ML18026A428

PP&L Response to NRC Concerns Re Loss of Spent Fuel Pool Cooling Following Loca,Sses,Units 1 & 2.

ML18026A428
Person / Time
Site:	Susquehanna
Issue date:	08/16/1993
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML18026A429	List: ML18026A428 ML18030A383
References
NUDOCS 9308170357
Download: ML18026A428 (42)

Text

ATTACHMENTTO PLA4012 PP8cL RESPONSE TO NRC CONCERNS REGARDING THE LOSS OF SPENT FUEL POOL COOLING FOLLOWING A LOSS OF COOLANT ACCIDENT SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 DOCKET NOS. 50-387 AND 50-388 9308i70357 9308ih PDR ADOCK 05000387 p PDR

7 y I ATTACHMENTTO PLA4012 TABLE OF CONTENTS 1.0 SPENT FUEL POOL/PLANT CONFIGURATION OVERVIEW ............ Page 1 2.0 TYPICAL REFUELING OUTAGE PRACTICES ....................... Page 2 3.0 ACCIDENT SCENARIOS/CONFIGURATION ....................... Page 4 4.0 PLANT RESPONSE TO WORST CASE ACCIDENT SCENARIO ... .. Page 5 4.1 SFP MAKE-UP/COOLING CAPABILITIES .. Page 5 4.2 REACTOR BUILDING ACCESS.................... .. Page 7 4.3 ENVIRONMENTAL IMPACTS............. . ~ .. Page 9

5.0 CONCLUSION

S.............................. ~........... Page 12 APPENDIX 1: ENHANCEMENTS . Page 13 A1.1 PROCEDURE ENHANCEMENTS............ ~..... . Page 14 A1.1.1 PROCEDURES REVISED: ............. ~..... . Page 14 A1.1,2 PROCEDURES TO BE DEVELOPED OR REVISED .. . Page 16 A1.2 TRAINING ENHANCEMENTS . Page 16 A1.3 MODIFICATION ENHANCEMENTS . Page 17 A1.3.1 SFP LEVEL AND TEMPERATURE . Page 17 A1.3.2 CASK STORAGE PIT DRAIN LINE MODIFICATIONS . Page 17 APPENDIX 2: UHS IMPACT ANALYSIS ............................ Page 18 F IG U RES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ Page 22 FIGURE 1: Fuel Pool Cooling/RHR Interface P&ID FIGURE 2: Reactor Building HVAC Diagram FIGURE 3: Plan View of Refueling Floor

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ATTACHMENTTO PLAR012 1.0 SPENT FUEL POOLtPLANT CONFICURATION OVERVIEW Susquehanna SES is a dual unit BWR 4, with Mark II primary containments. Each unit has a separate reactor building enclosure that surrounds each primary containment structure.

Both units share a common refueling floor that spans the entire top elevation of both reactor buildings. This is a large open room with no intervening walls to separate the units. Both reactor buildings and the common refueling floor, together, comprise secondary containment. The ventilation and SGTS systems to these areas are designed such that the reactor building of the non-LOCA unit is isolated from the LOCA unit and the refueling floor. Consequently, treatment of the airborne radioactivity in the accident unit and the refueling floor is provided by SGTS (See Figure 1).

Figure 2 provides a plan view of the refueling floor and indicates the relationship of the SFPs to the reactor vessels. The SFP for each unit is centrally located between the reactors and share a common cask storage pit. Likewise, the skimmer surge tanks (see Figure 3) for each SFP also communicate directly with the cask storage pit, as well as, the SFP of the respective unit. Each Unit's SFP is equipped with its own non-safety grade SFP cooling and clean-up system. The units also share a common clean-up system. The cooling water for the SFP cooling system is provided by the service water system, which is non-safety grade and not Diesel Generator backed. Each SFP is equipped with separate connections to each loop of Emergency Service Water (ESW) to provide a source of safety grade make-up water. Each of the two redundant ESW connections to each SFP can provide sufficient water to maintain water level under boiling conditions. Due to the communication of the cask storage pit with the skimmer surge tanks, adding water to one unit's SFP to maintain the water level above the weir will result in water level in the opposite unit's SFP to rise above the weir.'ach SFP is provided with a connection to the RHR system of the respective unit as a means of providing a safety grade backup cooling system. The connection to the RHR system utilizes Seismic Category I piping and any one of the four RHR pumps can be used in this mode. Each of the four RHR pumps that could be used receives power from a separate emergency diesel generator.'URING OUTAGE PERIODS I MAKE UP TO THE SFP CAN ALSO BE PROVIDED VIA ONE OF THE TWO (2 ) LOOPS OF CORE SPRAY . PLANT ADMINISTRATIVE PROCEDURES REQUIRE AT LEAST ONE LOOP OF CORE SPRAY TO BE OPERABLE AT ALL TIMES WHILE THE SFP GATES TO THE REACTOR ARE OPEN.

THERE ARE FOUR COMMON EMERGENCY DIESELS FOR THE STATION AND ONE SPARE.

ON EACH UNITI EACH RHR PUMP AND CORE SPRAY PUMP IS POWERED BY A SEPARATE DIESELI I ~ E ~ I THE A RHR PUMP FROM THE A DIESELI THE B RHR PUMP FROM THE B DIESEL ETC.

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ATTACHMENTTO PLA4012 2.0 TYPICAL REFUELINC OUTACE PRACTICES Prior to the start of an outage, the heat load in the isolated fuel pools is small and the time to boil is relatively large. The time to boil for the outage unit's pool (last outage was 18 months ago) is approximately 136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br /> or 5.7 days.'he non-outage unit's pool underwent the most recent offload (i.e., 6 months ago) and has a time to boil of approximately 81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br /> or 3.4 days.'"'he full core is offloaded during SSES refueling outages. The core heat load decays as a function of time during the outage. For a typical SSES full core offload, the decay heat will decrease by approximately 28% from the time the core is placed in the fuel pool until it is returned to the reactor and the cask storage pit gates are re-installed.4 The emergency heat load (EHL) of 33.9 million BTU/hour is the maximum design heat load which can be projected to possibly be resident in one SSES fuel pool.'he EHL can be dissipated by the RHR fuel pool cooling mode at 5,700 gpm maintaining the fuel pool temperature at less than 125'F. The current (Unit 1, 7th refueling outage) fuel pool heat loads (this includes the heat of both pools combined) is =28 million BTU/hour, which is less than the EHL. This heat load is calculated based on actual SSES operating history.

In a typical refueling outage, fuel offloading starts on day 6 (6 days after shutdown) and is completed by day 13. During the offloading period, RHR is operated in the shutdown cooling mode. Once the offload is complete, the cask storage pit gates are removed thereby connecting the outage unit's fuel pool and reactor cavity to the operating unit's fuel pool.

This increases the outage unit's effective "pool" volume by approximately 33 %. This pool configuration (2 fuel pools + cask storage pit + reactor well + equipment pit) is typically maintained until RPV reload is completed on day 37 of the outage.

The RHR shutdown cooling mode is typically operated until fuel offload is completed (day

'13), and is available for operation until work on the common portions of RHR is initiated.

For the Unit 1 7th refueling outage, work on the common portion of the RHR system is currently scheduled to begin on day 16 and be completed by day 26; however, the beginning and end of this work window is devoted to testing so that the system could easily THESE VALUES ARE BASED ON THE SSES U26RIO DATA SCHEDULED FOR THE SPRING OF 1994 AND ARE CONSERVATIVELY CALCULATED. ACTUAL TIME TO BOIL WOULD BE LONGER.

IT SHOULD BE NOTED THAT 2/3 OF THE PREVIOUS CYCLE CORE IS RETURNED TO THE REACTOR WHILE 1/3 WILL REMAIN IN THE POOL FOR STORAGE.

THIS HEAT LOAD CONSIDERS A FULL CORE OFFLOAD I A FULL FUEL POOL I AND POWER UPRATE .

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ATTACHMENTTO PLA4012 be returned to service during this period. The other loop of RHR, which is out of service, is currently scheduled to be returned to service on.day 26, however it is available for"testing on day 23 and could be put into service in an emergency situation. Therefore, the effective "window" when both loops of RHR would be out of service due to maintenance activities is 7 days. However, the work being performed in this period is typically individual tasks on the order of 8 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> durations rather than a 5 day activity.

Between day 16 and day 26 (when RHR is not available), the time to boil typically ranges from 40 to 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> (assuming 110'F fuel pool temperature).'his time exists even though a relatively large heat load is present in the pool because of the large volume of water created by the connected pools. Therefore, it would be possible to restore a loop of

'RHR for use in the fuel pool cooling mode within the 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> time to boil time frame.

Also, during the time period when the pools are connected, the outage unit's fuel pool cooling system is secured on day 17 and is taken out of service on day 21. The system is maintained available for several days to assure the operating unit's FPC system can adequately dissipate the fuel pool heat load. Thus, the operating unit's FPC system is cooling both fuel pools from day 17 to day 30. On day 30, RHR is returned to service in the shutdown cooling mode as refueling begins.

Typically, on day 11 (prior to completion of the fuel offload), the outage unit's HVAC system is isolated from the recirculation plenum. This is done to assure that the outage unit's environment is not affected by operation of the HVAC recirculation system. This prevents the environment of the refueling floor (zone III) and/or the operating unit's reactor building from being mixed with the outage unit in the event that a radiological release occurs on the refueling floor or the operating unit. On day 48, the outage unit is typically.

reconnected to the recirculation plenum. Work activities may require reconnection of the outage unit to the recirculation plenum during the work activity window.

After the fuel pools are returned to an isolated condition (day 39), the outage unit's fuel pool time to boil is approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, even though the pool volume is significantly smaller than it was during the outage. This is because the resident fuel bundles have had time to decay and only 1/3 of the core remains in the SFP. The operating unit's pool has a time to boil of =130 hours or 5~/2 days.

ACTUAL TIME To BOIL IS EXPECTED TO BE CONSIDERABLY LONGER DUE TO THE CONSERVATIVE ASSUMPTIONS USED IN THE CALCULATIONS.

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ATTACHMENTTO PLAA012 3.0 ACCIDENT SCENARIOS/CONFICURATION PP8 L's evaluation of the concerns associated with a loss of SFP cooling following a LOCA or LOOP event indicates that cooling can be restored to the SFP(s) prior to boiling.

Adequate time will exist for either restoration of the normal SFP cooling system or utilizing Seismic Class I connections to the safety-related Residual Heat Removal (RHR) system.

Procedural guidance for responding to a loss of SFP cooling is contained in Off-Normal and Emergency Operating procedures. The latter procedure specifically identifies the need to provide for SFP cooling in a LOCA situation, while the Off-Normal procedure addresses a loss of cooling for any reason.

The concerns raised regarding a loss of SFP cooling are associated with a Design Basis Accident (DBA) LOCA or a DBA LOCA concurrent with an extended loss of offsite power (LOOP). In a LOCA condition, the normal SFP cooling system will be automatically shed from the plant electrical system, along with other non-safety related equipment, to permit the startup of the large ECCS pumps on the LOCA unit. Once all of the ECCS equipment has started, the SFP cooling system and its support systems can be manually restarted. A similar manual (non-mandatory) load shed exists in emergency response procedures with regard to reactor building post accident temperature. However, as with the LOCA load shed, the SFP cooling system and its support systems could be manually restarted if they are manually shutdown. If this cannot be accomplished, the safety related RHR system could be used to cool the SFP in its Fuel Pool Cooling mode of operation. It should be noted that this mode of RHR does not perform a "safety-related function", however, it uses the same components as the safety-related functions of RHR and the portion of SFP cooling that it utilizes is ASME Class 3 and Seismic Category I.

In a LOOP, normal fuel pool cooling would be lost for a period of time until power could be restored. The SSES IPE (using NUREC-1032 criteria) indicates that there is a 99.53%

probability of restoring offsite power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based on these factors, it is reasonable to expect to restore the normal SFP cooling system prior to the SFP reaching a boiling condition. In the extremely unlikely event that the LOOP is of a longer duration, the safety-related and emergency diesel generator backed RHR system could be used to cool the SFP in its Fuel Pool Cooling mode of operation. Consequently, a LOCA/extended LOOP would be the most limiting event since it would result in the loss of the normal SFP cooling for the duration of the event.

A LOCA/extended LOOP could be postulated to occur while both units are operating or while one unit is in a refueling outage. The most limiting scenario is a LOCA/extended LOOP where both units are operating, since ECCS would be required to cool both reactor cores and SFPs (for isolated pools) at the same time. However, it also 'represents a condition where the lowest heat loads, and thus longest time to boil exists for the SFP. The outage situation represents a greater heat load in the SFP,'ut places less demands on ECCS due to the cross-tied pools and one unit shutdown.

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ATTACHMENT,TO PLA4012 Consequently, the design of Susquehanna is such that SFP boiling would not occur while one unit is in a refueling outage should a LOCA occur on the operating unit or with both units operating and cross-tied SFPs. This statement applies even if a DBA LOCA with a Reg Guide 1.3 source term is assumed concurrent with an extended LOOP on both units. This is due to the fact that the reactor cavity and both SFPs are cross-tied; and the outage unit's reactor building will remain accessible for the duration of a LOCA, since its HVAC system is normally isolated from the operating unit and the refueling floor. Crosstying the SFPs provides for a water volume of 173;146 ft'ompared to a volume of 48,690 ft'or a single isolated SFP. Therefore, even though the heat load is relatively large, the time to boil is always greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> and is usually greater than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. The outage also provides the greatest amount of make-up capability (ESW from both units and Core Spray from the outage unit). These factors allow sufficient time for operator action to restore cooling to the SFP. The cross-connecting of the SFPs allows either unit's cooling systems (i.e., normal Fuel Pool Cooling and RHR Fuel Pool Cooling mode) to provide cooling to both pools. This is accomplished by utilizing off-normal procedure ON-1 35(235)-001, "LOSS OF FUEL POOL COOLING/COOLANT INVENTORY" and operating procedure OP-149(249)-003, "RHR OPERATION IN FUEL POOL COOLING MODE". These procedures provide the detailed guidance in terms of ensuring proper SFP level and ESW and RHR flow rates.

Therefore, during an outage, SFP boiling will not occur and the remainder of this response will focus on non-outage conditions with isolated SFPs.

4.0 PLANT RESPONSE TO WORST CASE ACCIDENT SCENARIO 4.1 SFP MAKE-UP/COOLING CAPABILITIES The engineered safety grade SFP make-up system for Susquehanna is Emergency Service Water (ESW). The capability also exists to use the safety-related RHR Service Water (RHRSW) system via the RHR system as a back-up source of safety grade make-up (not proceduralized at this time) The actions required to respond to a Loss of Fuel Pool Cooling

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event are described in procedures which existed at the time the concern was raised. The use of ESW involves the manipulation of three manual valves (2") per loop of ESW in two different areas of the plant.

As noted earlier, initiation of ESW make-up to either SFP will result in both pools being filled when the pool height is raised above the weir regardless of whether or not the pools are cross-tied. With the pools isolated, water added to one pool will overflow to its skimmer surge tank; which when completely filled will overflow to the cask storage pit; which will overflow to the opposite units'kimmer surge tank; which will in turn overflow to the opposite units SFP (see Figure 3). Instructions exist in the procedures to fill above the weirs, thereby completely filling the skimmer surge tank, and then using the existing Page 5

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ATTACHMENTTO PLA4012 skimmer surge tank level instrumentation to determine that the pools are filled to at least the weir height. This instrumentation has 1E power and is seismically designed in accordance with GDC 63 and section 9.1.3 of NUREG-0800. This instrumentation is not envirorimentally qualified since it does not perform a function identified in section 7.1 of NUREG-0800 or Reg Guide 1.97.

Level measurement using this instrumentation is accomplished by opening a drain valve on the tank level instrumentation and observing whether there is a change in skimmer surge tank level. If none occurs, then the pools are filled above the weirs and in direct communication with the skimmer surge tank. If a change in level occurs, then the pool level is below the weir elevation and the operator would add water via ESW or other systems until the tank level is increased. This process of periodically adding water to compensate for boil-off would then be repeated over the course of the event until cooling is re-established. The above 'referenced procedures caution the operators against overfilling the SFP through uncontrolled make-up to the SFP. Modifications to add level and temperature indication in the control room for each SFP are being planned in order to minimize the actions taken by the operator and provide for enhanced assessment of SFP heat-up.

Procedures also identify the RHR Fuel Pool Cooling mode as a means of cooling the SFP, utilizing safety-related and Seismic Category I equipment, in the event the normal fuel pool cooling system cannot be restored. This is accomplished by the manipulation of six valves in two different areas of the plant. These valves range in size from 8" to 16" and are all ASME Section III manual valves. These valves are not required to be tested under ASME Section XI, however, PP8 L will be performing periodic stroking of these valves. Either loop of RHR may be used in this mode and the installation of a spool piece is not required, since the connection at SSES is completely "hardpiped".

Pre-operational test data from both units indicates that at least 5,700 gpm can be achieved by one pump through one loop of RHR. Calculations indicate that this flow can be sustained with more than adequate NPSH for the RHR pumps once the SFP has been pre-filled to the proper'level. This level is specified in procedures and provides sufficient level in the skimmer surge tank to ensure adequate head for the RHR pumps. This flow is sufficient to maintain temperatures 6 125 F at decay heat loads up to the Emergency Heat Load.

Ultimate Heat Sink (UHS) analyses performed to date indicate that this heat load can be accommodated. The design basis event for the UHS is a LOCA/LOOP with a LOCA on one unit and a simultaneous forced shutdown on the other unit. In this mode, the UHS can provide sufficient cooling water for safety related equipment for 30 days with the water temperature at or below the system's design cooling water temperature. The reactor decay heat assumed in the UHS design analyses ranges from approximately 140 MBTU/hr after 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, to 36.6 MBTU/hr after day 7, to 22.6 MBTU/hr on day 30 of the event. This decay heat data is used for both the LOCA and the non-LOCA unit.

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ATTACHMENTTO PLA4012 Three shutdown scenarios were evaluated to determine the impact that the operation of the RHR fuel pool cooling mode could have on the UHS minimum heat transfer (MHT) and maximum water loss (MWL) design basis analyses. The MHT analysis assumes conditions which result in the highest potential for pond heat-up. The MWL analysis evaluates worst case conditions for pond inventory losses. It is concluded that for all three cases evaluated, the existing design basis UHS MHT and MWL analyses are bounding. Appendix 2 provides specific details of the scenarios and the evaluation.

It is estimated that it will take approximately eight hours to place RHR in service in the FPC mode of operation, excluding the time to fill the SFP(s) to the appropriate level. The time to fill is dependent on the pool configuration and system availability, but is within the time to boil. It is estimated that make-up can be started within one hour of initiating action to begin the process. Assuming only one loop of ESW is available for makeup (35 GPM per pool) and pool level is at the weir height when makeup is initiated, the maximum time to reach the appropriate SFP level to sustain RHR system operation is 22.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This maximum fill time would exist when the pools are isolated and the cask pit is empty. With the fuel pools connected through the cask pit, the fill time is reduced to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (cask pit would be full) In an outage situation with both pools connected to the reactor well and

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equipment pit, the fill time would be 8.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. In the more likely situation where both loops of ESW are available to provide SFP make-up, fill times would'be halved. Fill time would be further shortened by use of other available makeup sources, such as condensate transfer and fire water systems. Procedure enhancements have been made as a result of the Loss of Fuel Pool Cooling issues. These procedure enhancements are also incorporated into the operator re-qualification training program at SSES. A detailed description of the enhancements is provided in Appendix 1.

4.2 REACTOR BUILDING ACCESS FSAR Chapter 18 documents PP8 L's evaluation with regard to operator access to the reactor building to perform actions required to respond to an accident like TMI where substantial core damage resulted. This evaluation indicated that there were no operator actions in the reactor building required to respond to this type of accident. The FSAR evaluation also indicated that no airborne dose contribution was assumed in evaluating operator access, based on PP8 L's interpretation of NUREG-0737. The contained source dose for each room was based on the contact dose from a point source. PP8L has subsequently performed access evaluations for specific actions which did consider an airborne dose contribution on a case by case basis. An airborne dose contribution has been included in the evaluations of operator access that have been performed in response to this issue.

For a severe accident, only one SFP would be expected to boil. This is primarily due to the fact that the non-accident unit's reactor building can be isolated from the accident environment by manually closing (from the'ain control room) remote actuated safety-Page 7

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ATTACHMENTTO PLA4012 related dampers. These dampers open when a LOCA and LOOP signal is received, thereby aligning all three ventilation zones to SGTS (see Figure 1). When a LOCA signal alone is received, only the dampers on the LOCA unit open (permitting the LOCA unit to be connected to Zone III), thereby preventing contamination of the non-LOCA unit. Manually closing the dampers on a LOCA/LOOP, with high radiation in the accident unit, will afford the same barrier against the spread of contamination to the non-accident unit as is afforded under LOCA only conditions. Three zone mixing is automatically initiated when the LOOP condition is present because of the loss of normal HVAC systems to both reactor buildings.

Since the recirculation airflow would be cut off when the dampers are closed, PP8 L has evaluated the temperature rise that the non-LOCA unit would experience during the loss of recirculation flow. This evaluation shows that temperatures willremain within design values for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This provides ample time for the non-accident unit to reach cold shutdown (time to reach cold shutdown is typically 6-12 hours).

Isolation of the non-LOCA unit from the LOCA unit environment would permit access to restore normal Fuel Pool Cooling, if offsite power is available, or place RHR Fuel Pool Cooling mode in operation should offsite power continue to be unavailable, This enables the operators to prevent SFP boiling on the non-LOCA unit. Should a severe accident occur with the accident conditions described in Chapter 18 of the FSAR on Unit 1, the ESW valves located on El. 670'ould be inaccessible for the duration of the response to the LOCA (30 days per Ultimate Heat Sink design basis). The primary reason for the inaccessibility is the conservative assumption of assigning the peak contact dose of the most radioactive source to the entire room. The remaining ESW valves on Unit 1 (El. 749') would be accessible after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If the FSAR Chapter 18 LOCA occurs on Unit 2, all ESW valves are accessible after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.'ccess to the RHR FPC valves located on El. 704'f either Unit 1 or 2 would be precluded for a FSAR Chapter 18 LOCA. The RHR FPC valves located on EL. 749're in the same location as the ESW valves and similarly would be accessible after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> for an FSAR Chapter 18 LOCA. In this case, the time to boil will be in excess of 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />, thereby allowing ample time for the operators to take the actions described in Section 6.1 above.

PP&L has performed calculations to determine realistic dose that an operator would receive accessing the RHR and ESW valves using a Reg. Guide 1,3 source term from airborne and contained sources, but taking into account distance and time factors. For RHR, the amount of piping containing the radioactive source is so, large that significant distance reductions are not possible. The time involved to open the large valves also results in unfavorable stay times. Consequently, a realistic Reg. Guide 1.3 dose evaluation was not performed for the RHR valves. For ESW, distance reductions were possible and are based on the proximity of the valves to the Core Spray piping which acts as the radiation source. Also reductions THE ESW VALVES THAT ARE LOCATED ON EL. 670 'N UNIT 1 ARE LOCATED ON EL 683 IN UNIT 2.

~ THIS ACCOUNTS FOR THE DIFFERENCE BETWEEN UNITS WITH REGARD TO FSAR CHAPTER 1 8 ACCESSIBILITY~

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ATTACHMENTTO PLA4012 in stay time were possible based on a simulated walkdown by an operator. The dose an operator would experience accessing the ESW valves on EL. 670'n the Unit 1 reactor building 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a Reg Guide 1.3 source term and realistic assumptions is 4.22 Rem.

The realistic assumptions are distance of the valves from the source and a walkdown transit time inside the reactor building of 1.8 minutes to and from the valves and a time of 1 minute to open each of the two valves (2 minutes total for valve opening). Similarly, the dose to an operator opening the 4 ESW valves (1 minute each valve) on EL. 749's less than 1.6 rem.

4.3 ENVIRONMENTAL IMPACTS PPRL is performing an assessment of the environmental effects resulting from a boiling SFP assuming operator access to the LOCA unit is not possible. These effects include flooding and the elevated heat load/humidity imposed upon equipment in the reactor building due to the boiloff. The determination of the environmental effects is based upon taking operator actions to mitigate the effects of a boiling SFP, in the unlikely event cooling cannot be restored to the SFP. The principal operator action to limit the environmental impact of a boiling SFP is shutting down the reactor building recirculation system. Turning off the recirculation system will prevent the spread of the boiling environment to the accident unit's reactor building.'lso, preliminary evaluations indicate that turning off the recirculation system will still permit the SGTS system to maintain a negative pressure in secondary containment since the mass of steam added by the boiling pool is less than the capability of the SGTS system. It should be noted that the primary purpose of the reactor recirculation system is to mix the post-accident environment so as to provide a uniform concentration of airborne radiation within the reactor building. While procedures have not been revised yet to reflect shutting down the recirculation system, the need to mitigate the environment from a boiling pool is recognized in emergency plan procedures, thereby allowing the emergency organization to take this action. EP-PS-102, "Technical Support Coordinator: Emergency Plan-Position Specific Procedure" provides specific instructions to the Technical Support Coordinator to monitor the SFP and take actions to preclude boiling. The Technical Support Coordinator is a pager-activated position under PPBL's Emergency Plan and would be activated to respond in the event of an accident.

The steam from the boiling pool is expected to condense on the structural surfaces of the refueling floor elevation, with a large portion condensing out in the form of rain without direct contact with these surfaces. A portion of the boiloff (=10%) will be removed via the SGTS system and the remaining 100% relative humidity air will condense throughout the THESE ENHANCEMENTS HAVE NOT BEEN PROCEDURALIZED AS OF THE DATE OF THIS REPORT. THEY ARE PLANNED TO BE INCORPORATED INTO EP-PS-102 .

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ATTACHMENTTO PLAP012 refueling floor.9 This condensate will find its way to the reactor building sump room from the refueling floor via the drainage system. Since the refueling floor is a common area, the drains on the Unit 1 side will drain to Unit 1 and Unit 2 dr'ains to Unit 2. Therefore, the condensate would be equally split between the units. Based on this, an assessment was performed to determine the extent of flooding that would result from a boiling SFP.

For the flooding assessment, a decay heat corresponding to an isolated SFP at the end of an outage (day 55) was chosen. This results in a boiloff rate equivalent to =11 gpm for the isolated SFP. Use of the boiloff rate (rather than the maximum makeup capability of ESW) is appropriate since off-normal procedures direct the operators to periodically fill the SFP to offset boiloff rather'than leaving the valves open continuously. The design of the Ultimate Heat Sink, from which the ESW system takes suction, includes sufficient volume for ESW to provide makeup to the SFP for a period of 30 days at 35 gpm. Using this as the duration of the LOCA, and the boiloff rate from the HVAC analysis, the total volume of water that would boiloff, assuming no operator action to restore cooling within 30 days is:

11 gal/min x 30 days x 60 min/hr x 24 hr/day = 475,200 gallons Moisture extracted by SCTS (=10%) = - 47 520 gallons TOTAL = 427,680 gallons This volume would then be split equally between the two units, as described above, such that approximately 213,840 gallons would accumulate in the basement of each reactor building. If offsite power is unavailable for the sump pumps, the condensate would overflow the sump and start to accumulate on the floor. Water tight doors rated for a 23 foot head of water isolate the Sump room from the ECCS rooms except for an adjacent Core Spray Pump Room." Therefore, water would begin to accumulate in these two rooms at the same rate. The other Core Spray room is protected by water tight doors and therefore is not subject to flooding. It should be noted that only one loop of Core Spray is required to assist in long term core cooling at this point in the LOCA, so that the flooding of one Core Spray room does not represent a threat to long term core cooling. The volume of the Sump room/Core Spray Pump room combination is such that water could accumulate to a depth of 11 feet within for 30 days at the 11 gpm boil-off rate. This is well within the 23 design limit of the doors. Therefore, for the isolated pool case, the Sump Room/Core 'oot THE SGTS -HEATERS ARE DESIGNED FOR 1 0 0 ~o HUMIDITY AIR AND ARE EQUIPPED WITH DEMISTERS s HEATERS AND DRAINS TO REMOVE THE MOISTURE ~

THE AFFECTED CORE SPRAY PUMP ROOM IS THE "A" PUMP ROOMs WHICH CONTAINS THE "A" & "C" CORE SPRAY PUMPS .

Page 10

ATTACHMENTTO PLA4012 Spray Pump Room combination provides adequate volume for accumulation of boiloff over a 30 day accident event duration.

A preliminary assessment of the temperature impact to the reactor building under these conditions indicates that Environmental Qualification (EQ) temperature limits are exceeded for a small number of safety related equipment in the unlikely event of pool boiling, without actions to restore cooling to the SFP. The assessment indicates that when the reactor building HVAC recirculation system is shutdown at the onset of boiling, the temperatures within the accident unit's reactor building will remain within EQ limits for the major ECCS components. The peak temperature in the ECCS switchgear and load center rooms will reach a maximum temperature of =105 at 30 days into a LOCA/extended LOOP. This temperature slightly exceeds the EQ limit of 104'F, but is well within the capabilities of the equipment to perform its function for the brief period of time it will be operating above the EQ temperature. The room temperatures for 15 additional instruments/MCCs/panels will be more significantly exceeded at 30 days. However, an evaluation of the impact of the higher temperatures indicates that the LOCA qualification life would be reduced from 100 to 46 days for the most temperature limiting device." A review of the safety function of these components indicates that they perform their functions very early in the LOCA event and would not be required to operate at the time their EQ temperatures would be exceeded.

Even if the components were to fail, only the failure in three 480 VAC MCC's would result in any ECCS functions being affected. The failure of components within these MCC's could cause the "C" RHR and Core Spray pumps to be lost; however, at the time in the event when this would happen (day 20 - 55), there would still be sufficient RHR and Core Spray pumps available to cool the reactor core.

In summary, the capabilities described above in combination with the procedures discussed in this report provide ample assurance that the consequences of a postulated boiling SFP will be mitigated. Consequently, adequate long term reactor core cooling can be maintained, along with SFP cooling and makeup to a boiling SFP.

THE REDUCTION IN LIFE WOULD OCCUR AFTER THE EQ TEMPERATURE IS EXCEEDED WHICH WOULD BE AT LEAST 20 DAYS AFTER THE SFP BEGINS TO BOIL.

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ATTACHMENTTO PLA4012

5.0 CONCLUSION

S PP&L has thoroughly evaluated the concerns associated with the SFPC system at SSES and have determined that safe operation is assured. The original design of the systems, structures, and components at SSES adequately addresses a loss of SFP Cooling concurrent with LOCA and/or LOOP events. In addition, adequate response capability exists to prevent SFP boiling at SSES. These conclusions are based upon the following:

1. Based on the SSES IPE, the probability of a DBA LOCA with an extended LOOP and a

, boiling SFP is on the order of 10".

2. Time to boil is typically much greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> at any point in which the accident scenario is of concern. Sufficient time will be available to take appropriate operator actions to restore SFP cooling and prevent boiling.

3. ESW makeup to the SFPs will eventually fill both pools, therefore access to the accident unit is not required to provide makeup water to its SFP.

4. Cooling can be restored prior to SFP boiling by either:

a) restoration of normal SFPC system, or b) the safety grade RHR system (FPC mode).

5. Procedures have been improved and operators have been trained to assure the proper response to a loss of SFP cooling. Modifications to install improved level and temperature indication to the SFPs will further strengthen the operator response to this event.

6. Access to the non-LOCA unit is always assured due to the capability to isolate its ventilation system from the accident unit from the main control room.

7. Modifications are being evaluated to support removal of the cask storage pit gates during non-outage periods will provide capability to utilize the non-LOCA unit SFPC system, if available, or the RHR FPCA mode during all accident conditions to prevent both SFPs from reaching a boiling condition."

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ATTACHMENTTO PLA<012 APPENDIX 1: ENHANCEMENTS Page 13

I ~ P ATTACHMENTTO PLA4012 As a result of the engineering evaluations, PP8 L has made enhancements and is considering further improvements in the area of modifications. Consideration is being given to removing the cask storage pit gates at SSES to provide additional flexibility in responding to a loss of SFP Cooling event. In addition, procedures'ave been updated to improve operator awareness of the concerns, and operator training has been provided to enhance response to a loss of SFP Cooling event. The SFP monitoring equipment meets licensing requirements; however, enhancements are being planned in order to improve operator response to such events. A detailed discussion of the various enhancements follows.

A1.1 PROCEDURE ENHANCEMENTS A1.1.1 PROCEDURES REVISED:

1. ON-135(235)-001, Rev. 12, "LOSS OF FUEL POOL COOLINC/COOLANT INVENTORY ~ (DATED 1 2/22/92)

This revision identified:

That the pools can be crosstied should pool cooling not be able to be reestablished. (This action increases the time to boil and makes either available to cool or makeup to both pools.)

units'ystems That fire protection water can be used to makeup to the pools. (Provides an additional source of makeup water to those already identified in the procedure.)

Alternate methods for determining fuel pool level should access to the refueling floor be restricted. (Allows level determination via use of the skimmer surge tank instrumentation should normal level detection methods not be available.)

Cautions should pool boiling occur with the ventilation systems for zone I, II, III isolated regarding increased airborne radiation, the potential need to evacuate the refueling floor and potential impacts to safety related equipment in the reactor buildings. (Draws specific attention to the potential impacts of a boiling pool.)

The need to maintain pool level, notify HP and monitor building releases should pool boiling occur.

All available means should be employed to cool the pools prior to allowing the pools to boil. (Increased the emphasis to not allow the pools to boil.)

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ATTACMMENTTO PLA4012

2. ON-104(204)-001, U¹1: PCAF 1-92-1390 to Rev. 7 & U¹2: PCAF 2-92-0605 to Rev.,

7~ RESPONSE TO LOSS OF OFFSITE POWER I DATED 1 2/1 7/92 This change provides instructions upon loss of offsite power that the fuel pool cooling system will trip and that ON-135(235)-001 "LOSS OF FUEL POOL COOLING/COOLANT INVENTORY" (described above) shall be performed.

3 OP 1 49(249) 003~ RH R OPERATION IN FUEL POOL COOLING MODE U¹1: PCAF 1-92-1389 to Rev. 9 R U¹2: PCAF 2-92-0604 to Rev. 9, Date: 12/1 5/92:

That should an RHR flow of 6000 gpm be required, that controlled step increases to 6000 gpm should be achieved allowing stabilized conditions to be reached at each step increase. (This assures pool level does not decrease due to pool cooling such that operation at the higher flow rates is not adversely impacted.)

Rev. 11 for U¹1 5 U¹2, Date: 4/21/93:

The need to fill the fuel pool to an appropriate level of 8 inches below the fuel pool curb to obtain RHR flow of approximately 6000 gpm. It also identified that fuel pool filI methods are delineated in ON-135(235)-001, (DATED 12/2/92),

"LOSS OF FUEL POOL COOLING/COOLANT INVENTORY". Filling the fuel pools to an appropriate level to assure proper NPSH for the RHR fuel pool cooling mode of operation is based on the pre-operational testing performed on the RHR fuel pool cooling mode (ref. pre-operational test reports P49.1 and P249.1) and is further confirmed by PP8 L calculation M-RHR-039 Rev. 0 issued 5/17/93. The specified height provides an adequate head of water to assure NPSH by keeping the skimmer surge tank full during RHR operation. Thus adequate NPSH is assured when the fuel pool height is adequately maintained and monitored.

A caution to closely monitor fuel pool level during filling as the addition of cold water could cause a fuel pool level drop.

A caution to closely monitor fuel pool level when RHR is placed in service as the cooling of the pool could cause the level to decrease.

4. EP-PS-102, Rev. 6, "TECHNICAL SUPPORT COORDINATOR: Emergency Plan-Position Specific Procedure", DATE 12/30/92 This change added a section discussing the concerns associated with loss of FPC events. The added section provides instructions to determine when pool boiling can be expected and the actions that can be taken to prevent or mitigate the consequences of fuel pool boiling. Included are cycle specific time to boil curves.

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t ATTACHMENTTO PLA4012 A1.1.2 PROCEDURES TO BE DEVELOPED OR REVISED

1. OUTAGE PLANNING:

Enhancements to the SSES outage planning process will be made to provide for formal engineering evaluation of the SSES outage schedule for conformance to the SSES design basis. Administrative controls to Nuclear Department Administrative Procedures and functional unit procedures, as necessary, will be made.

The SSES Unit 1 fall 1993 refueling outage evaluation is complete and is currently undergoing review.

2. VALVE STROKING PROCEDURES:

A. Procedures will be developed to perform valve stroke testing on the manually operated RHR Fuel Pool Cooling valves, These procedures will be implemented during the next set of unit outages.

B. Procedures will be implemented to test the ESW (as part of ISI program testing on 2 yr intervals) makeup valves. Procedures are issued for Unit 1 (and implemented) and procedures for Unit 2 will be developed and are planned to be performed by the end of the year.

3. REVISION TO EP-PS-102 FOR HVAC ACTIONS This change would provide direction to shutdown the Reactor Building Recirculation fans if SFP boiling appears imminent and it will not be possible to restore cooling to the SFP because of accessibility to the reactor building or other reasons. It also will provide direction for the isolation of the non-accident unit from the recirculation plenum to prevent the spread of airborne radiation.

A1.2 TRAINING ENHANCEMENTS SSES operators have been informed of the issues and procedure changes through "hot box" training and through the quarterly Operations re-qualification training program. The training consisted of a review of the concerns, the engineering evaluation results, and procedure changes which resulted. Additionally, engineering support personnel received a briefing of the issues and evaluation results in the first quarter of 1993.

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~ 2 ATTACHMENTTO PLA4012 A1.3 MODIFICATION ENHANCEMENTS A1.3.1 SFP LEVEL AND TEMPERATURE An IMO CEMS level measurement system which has been seismically designed, but not qualified will be installed." The system includes a new probe mounted on a new seismically designed bracket in the fuel pool on elevation 818'. The probe output requires three conductors which will be routed in embedded conduit on elevation 818'o the Lower Relay Room in the Control Structure. A wall mounted electronics panel will be installed in the Lower Relay Room which includes a level transmitter. The output of the level transmitter will be wired to a dedicated recorder in the main control room on a back row panel.

The temperature measurement system will utilize the spare RTD in the existing temperature sensor which is seismically mounted in the fuel pool. This is already wired out to a panel on EL. 818', and will be routed to a temperature transmitter in the new wall mounted electronics panel in the Lower Relay Room. The output of the temperature transmitter will be routed to the dedicated recorder in the control room back panel used for level.

Power for both instrument loops and the recorder will be non Class 1E, but will be supplied from a Class 1E source through an isolator so as to provide additional reliability.

This modification is currently scheduled to be installed in the 1st quarter of 1994.

A1.3.2 CASK STORAGE PIT DRAIN LINE MODIFICATIONS PPAL is considering not installing the gates between the cask storage pit and SFPs during non-outage periods. This will greatly improve the ability to respond to a Loss of SFP cooling event since either units'ystems could be used to cool both SFPs. However, while this improves the response to a loss of SFP cooling event, it increases the risk to a SFP draindown event through valve mis-operation. During outages, when the gates are removed, additional measures such as the temporary installation of a standpipe over the cask pit drain are taken to prevent inadvertent draindown. PP&L is evaluating various options for installing permanent physical barriers to assure that inadvertent draindown does not occur. These measures will retain the ability to draindown the pit for spent fuel cask movements while providing positive protection against draindown during periods when the cask pit is crosstied to the SFPs.

FORMAI SEISMIC QUALIFICATION UNDER PP&L'S SQRT PROGRAM IS NOT REQUIRED BASED ON THE SAFETY FUNCTION OF THE INSTRUMENTATION.,

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~ k ATTACHMENTTO PLA-4012 APPENDIX 2: UHS IMPACT ANALYSIS Page 18

ATTACHMENTTO PLA4012 The design basis event for the Ultimate Heat Sink (UHS) is a LOCA/LOOP with a LOCA on

.one unit and a simultaneous forced shutdown on the other unit. In this mode, the UHS can provide sufficient cooling water for safety related equipment for 30 days with the water temperature at or below the system's design cooling water temperature, The decay heat assumed in the UHS design analyses for each reactor ranges from approximately 140 MBTU/hr after 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, to 36.6 MBTU/hr after day 7, to 22.6 MBTU/hr on day 30 of the event. This decay heat data is used the analysis for both the LOCA and the non-LOCA unit.

Three shutdown scenarios were evaluated to determine the impact that the operation of the RHR Fuel Pool Cooling (FPC) mode could have on the UHS minimum heat transfer (MHT) and maximum water loss (MWL) design basis analyses. The MHT analysis assumes conditions which result in the highest potential for pond heat-up. The MWL analysis evaluates worst case conditions for pond inventory losses. It is concluded, that for all three, the existing design basis UHS MHT and MWL analyses are bounding. The scenarios evaluated are:

CASE 1 Unit 1 shutdown for refueling. Core is off-loaded and a heat load of 33.9 MBTU/HR is resident in the connected pools. In this scenario, this unit's RHRFPC mode is used to cool the pool.

Unit 2 experiences the LOCA.

A LOOP occurs concurrently.

In this case, in terms of heat load to the UHS, the Unit 1 reactor decay heat is replaced by the heat load from the fuel pools. As indicated above, the heat load from the fuel pools is exceeded by the reactor decay heat rate for at least 7 days into the accident. The design UHS MHT temperature response peaks at about 1.8 days into the transient. Therefore, the UHS MHT design basis scenario bounds this shutdown case since the LOCA unit reactor decay heat contribution is the same for both cases, and the design case's non-LOCA unit decay heat rate far exceeds the heat load contribution from the fuel pools during the critical period for pond heat-up'otential (i.e., less than 3 days).

CASE 2 Unit 1 has just completed a refueling outage and the heat load in the connected fuel pools is 7.7 MBTU/hr (based on U17RIO decay heat values). Time to boil is 3.6 days.

Unit 1 or Unit 2 has the LOCA.

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r 4 ~ t ttC

./

ATTACHMfNT TO PLA4012 A LOOP occurs concurrently.

RHR Fuel Pool Cooling initiated on day 2.

Non-LOCA unit is cooled by shutdown cooling until day 2 when RHR is swapped between the fuel pool cooling and shutdown cooling modes.

The shutdown scenario for this case is identical as that for the UHS MHT design basis scenario except that, approximately two days into the accident, the non-LOCA unit RHR heat exchanger in shutdown cooling mode is swapped between fuel pool cooling and shutdown cooling modes as required to.maintain shutdown cooling and prevent the fuel pools from boiling. This condition is also bounded by the UWS MHT design basis analysis in that:

a. The fuel pool cooling heat load of 7.7 MBTU/hr is less than the design basis reactor decay heat rate for the 30 day accident.

b. Switching the non-LOCA unit RHR heat exchanger to fuel pool cooling at approximately two days into the transient effectively reduces the heat load into the spray pond at this time. Subsequent non-LOCA unit RHR heat exchanger swapping between shutdown cooling and fuel pool cooling modes will result in a lower than design integrated pond heat load for the period between 2 and 30 days.

Consequently, the UHS design temperature will not be exceeded.

c. *An added conservatism in this assessment is that the UHS MHT design analysis assumes both units reactor. decay heat rate based on 2 effective full power years (EFPY) of operation whereas in this case one unit has just returned from a refueling outage and thus generates much less decay heat.

CASE 3 Unit 1 has just completed a refueling outage and the heat load in the connected fuel pools is 7.7 MBTU/hr (based on U17RIO decay heat values). Time to boil is 3.6 days.

Unit 2 or Unit 1 has the LOCA.

A LOOP occurs concurrently.

RHR Fuel Pool Cooling initiated on day 2.

Alternate shutdown cooling is used to cool the non-LOCA unit and RHR fuel pool cooling used to cool the fuel pools starting on day 2.

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<I

~.

ATTACHMENTTO PLA4012 The shutdown scenario for this case is ~essenttatl the same as the UHS MHT design basis scenario except that a maximum of 7.7 MBTU/hr of fuel pools heat loads is added at about 2 days into the transient. This event is deemed to be bounded by the UHS MHT design basis analysis in that:

a. As indicated in the assessment for Case 2 above, the design basis decay heat for both units is identical and equivalent,to 2 EFPY of operation. However, this case's non-LOCA unit is assumed to have just returned from a refueling outage, in order to maximize the fuel pool heat load, and thus generates less decay heat.

b. The UHS design decay heat rate 2 days after shutdown is over 60% higher than that after, say, one month of operation. This conservative estimate results in over 30 MBTU/hr excess decay heat at two days into the transient as compared to the 7.7 MBTU/hr added heat load from the fuel pools.

c. As an added conservatism, calculations also demonstrate that the two loops/four spray arrays operation required for this case after 2 days is more efficient for pond cooling than the single loop/two arrays operation used in design basis analyses.

UHS

SUMMARY

Thus, it has been demonstrated that all the above shutdown scenarios for the UHS response to use of the RHR Fuel Pool Cooling mode are bounded by the design basis UHS MHT analysis. The UHS MWL response to these conditions is considered bounded by the design UHS MWL analysis in that:

a. The design basis UHS MWL analysis maximizes drift and evaporation water losses by assuming 2 RHR heat exchangers in service per unit for the duration of the 30 da transient.

b The worst case scenario from a UHS MWL standpoint, Case 3, may potentially use 2 RHR heat exchangers per unit but only after two days into the transient.

c. The UHS MWL design basis decay heat load is higher than that of the fuel pool cooling scenarios and thus result in spray water temperatures and consequently, higher evaporative losses.

d. The design basis UHS MWL analysis assumes loss of pond inventory for fuel pool make-up (70 CPM starting 24 hrs. into the transient) which is, of course, avoided if the fuel pools are not allowed to boil.

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ATTACHMENTTO PLA4012 FICURES Page 22

REFUELING FLOOR RE CIRC UNIT Nl REACTOR

~ SUCTION BLDG RE CIRC DISCHARGE UNIT N2 REACTOR BLDG SGTS FlGURE 1

LMIT 2 LAIT I HH gI SlKN4 SKPARATER ANT tÃYFTT CI -CI STTOIAlK PIT I I V'LVK PITS I0 I IKACTIN'IKlD STTNAGE AD 1AQQQtH NKh REACTOR BUILDING =L.8IS'-0

UNIT 2 UNIT I ~ y )'

~~

CASK PIT:~~

SPENT g

~

)UEL NKIKKER BKIKNEI SPAT SUlNK SllRGR 174K TANK 25307IA 2530118 153019A 25309IA FROH ESII 25309OA 2535OO 253001 10 2530loa FROM FUEL 1301 ES11 POOL lO RIN

~ ~ 2S3O9O8 253S 2530108 P 253o 918 253021 153001 TO CLEANN 251070 IO Sup.

POOL IKILT EX%,

251060 ORY HELL SPRAV 251 Food TO LPCj IA IA P BSU HV 251 F003 IB A HV 251 18 F006 A

RIN RHR HEAT A PINP EXCH, SERVICE A LOOP HV 251 NAIER Fool A

FROH SUPPRF SS ION POOL FUEL POOL COOLIHO PUIIP8

0 4 s'.+

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