbertCommonwedith l III-4

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 total diesel load to determine the total kW running load that is automatically sequenced onto diesel generator EDG-3 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-poweE 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: Gelbert Commonwedith III-5

g 1 (g Y "1 % Voltage Dip =.100 / [1 + GRKVA

• 100 / (SKVA
• Cx"d)]

Where GRKVA= Generator Rated kVA SKVA = Inrush Reactive Power Locked Rotor. Amps

• Volts
• 1.73/1000

= Cx"d = Corrected Sub-Transient Reactance (x'd - x"d)

• 2/3 + x"d where

= 4 Transient Reactance of generator in percent x'd = Sub-Transient Reactance of generator in percent x"d = Results in each case were comparable. l 1 Gilbert' Commonwealth III-6 -_________-.-______a

i IV. RESULTS From the KW totalloads depicted on Tables 1 and 2, it is concluded that with the implementation of the design changes descrioed in Appendix B of this report, the total diesel generator KW running load is below the 30 minute j maximum rating of 3300 KW for each of the postulated accident scenarios. Prior to exceeding this rating time frame, operator action will be taken to maintain the dieselloading within its 2000 hour rating of 3000 KW. j From Table 3 it is concluded that the voltage dips produced by each of the block loads are acceptable levels, and will not cause de-energization of any safety related electrical components. i Gilbert C OM'tWtonwealth IV-1 .________u

T o 3 h &2( ls A A a 8 9 7 7 7 7 7 7 7 1 i 6 6 6 6 6 6 6 t 1 F P i I l n I l i A A e h N t L W-I 1 1 5 4 5 1 4 t D M A K T A 8 8 9 7 3 9 7 u 0 Go1 1 1 8 8 9 7 8 o oDL 3 3 2 2 2 2 2 T h E T g uo G) r Nl I t h Ns 4 4 4 4 4 4 4 t I D 3 3 3 3 3 3 3 t A A 5 5 5 5 5 5 5 n MO a T E L t R s E n E o A 0 0 0 0 0 0 0 c H 0 00 0 00 00 0 00 0 000 000 000 n 2 S P 2 5 0 2 50 25 0 25 0 250 250 250 i W 5 67 56 7 56 7 56 7 567 567 567 a Y 1 1 1 1 1 1 1 M R me A r M A 0 0 0 0 0 0 0 d 401 4 01 401 n 1 4 01 40 1 4 0 1 4 0 1 M 4 82 4 8 2 48 2 48 2 482 4 82 482 P a U W 5 62 56 2 56 2 56 2 562 562 562 1 1 1 1 1 1 1 s s i S sy l' D l 1 a E A 2 20 22 0 00 0 93 0 005 930 930 n L O P 1 4 3 1 4 3 59 5 26 0 946 260 260 a F 5 64 56 4 56 5 56 5 576 565 5' 65 L E o H i A E r S a T A n A 0 e 0 C 9 4 94 00 0 00 0 000 000 000 s 7 0 70 c 1 0 0 P 6 6 E S 1 2 1 2 f 1 1 R B o t U n L e I A 0 0 d A 6 3 63 4 7 0 4 7 0 4 70 4 70 4 70 n 1 7 5 75 05 0 05 0 050 050 050 5 5 P F i 2 3 23 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 e 2 2 l 3 3 p D e GD dn E 6 a' 3 i A 60 60' 27 5' 275 e 'B 0 7 0 07 0 07 0 // 2 //2 279 070 a I 7 30 73 0 73 0 56 9 569 605 730 r P U 6 76 67 6 67 6 73 / 73/ 562 676 07 0 070 t M 670 670 a 6 6 h e t so 1 1 l l l 1 1 WIPA WPA WIPA WIPA WIPa WIPA WIPA hn t P P P P P P t P I kI G kI G kI G kI G kI G kI G kI G e ed r i acc s da ae B. oh D K K R E l t gf K A K E A o n I D S E in A E A EI ET R. E n o E R E N S NU B B. B at R B R N ii I I B D B LI L O ER U ma E L K K NE T r R I I E I GA R A LA AA AA D U eu A A A I RC T C AC E E E E. L G L Rd I S SI A O. N O AO T R T R B. W TA 1 N L1 L SL SB SBR FI OF I ) 1 (

{ -g Dt 7 7 2 2 5 1 5 2. m y" n N L r bB a A h 5 5 7 7 7 9 7 7 o o1 1 1 8 8 7 7 7 8 e d" o T 0 t t 3 3 2 2 2 2 2 2 ee T t t h e T r at u o i 3 h tf d i i e &2 s ns s t l in a A A a M9 7 7 7 7 7 7 7 7 sa e i 1 6 6 6 6 6 6 6 6 t df r PP i a n c n oi e l A A a d i l e l r h at n N t u" e L I 0 0 5 5 9 5 8 5 t nB h Dt A h t T A 9 9 0 0 8 0 0 0 u am "e 0 0 0 8 8 7 8 7 8 o d o 3 3 2 2 2 2 2 2 h 1 n T h T g et a u r ag e o G l h nd a n g r s N i i I t t k INS 4 4 4 4 4 4 4 4 a a D 3 3 3 3 3 3 3 3 t Fo n A A 5 5 5 5 5 5 5 5 n l i h o a r r f f h i E L t Ao T R s s E n n r o n o o E c ri f 0 ut 1 0 0 0 0 0 07 0 07 I P 07 0 07 07 I 0 0 98 98 98 98 98 98 07 07 07 0 0 0 0 0 0 n t p e 98 98 t eo 6 6 6 6 6 6 6 6 i S W 34 34 34 34 34 34 34 34 a Y .R 7 7 7 7 7 7 7 7 a Re s m xh n R e = t e e r s p A l d pa m M 20 20 20 20 20 20 0 0 1 0 0 0 0 0 0 20 P 0 0 0 0 0 0 0 0 n mh .o 20 34 34 34 34 34 34 34 34 M W 45 45 6 6 6 6 6 6 6 7 7 7 7 7 7 7 7 a o roW c 45 45 45 45 45 45 s 6 s o U Ctkt is l a S y or7 S re6 . 2 l C D a t pyR 1 0 0 0 0 E. A P-320 320 20 20 600 20 600 20 n n ob 7 7 7 0 o 636 636 08 08 864 08 864 08 0 0 0 0 a O F 688 688 78 78 689 78 689 78 C el e I h eh E 1 1 1 1 H L o st dT el i . A E r n a .i l T S n atdi f s n "A t A A e 0 0 o ne 70 70 c I C 94 0 0 94 000 000 000 000 000 000 s ad P 6 6 F c "e g h S 1 2 1 2 f i E B o 1 1 chi R yath t U n lpe ne e L A 0 0 d ph o b I I 75 75 050 050 050 050 050 050 ut A I 63 63 470 470 470 470 470 470 n P 5 5 2 2 S odl l F I 23 23 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 e 3 3 t ai D p ynow 2 e t l i d u sl w P 8 3 i t 3 8 n De a t o F 60( 60 l 275 2 7 5' e yutof A E 730 730 730 730 569 569 605 735 r c n I ) P 070 070 070 070 //2 //2 279 072 a n i gA U 676 676 676 676 73/ 73/ 562 674 g mi -n1 e 070 070 t lh 670 670 a c h r0uP 6 6 e1 m dU t e n e M e. s .E o r o is( r WIPM WPM WIPM WIPM WIPM WIPM WIPM WIPM hn xt t cupu P P P P P P P t e P I kI.G kI G kI G kI G kI G kI G kI G kI G el ah m s e d pl t s ri ,ue acmargpr c p s o uni d aCdiPS ae t B. ohocreaHC l eR K L R E l t ro e D K gf t r ph n L E A n onpoti K A K A D S E in o r sye I A E A M CA EI ET R. E n oC oi E R E N S NU B B. B at e al a l s S O N R B R i t l i a I I B D. B ML LI L O ER U E ei e mas r st r R re e ei c E M L P K MK MK NE T U e u h pi n n GA R A LA G A AA A A L D G L RdT odIi I RC T C AC 0E E E E E. W N TA IS SI R TR TR B. 0 AO N O MO 0 SB SBR FI OF l LL L SL I 1 I ) ) ) 1 2 3 ( ( ( (llliI

TABLE 3 BLOCK LOAD VOLTAGE DIP

SUMMARY

LOAD. ' STARTING % VOLT. % VOLT.. BLOCK kVA PROM CURVE CALCULATED 1 7436 68.1 73.8 2 4325 80.1' 82.9 l 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. l l :! l. t = _ _ _ _ _ _ _ _._

i l 1 l APPENDIX A - SCENARIOS I A.1. LOSS OF COOLANT ACCIDENTS l I 11.0 GENERAL J .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 - was established considering ESF systems response for large line break LOCAs j 2 2 I to.5 ft ), intermediate line break LOCAs (.5 ft2 to.01 ft ) and (14.14 ft2 2 smallline break LOCA'(less than.01 ft ), l i 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 BREAK LOCA The _ LOCA analysis (Reference BAW-10103A, Rev. 3) indicates that for a large break LOCA, the HPI/LP! 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 >4 psig or >30 psig may be assumed to be present to start the HPI, LPI, and. BS pumps. The reactor building cooling fans are automatically transferred to ES operation. 2.1 "B" EDG Failure Case Postulating the failure of the emergency diesel generator 3B leaves the emergency diesel generator 3A to provide AC power to the ESF features required to mitigate the consequences of a large LOCA. j 1 i ca.m.-n.an A.1-1 H-. j

l l \\ l The HPI pump (MUP-1A) 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 HPl 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-1A) is also initiated by ESAS. Since the RCS pressure decreases rapidly below its maximum discharge pressure, the LPI pump will operate at a flow of 3250 gpm, which includes 100 gpm pump recirculation flow, and continue at this flow throughout this 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. EFW will be initiated on loss of the reactor coolant pumps or ESAS. Both EFW pumps (EFP-1 and EFP-2) will start initially and be controlled by EFIC to shore the flow requirements of 430 gpm each, including recirculation. Steau, is available to operate EFP-2 during the depressurization of the OTSG from normal operating pressure to the EFIC reduced fill rate flow setting at 800 psig OTSG pressure. Once secondary pressure is reduced to 800 psig, the fill rate is limited to 2" per minute per generator, resulting in a total required flow of 365 gpm. This flow rato demand will be shared by both EF pumps. However, the entire flow rate could be supplied by EFP-1 with the resultant load being less than that assumed in the diesel loading calculation. 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. Poth HPI 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 I HPI pumps will operate at their maximum flow of 600 gpm and continue to operate until operator action is taken to terminate them. l l l Gelbert CommonweeHh A.1-2

c i l I The LPI pumps (DHP-1 A and DHP-1B) are also initiated by ESAS. Since the. RCS pressure decreases rapidly below their maximum discharge pressure, the 4 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-18) 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-2 was j assumed to fall to start in this analysis. EFP-1 will produce flow of 860 gpm, ) including recirculation while OTSG pressure is high. Upon a decrease of OTSG pressure (800 psig), the flow will decrease to 365 gpm. i 3.0 INTERMEDIATE BREAK As the break size decreases to the smaller break sizes (i.e from.5 ft2 to.01 2 ft ), 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. Because the ESAS loading sequencers which control the HPI and the LPI pumps are actuated by elther 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, j resulting in diesel loadings which are indicative of small and large break l 1 LOCAs. During the initial depressurization, following ESAS actuation by RCS pressure less than 1500 psig, the primary and secondary systems remain i i coupled and EFW flow is required to maintain OTSG levels. For this time j period, the diesel loading is identical to the small break LOCA. As the scenario evolves and additional equipment begins operation, the diesel load Gdbert Commonwealth A.1-3 = - _ - _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ -

approaches that of the large break LOCA. Therefore, the loads at any given time will be bounded by the large break LOCA case. 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 i Postulating the failure of the emergency diesel generator 3B leaves 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 DHP-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. l EFW will be initiated on loss of the 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 of 430 gpm each, including recirculation. Steam is available to operate EFP-2 during the depressurization of the OTSG from normal operating pressure to the EFIC reduced fill rate flow setting at Gdbert Commonwealth A.1-4 w.

.p I h 800 psig OTSG pressure. Once secondary pressure is reduced to 800 psig,'the fill rate. is limited to 2". per minute per generator, resulting in a total required flow of 365 gpm. This flow rate de. mand will be shared by both EF pumps. However, the entire flow rate could~ be supplied by EFP-1 with the . resultant load being less than that assumed in the diesel loading' calculation. 3.2 EFP-2 Failure Case 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 EPP-2 to fill the OTSG's. Both HPI pumps (MUP-1A and MUP-1B) are actuated by ESAS as a result of RCS pressure less than 1500 psig.or RB pressure greater than 4 psig. The 1 HPI 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-1 A 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. EFP-1 will produce a flow of 860 gpm, including recirculation while OTSG pressure is high. Upon a decrease of OTSG pressure (800 psig), the flow will decrease to 365 gpm. l om. .m, l A.1-5 1 i __-______A

4.0 SMALL BREAK LOCA Small breaks of less than.01 ft2 results in the actuation of HPI without a 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. i ' 4.1 "B" EDG Failure Case Postulating the failure of the diesel generator 3B leaves the emergency 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-1A) 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. 51-1170352). 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. 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-i u-co... .. a A.1-6 {

1170352). The RB Spray pump will therefore 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 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 EFP-2 of approximately 1100 gpm which includes 200 gpm pump recirculation flow (Ref. B&W Document No. 51-1170352). EFP-1 is assumed to contribute approximately 550 gpm of this flow requirement throughout this analysis. 4.2 EPP-2 Failure Case Postulating the loss of the Turbine Driven Pump (EFP-2) allows the assumption that two emergency diesel generators are available. 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. .01 ft2 Break Size 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 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 N o. 51-1170352). The HPI pumps were conservatively assumed to 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. Gdbert Commonweahh A.1-7

The LPI pumps (DHP-1A and DHP-18) 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. 51-1170352) The RB Spray pumps (BSP-1A and BSP-1B) will not be j 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 determined by B&W analysis (Ref. B&W Document No. 51-1170352) 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 esse) will operate at a maximum flow of 1070 gpm based on system hydraulic 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 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 two HP! 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. 51-1170352). The HPl pumps were conservatively assumed to 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 (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 Gilbert Commonwealth A.1-8

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. 51-1170352). The RB Spray pumps (BSP-1A 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 determined by B&W analysis (Ref. B&W Document No. 51-1170352) 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. Gdbert Commonwealth A.1-9

1 l,!l Y I s

y;,

LARGE BREAK LOCA 'B' EDG FAILURE CASE ay HPI Pump Flow 600 F 1 3250'- LPI Pu'mp Flow o W R RB Spray Pump Flow a 1600 t e S EFP-1 (gpm) 430 : 1 EFP-2 FAILURE CASE Same as above for HPI, LPI, RB Spray J ') EFP-1 860 5 10 15 20 25 30 35 40 Time iMinutest i Gilbert Commonwealth A. I.10

,'s j s 1 r INTERMEDIATE BREAK LOCA 'B' EDG FAILURE CASE-a; I 600-HPI Pump' Flow LPI Pump Flow 3250 .100 1600 - RB Spray Pump Flow EFP-1 Flow 430 o w R 1 a i e S EFP-2 FAILURE CASE ^ Same as above for HPI, LPi, RB Spray 860 _g. 10 20 30 '40 Time (Minutes) Gdbert Commonwealth A.1 11 a_-_______

g. p."- I' Y \\ %yi: s .Commonweattn A.2-6

>.i-L SLBA INSIDE RB - EFP-2 FAILURE CASE 600 l

295 HPI PUMP FLOW 100 m. 'F LPI PUMP RECIRC. FLOW L-O W 940 R l A EFP-1 FLOW T E. 0 (RB SPRAY NOT INITIATED) s (gpm) 5 10 15 20 25 30 35 40 TIME (Minutes) i i i i 1 1 l Gilbert Commonwealth A.2-7

!-i SLBA OUTSIDE RB -'B' EDG FAILURE CASE k '600 .295 HPI PUMP FLOW 100 F LPI PUMP RECIRC. FLOW L-O W ~665 R A EFP-1 FLOW T E O (RB SPRAY NOT INITIATED) .- s - (gpm) 5 10 15 20 25 30 35 40 TIME (Minutes) l: 1 omrtc... .iin A.2-8 L'

1 1:y.- l 'I 'SLBA OUTSIDE RB - EFP-2 FAILURE CASE '600 4 295' y HPI PUMP FLOW 100-y LPI PUMP RECIRC. FLOW L-O W i 1070 l R-A EFP-1 FLOW T E-0 (RB SPRAY NOT INITIATED) S (gpm) - 5 10 15 20 25 30 35 40 TIME (Minutes) e Gelbert> Commonwealth A.2-9

APPENDIX A - SCENARIOS I 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 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 eccident j and the sequence of events. The scenarios for this analysis are presented below for the two assumed failure cases. 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 loss of offsite power coincident with the FLBA. This results in a reduction in energy input into the RB. Therefore the RB pressure resulting from the FLBA in this scenario was determined to be below 23 psig. (Ref. B&W Document No. 51-1170352). 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. Gilbeft Commonwealth A.3-1

RB Spray will not be initiated because the RB pressure does not reach the 30 psig ES setpoint. Therefore the RB Spray pump, BSP-1 A will not be operating. EFW will be initiated on either low OTSG pressure / level or ES actuation. EFIC will initiate isolation of both OTSGs and control EFW flow to the 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 23 psig. i ES actuation is initiated on the 4 psig RB high pressure setpoint. Both HPl 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 saf'ety relief valve pressure setpoint. j The pumps will continue at this flow rate until operator action is taken to depressurize the RCS. l 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 l pressure is maintained at a greater pressure than the maximum discharge pressure of both LPI pumps, DHP-1A 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. l RB Spray will not be initiated because the RB pressure does not reach the j 30 psig ES setpoint. Therefore, the RB Spray pumps, BSP-1A and BSP-1B, will not be operating. Odbett Comm0ftwealth A.3-2

\\ ' EFW is. Initiated 'on loss of ' RC pumps, low OTSG pressure / level or. ES actuatl'on. -' EFIC will initiate isolation of both OTSG's and control EFW flow-to the unaffected OTSG initially at the maximum fill rate of 330 gpm. Since steaming and EFW recirculation are occurring at the same time an additional ~ y flow of 670 gpm is required to maintain the maximum fill and recirculation rate requirements. : However, the maximum fill rate to a OTSG (for a single OTSG solely available scenario) the flow rate is controlled by EFIC at 740 GPM. Therefore, the total flow from EFP-t (the only pump available in this-failure case) will be 940 gpm including recirculation. EFP-1 is assumed to continue at this rate until the OTSG is refilled. ] ) I k r GilbestCommonwealth A.3-3 = _ _ _ _ _ _ -

FW LINE BREAK ACCIDENT INSIDE RB 'B' EDG FAILURE CASE 295 HPI PUMP FLOW 100 y LPI PUMP RECIRC. FLOW 'L 'O W-500 R A. EFP-1 FLOW T. E O - (RB SPRAY NOT INITIATED) S (gpm) 5-10 15 20 25 30 35 40 TIME (Minutes) l GilbertCommonwealth A.3-4

FW LINE BREAK ACCI' DENT INSIDE RB - EFP-2 FAILURE CASE 295 HPI PUMP FLOW 100 m F LPI PUMP RECIRC. FLOW L O. ~W 'R '940 m A EFP-1 FLOW -T E O (RB SPRAY NOT INITIATED) s . (gpm) 5 10. 15 20 25 .30 35 40 TIME (Minutes) I q l l l Gilbert Commonwealth A.3-5 l

l L APPENDIX A - SCENARIOS A.4 STEAM GENERATOR TUBE FAILURE ACCIDENT i l The basic accident dynamics are described in FSAR Section 14.2.2.2. Since loss of offsite power is assumed coincident with the steam generator tube failure accident, the reactor, the RC pumps, and the turbine are assumed to be tripped at time zero of this scenario analysis. The RB pressure is not affected by this accident since the RCS leak is contained within the OTSG. 1. 'B' EDG Failure The primary and secondary coolant system parameters and sequence of events for 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. The HPI pump (MUP-1 A) is actuated by the Engineered Safeguards Actuation System (ESAS) as a result of RCS pressure below 1500 psig. The HPI pump is conservatively assumed to operate at its maximum flow of 600 gpm to compensate for leakage through the ruptured tube and RCS shrinkage. This maximum flow will continue until operator action is taken to reduce RCS pressure. LPI is initiated by ESAS. The RCS pressure will remain above the maximum discharge pressure of the LPI pump based on the FSAR description. Therefore, there will be no flow into the RCS and DHP-1A will be on recirculation at a flow of 100 gpm. RB Spray will not be initiated because the RB pressure does not change for this accident. Therefore, the RB Spray pump, BSP-1 A will not be operating. EFW will be initiated on loss of RC pumps, low OTSG pressure / level or ES actuation. EFIC will control EFW flow to both OTSGs initially 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 Gdbert Commonwedith A.4-1

1: l-I -e x (470 gpm) is occurring 'at t'he same time, the total flow to both OTSG's would L-beLa total of 800 gpm. The total flow from both EFP-1 and EFP-2 is 1000 (gpm. The total flow includes 200 gpm pump recirculation flow'and assumes the RCS leakage' satisfied the affected OTSGs maximum fill rate. EFP-1 is assumed to contribute 500 gpm flow during the period to refill the OTSG. 2. EFP-2 Failure Case LThe primary and secondary coolant system parameters and sequence of F, events for 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 1 scenario analysis.. HPI Pumps MUP-1A and MUP-1B will be actuated by ES as a result of RCS. pressure below 1500 psig and will operate at an initial total flow of 850.gpm (Ref. B&W-Document No. 51-1170352). -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.ls always greater than the - maximum discharge pressure of both LPI pumps, DHP-1A and DHP-18. I.; 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 OTSGs 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 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 total regired flow from EFP-1 (the only pump available in J GilbertCommonwealth A.4-2 l c____u_ __1_ _ .. J

l-I i f!- this failure' case) wSuld be 1000 gpm. EFP-1 would continue at this flow i until the OTSG is refilled. 1 5 1 Gdbert Commonwealth A.4-3

1, .6 STEAM GENERATOR TUBE FAILURE ACCIDENT 'B' EDG FAILURE CASE <3: .i s ib 600 HPI PUMP. FLOW q 100 p' LPI PUMP RECIRC. FLOW L 'O W 500 R A EFP-1 FLOW 'T E. 0 (RB SPRAY NOT INITIATED) i s (gpm) l 5 10 15 20 25 30 35 40 TIME (Minutes) i-l G%rtTommonwealth A.4-4 l-E: 1

!? STEAM GENERATOR TUBE FAILURE ACCIDENT - EFP-2 FAILURE CASE. 425 y 2-HPI PUMP FLOW 100-p ' LPI PUMP RECIRC. FLOW L O W 1000 R A-EFP-1 FLOW T E O (RB SPRAY NOT INITIATED) S - (gpm) l 5 10 15 20 25 30 35 40 TIME (Minutes) 1 l ( Gilbert,(OmmOnwedittt A.4-5

w f APPENDIX B - DESCRIPTION OF MODIFICATIONS . B.1 : MODIFY EFP-1 AUTO TRIP CIRCUIT AND EDG-3 A MONITORING TIMER - M AR T87-10-19-01 This modification involves the removal of the auto trip of the Motor Driven i - Emergency Feedwater Pump EFP-1 after 30 minutes and its associated timers and alarms. In their place, new mechanical rotary type timers and associated 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, and subsequent alarms are being 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 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-1 A, has been operating above 3000 KW. 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-3A 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 3000 KW, the amount of time operated in the 30-minute rating will be maintained by the mechanical rotary timers such that 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-3A has operated in its 30-minute rating. Gdbert Commonwealth B.1-1

I '.. '; \\ 1 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 ; .g 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-falldown criteria. L I l I l I l 1 _,m._._.,. j a.1-2 J

APPENDIX B - DESCRIPTION OF MODIFICATIONS B.2 = ASV-5/204 POWER SEPARATION - M AR T87-10-09-01 This 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 ES A channel power source. MOV ASV-204 will be controlled automatically by an auxiliary contact from the 'A' channel EFIC logic or manually by a new control switch mounted in the PSA Section of the main control board. Status lights indicating valve position will be located with the new control switch. The power source, cable routing, and wiring of MOV ASV-204 will be ES A Channel. However,10CFR50 Appendix R does not apply as ASV-5 will remain on the ESB Power and meet the present Appendix R cominitments. Therefore, control and indication will not be included at the Remote Shutdown Panel and ASV-204 is not required to comply with 10CFR50 Appendix R separation criteria. Existing annunciator / events recorder alarms will be modified 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 " Turbine Steam Supply Not Ready" to operate in conjunction with ASV-5 alarm logic. e l l l l l Gilbert Commonwealth B.2-1 E

l l I APPENDIX B - DESCRIPTION OF MODIFICATIONS B.3 EDG-1 A EMERGENCY LOAD SHEDDING - HEAT TRACING - M AR T87-10-03-01 This modification 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 Tracine Transformer Power Source HTTR-1 A 480V ES MCC 3A1, Unit 11AR HTTR-2A 480V ES MCC 3A2, Unit 6AL HTTR-3A 480V ES MCC 3A1, Unit 11 AL HTTR-4A 480V ES MCC 3A2, Unit 6AR HTTR-5A 480V ES MCC 3A2, Unit SAR This modication provides for the automatic load shedding of the heat tracing load from the Emergency Diesci Generator 3A following a 480V 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. The contactors providing the isolaition 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 3 Al and the contactors for heat tracing transformers HTTR-2 A, HTTR-4 A, and HTTR-SA 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 pushbutton. The box, components and wiring within are considered safety related and are seismically and environmentally qualified for their respective mounting locations. The existing power feeds downstream of the contactors are considered non-safety related. 1 mne...nma B.3-1

~ APPENDIX-C.-. SAFETY EVALUATIONS - EDG-1A TIMER MODIFICATIOtl SAFETY EVALUATION Shast=1 Of 2 MAR NO. T87.10 19.01 L SAFETY EVALUATION:. Answer the following questions and provide specific justification (use attachment 11 recessary). 1. 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? YES,,_,,, NO X 8ecauses L 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. 8ecauses See Attached Sheet 3. Is the margin of safety, as defined in the basis for any Technical Specification, REDUCED 7 YES NO X 8ecauset See Attached Sheet LICENSE REVtSION REQUIRED: Final Safety Analysis Report YES _X_, NO _ Technical Specification: YES _ NO,g,,, NRC Authorization for Change Required: YES _ 'No L Semi-Annual Reporting to NRC Required: YES NO A 10CFR30.59 CHECKLIST Does the proposed action change the Final Safety Analysas Report or require additional description to be added to the i Final Safety Analysis Report? YES (X) NO( ) Notify Manager, Nuclear Licensmg 1r and Fuels Manarement 3, w Is a Change to the Technical Specifications Required? YES( ) qr NO (X ) w is any unreviewed safety question involved, i.e., (1) is the probability of an occurrence or the consequences of an accident, 1 or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased? YES _,,, NO 1 (2) is the possibility for an accident or malfunction of a dif ferent type than any previously evaluated in the Safety Analysis Report created? YES NO X (3) is the margin 7 safety, as defined in the basis for any Technical ~ Specification reduced? YES,,,, NO _X_, Any answer YES ( ) h All answers NO (X) u Request and receive NRC Authorization for change 1P 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 d Initiate installation of Modification

• Required changes to Technical Specifications Prepared by should be processed in parallel to this checklist.

Name Date

Fl0 a ANALYSIS / CALCULATION g.4EET 2 OF 2 p9 Corporation Crystal River Unit 3 oate OCTOBER 19,1987 REl/ MAR No. MAR T87-10-19-01 Project : EDG-1A MONITORING TIMER ATTACHMENT TO MODIFICATION SAFETY EVALUATION 1. A malfunction of any component in the Elapsed Time Indicator (ETI) or in the 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. FSAR Sections 7.2.4, 8.2.2.6, and 10.2.1.6 have been reviewed. 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. Design Engineer Date Verification Engineer Date Supervisor, Nuclear Engineering Date Rev 7/81 912244 1

f 9: APPENDIX C - SAFETY EVALUATIONS - ASV-204 MODIFICATiott SAFETY EVALUATION A -Sheat 1 Of 3-a-. MAR No. T87. - 10. 09. 01 - SAFETY EVALUATION: Answer the following questions and provide specific justification (use attachment if necessary).- s

1. - Is the probab' ility of an occurrence or the consequence of an accident or malfunction,

sc, of equipment important to safety as previouslygvaluated in the Final Safety Analysis Report, INCREASED? YES _ NO _^ Because: <a -See Attached Sheet

• ~

0

2. 'is the possibility for an accident or malfunction of a diff' erent type than any previously evaluated in the Final Safety Analysis Report, CREATED?. YES _ NO X

1 Because: -See Attached Sheet-Is de margin of safety, as defined in the basis for any Technical Specification, 3. REDUCED? YES __, NO l Because: 1 See Attached Sheet i 1 4 l 't, i LICENSE REY!S!ON' REQUIRED: ' Final Safety Analysis Reports YES No X. a Techrucal Specifiestion: YES ~- NO T ' NRC Authorization for Change Required: YE5[ 'NO 2 .j i Semi. Annual Reporting to NRC Required: YES _ NO 1 i 10CFR$0.59 CHECKLIST m. Does the proposed action enange the Final Safety Analysis Report or require additional description to be added to the { Final Safety Analysis Reoort? YES(. ) 1 NO (X ) i Notsty Manager, Nuclear Licensing 1r and Fuels Manacement I . y-Y is a Change to the Technical 4 Specifications Required? YES: ) -gr NO(X) =- is any unreviewed safety question involved, i.e., (1) Is the probability of an occurrence or the consequences of an accident, ' the Satery Analysis Report increased? YE5_ NOor malfunction of equipmen X (2) is the possibility for an accident or malfunction of a dif ferent type than any previously evaluated in the Safety Analysis Report created? i YES NO X i (3) is the margin of safety, as defined in the basis for any Technical Specification reduced? YES NO X 1 Any answer YES ( ) h All answers NO (X) I y l: Request and receive NRC I i-Authorization for change I s Document ChanEe includmg: (1) Description of change (2) Written Safety Evaluation which provided basis f or items (I), (2). and (3) above. I I i Author ration Received ( ) Description Safety Evaluation Ccmolete h 4 initiate Installation of Modification

• Recuired changes to Technical Specifications f;

) shoul3 De processed m parallel to this checklist. Prepared by U Name Date ~ u_:__ = -

@ Corporation EN,h* ANALYSIS / CALCULATION p SHEET 2 OF 3 a Crystal River Unit 3 1 REl / MAA No.' M AR T87-10-O'9 Date OCTOBER 19,1987 l Project : ASV-5/204 POWER SEPARATION ATTACHMENT TO MODIFICATION SAFETY EVALUATION 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 connected in parallel to a common 250/125 VDC ES "B" channel power and control source. This modification electrically separates ASV-204 from ASV-5 and repowers ASV-204 from 250/125 VDC ES "A" channel power. Also, separate control room controls and separate "A" channel EFIC interlocks are a 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 1 opening ASV-204 whenever the EFIC System calls for emergency feedwater 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 l 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 redundant power channel, and will be provided with its own remote manual control and with separate EFIC interlocks for automatic operation. The type l 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 malfunction of a different type than any previously evaluated in the FSAR is l not created. FSAR Sections 7.2.4, 8.2.2.6, and 10.2.1.6 have been reviewed. I l oesign E ngineer Date verification Engineer Date Supervisor,NuddJr Engineering oate I Rev 7/81 912244

p Flo a po ANALYSIS / CALCULATION SHEET 3 OF 3 corporation Crystal River Unit 3 l REl/ MAR No. MAR T87-10-09-01 Date OCTOBER 19,1987 l ) l Project ; ASV-5/204 POWER SEPARATION i 3. This modification enables the turbine driven Emergency Feedwater Pump l (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 i 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. { l3 oesign Engmeet oate Verification Engineer Date Supervisor, Nucicar Engineering Date I Rev 7/81 912244

,H ' APPENDIX C --SAFETY EVALUATIONS - HEAT: TRACING MODIFICAT10tl 5AFETY EVALUATION - Shest'l of 2 . MAR No. E d-h c 5AFETY EVALUATION:, Answer the following questions and provide specific justification Luse attachment 11 necessary). 4, s;e !... 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.

YE5 NO X Because; ~See Attached Sheet 4 12. 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 1 Because: See Attached Sheet 4 3.. 'Is the margin of safety, as defined in the basis for any Technical Specification, REDUCED 7. YES _ NO 1 Secauset See Attached Sheet LICENSE REVISION REQUIRED: Final Safety Analysis Report: YE5 _ _ NO X Technical Specification: YES NO T NRC Authorization for Change Required:_ YES _ 'NO T Semi-Annual Reporting to NRC Required: - YES,,_, NO x 10CFR50.59 CHECKLIST Does the proposed action change the Final Safety Analysis ~0'4 Report or require additsonal description to be added to the Final Safety Ansivsis Reoort? YES(X) NO( ) Notif y Manaeer, Nuclear L4censsng 9P and Fuels Manacement -9P w is a Change to the Technical Specifications Required? YE54 ) NO (X ) ir I w is any unreviewed safety question involved, i.e., (1) is the probability of an occurrence or the consequences of an accident,

j or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased? - YES _ NO 1 (2) is the possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report created?

j YES No X -~

0) is the margin Esafety, as defined in the basis for any Technical Specification reduced? YE5_ NO1 Any answer YES I )

h All answers NO (X) Request and receive NRC Authorization for change ir Document Change includsng: (1) Description of change ) (2) Written Saf ety Evaluation which provided basis ior items (1), (2), and 0) above. 2 ,c Authorization Received ( ) s Description Safety Evaluation Complete 4 4 Inst ate installation of Modification

• Required changes to Technical Specifications should be processed in parallel to this checklist.

Prepared by tvame Date

9 corporation Florida Power ANALYSIS / CALCULATION SHEET 2 OF 2 Crystal River Unit 3 t R11/ MAR No. MAR T87-10-03-01 Date OCTOBER 5,1987 Project : EGDG-1A EMERGENCY LOAD SHEDDING - HEAT TRACING 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 4 this determination. J 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 l 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 Boration Systems. Technical Specification Sections 3/4.1.2 and 3/4.8.1 were examined. oesign E ngineer oate Venfication E ngineer Date Supervisor, Nuclear ingineermg Date ) l Rev 7/81 912244

f L.' l t) .p. l' i c i ATTACHMENT 3 FPC DIES EL. LOADING EVAL UATION S INCLUDING B UILDING SPRAY ACTUATION SETPOINT EVAL UATION (B &W DOCUMENT NOS. 51-1170352-00 AND 01) f s i i l}}