ML17157C138

Safety Consequences of Boiling Spent Fuel Pool at Susquehanna Steam Electric Station.

ML17157C138
Person / Time
Site:	Susquehanna
Issue date:	07/27/1992
From:	Lochbaum D PENNSYLVANIA POWER & LIGHT CO.
To:
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ML17157C139	List: ML17157C138 ML17157C140 ML17157C141 ML17157C142 ML18026A248
References
NUDOCS 9301050141
Download: ML17157C138 (41)

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SAFETY CONSEQUENCES OF A BOILING SPENT FUEL POOL AT THE SUSQUEHANNA STEAM ELECTRIC STATION JUlp 21 g 1992 Prepared hy:

Dav d . Lochbaum Donald C. Prevatte 930i0501%i 92if27 PDR ADOCK 05000387 8 PDR

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EXECUTIVE

SUMMARY

Engineering Discrepancy Report (EDR) G20020 was written in April 1992 after nine potential problems associated with the boiling spent fuel-pool event were identified during system evaluations to support the power uprate project for PP&L's Susquehanna Steam Electric Station. The major concerns raised in EDR G20020 are:

1 Regulations require that instrumentation shall be provided for the fuel storage systems to detect conditions that may result

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in loss of heat removal capability and to initiate appropriate safety actions.

Contrary to this requirement, the water level and temperature instrumentation for the spent fuel pools do not satisfy Class 1E criteria and are not included in the equipment qualification program. These instruments will fail following a loss of offsite power and may fail following a loss of coolant accident. The ultimate consequence of such failure could be containment.

'n irradiated fuel meltdown outside primary 2 Regulations require that nuclear power plant designs limit personnel radiation exposures to < 5 Rem per individual for

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control room occupation and actions required to mitigate or recover from an accident.

Contrary to this requirement, the manual ESW valve manipulations required to provide makeup to a boiling spent fuel pool following a loss of coolant accident could require a radiation exposure significantly higher than 5 Rem. The ultimate consequence could be significant radiation overexposure or inability to provide ESW makeup and an irradiated fuel meltdown outside primary containment.

3 ~ Regulations require that structures, systems and components important to safety be designed to accommodate the effects of the environmental conditions associated with postulated accidents.

Contrary to this requirement, the effects of ESW makeup water to a boiling spent fuel pool have not been considered in the SSES design. The effects include flooding, high temperature, and high humidity. The ultimate consequences could include failure of multiple ECCS and other safety related systems.

4 ~ Regulations require that electrical equipment be qualified to the temperature for the most severe design basis accidents.

Contrary to this requirement, the SSES reactor building temperature analyses used in equipment qualification evaluations do not account for the heat load from a boiling spent fuel pool. The ultimate consequences could include failure of multiple ECCS and other safety related systems.

-Safety Consequences of a Boiling Spent Fuel Pool SYSTEM DESCRIPTION Each of the two operating nuclear power plants at the Pennsylvania Power and Light (PP&L) Company's Susquehanna Steam Electric Station (SSES) has a spent fuel pool. Each spent fuel pool is designed to store up to 2,840 irradiated fuel bundles discharged from the reactor core after approximately four and a half years of operation. As of July 1992, the Unit 1 spent fuel pool contained 1400 irradiated fuel bundles and the Unit 2 spent fuel pool held 1004 irradiated fuel bundles.

The irradiated fuel bundles stored in the spent fuel pools generate heat from the nuclear decay of fission products. The amount of heat generation exponentially decreases with time as a function of the half life of the fission products.

The spent fuel pools are located in a common refueling area within the secondary containment structure. Each spent fuel pool is connected to a reactor cavity and to the other spent fuel pool.

The reactor cavity is the area above the reactor pressure vessel which is flooded during a refueling outage after removing the drywell shield blocks, drywell head and reactor pressure vessel head to permit fuel transfer between the reactor core and the spent fuel pool. These connections are normally isolated, except during refueling outages, using gates.

Unit 1 Unit 1 Unit 2 Unit 2 Reactor Spent Spent Reactor Cavity Fuel Fuel Cavity Pool Cask Pool Pit Each spent fuel pool has a fuel pool cooling and cleanup system (FPCCS) which circulates water from the fuel pool through a heat exchanger and demineralizer to maintain proper fuel pool water chemistry and to keep its temperature < 125oF. - The FPCCS has a design capacity of 13.2x106 BTU/hr. As of July 1992, the decay heat load in the Unit 2 spent fuel pool was =2.1x106 BTU/hr while the decay heat load in the Unit 2 spent fuel pool was =2.97x106 BTU/hr. Heat from the FPCCS heat, exchangers is transferred to the service water (SW) system which in turn dissipates the energy to the atmosphere via the cooling tower. The FPCCS and the SW system are non-safety related systems which are not designed to satisfy seismic, Class 1E power, equipment qualification and single failure criteria. The FPCCS is designed such that which drains water from the spent fuel pool.

it cannot fail in a way July 27, 1992 Page 2

Safety Consequences of a Boiling Spent Fuel Pool If the FPCCS is unavailable, the fuel pool cooling assist mode of the residual heat removal (RHR) system is designed to circulate water from the spent fuel pool through a heat exchanger to keep the fuel pool from boiling. The fuel pool cooling assist mode of RHR is manually initiated by opening valves in the reactor building.

The fuel pool cooling assist mode of RHR has a capacity of 32.6x106 BTU/hr. Heat from the RHR heat exchanger is transferred to the RHR service water (RHRSW) which in turn dissipates the energy to the atmosphere via the spray pond. The fuel pool cooling assist mode of RHR is a non-safety. related function which is not designed to satisfy seismic, Class 1E power, and single failure criteria. The fuel pool cooling assist mode of RHR is designed such that cannot fail in a way which drains water from the spent fuel pool.

it If both the FPCCS and the fuel pool cooling assist mode unavailable, the spent fuel pool water will boil unless of RHR are cooling is re-established. The time required to reach boiling is a function of the decay heat load in the spent fuel pool, the initial temperature of the water, and the volume of water available. The volume of water available is primarily dependent upon the presence or absence of the gates between the spent fuel pools and the reactor cavity. The emergency service water (ESW) system is designed to provide makeup to the boiling spent fuel pool to compensate for water lost through boil-off and evaporation. The ESW makeup supply is manually initiated by opening three valves in the reactor building. The ESW system uses water from the spray pond. The ESW system and the spray pond are safety related systems which are designed to satisfy seismic, Class 1E power, and single failure criteria as applicable. The design provision at SSES is for the ESW system to provide adequate makeup to a boiling spent fuel pool if cooling is lost.

The reactor. building heating, ventilating and air conditioning (RB-HVAC) system circulates tempered air through each reactor building and the refueling zone during normal operation. The RB-HVAC system maintains these areas at a slight negative pressure relative to the outside environment to prevent .leakage. of potentially airborne radioactivity to the atmosphere. The exhaust from the potentially contaminated areas is filtered to remove radioactive materials. In an emergency, the supply and exhaust lines are isolated and the RB-HVAC system recirculates air throughout the reactor building affected by the emergency and the refueling zone. During a loss of offsite power (LOOP), the supply and exhaust lines are isolated and the RB-HVAC system recirculates air throughout the both reactor buildings and the refueling zone. The RB-HVAC system in recirculation mode does not provide any cooling function, so the reactor building and refueling zone air temperatures increase based upon piping, lighting, transmission and equipment heat loads.

July 27, 1992 Page 3

Safety Consequences of a Boiling Spent Fuel Pool The standby gas treatment system (SGTS) is designed to maintain the secondary .containment at a negative pressure relative to the outside environment in an emergency. The SGTS takes suction on the recirculation plenum of the RB-HVAC system and processes this air through a filter train to remove radioactive materials. The SGTS is normally in standby except during testing. The SGTS iscriteria. designed to satisfy seismic, Class 1E power, and single failure The emergency core cooling systems (ECCS),and the reactor core isolation cooling (RCZC) system are located in the lower elevations of each reactor building. These systems provide water to .the reactor pressure vessel during transients and accidents. These systems are normally in standby except during testing. The ECCS are designed to satisfy seismic, Class 1E power, and single failure criteria.

BOILING SPENT FUEL POOL DESIGN ANALYSIS SSES Final Safety Analysis Report (FSAR) Appendix 9A reports the results of an analysis performed to quantify the radiological consequences of a loss of spent fuel pool cooling event. The analysis assumed the initiating event was an earthquake which resulted in the failure of the FPCCS on both units. The analysis concluded that the secondary containment design with SGTS operation kept offsite doses to a small fraction of 10 CFR 100 limits even with conservative assumptions of initial fuel failures in the spent fuel pools.

CONTAINMENT DESIGN ANALYSES SSES Final Safety Analysis Report (FSAR) Chapter 6 reports the results of analyses performed to demonstrate the capability of the safety related systems to mitigate the consequences of postulated accidents such that the containment design parameters are not exceeded. The postulated accidents included main steam line breaks and loss-of-coolant accidents (LOCAs) with and without concurrent loss of offsite power. A design basis accident (DBA) for SSES is defined as a LOCA with a simultaneous LOOP and safe shutdown earthquake and the worst case single failure which results in the maximum containment pressure and temperature conditions. SSES FSAR Chapter 6 indicates margin to containment design parameters for the analyzed postulated accidents.

Reactor building room temperatures following postulated accidents were analyzed for equipment qualification. A procedure to manually shed all the non-Class 1E power loads in the reactor building =24 hours after a LOCA without a LOOP was developed to prevent room temperatures from exceeding equipment qualification limitations.

July 27, 1992 Page 4

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Safety Consequences of a Boiling Spent Puel Pool CONCERNS OVER BOZLZNG SPENT FUEL POOL EVENT Engineering Discrepancy Report (EDR) G20020 was written in April 1992 after nine potential problems associated with the boiling spent fuel pool event were identified during system evaluations to support the power uprate project. The four major concerns raised in EDR G20020 are:

Znadequate Znstrumentation A. Re u a or Re u' e ts 10 CFR 50 Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (l) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate 'safety actions."

Regulatory 'uide 1.97 defines accident-monitoring instrumentation to include "those variables to be monitored that provide the primary information requfred to permit the control room operators to take the specified manually controlled actions for which no automatfc control is provided and that are required for safety systems to accomplish their safety function for design basis accident events."

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design includes "the i nstrumentati on provided for ini tiati ng appropriate safety actions."

Standard Review Plan (NUREG-0800) 9.1.3 for the spent fuel pool cooling and cleanup system states that the "safety function to be performed by the system in all cases remains the same; that is, the spent fuel assemblies must be cooled and must remain covered wi th water during all storage conditions."

Standard Review Plan (NUREG-0800) 7.1 states that "fnformation systems important to safety include those systems whi ch provide informati on for manual ini tfation and control of safety systems, to indicate'hat plant safety functions are bei ng accomplished, and t o provi de information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accfdents."

July 27, 1992 Page 5

Safety Consequences of a Boiling Spent Fuel Pool B. ~Co ~ce ~s The ESW system is required to provide makeup water to the spent fuel pools following loss of fuel pool cooling to keep the irradiated fuel covered thus preventing fuel damage from overheating. A loss of offsite power or LOCA can result in loss of fuel pool cooling since the FPCCS and the SW system are not normally supplied by Class 1E power. The loss of offsite power will also disable the spent fuel pool temperature and level instruments monitored by the operator and used to initiate the safety'ction of providing ESW makeup to the boiling spent fuel pool. The post-LOCA environment in the reactor building may disable the spent fuel pool temperature and level instruments since they are not covered under the equipment qualification program. Therefore, the existing spent fuel pool temperature and level instrumentation is inadequate to ensure the required safety action of providing adequate makeup to. a boiling spent fuel pool is properly initiated and monitored under all postulated accident conditions.

Zf the spent fuel pool is permitted to boil without adequate makeup, its water level will drop. A study by the PP&L Nuclear Safety Assurance Group (NSAG Report 13-84, December 1984) reported that water level in the spent fuel pool dropping to within five inches of the top of the irradiated fuel "would cause radiation levels on the 818'levation of the reactor building in excess of 200,000 rem/hour." At that dose rate, an individual on the refueling floor would receive a lethal radiation exposure in approximately 16 seconds. This severe condition is just the beginning of the adverse consequences of spent fuel pool boiling without adequate makeup. At this point, the radiation source term results in offsite doses exceeding 10 CFR 100 limits and in doseaccess.rates within the reactor building that prevent any personnel

.-The situation progresses ultimately to uncovering irradiated fuel bundles in the spent fuel pool and fuel damage froma overheating. The situation has the potential for substantial meltdown of irradiated fuel outside the primary containment.

July 27, 1992 Page 6

Safety Consequences of a Boiling Spent Fuel Pool 2 Manual ESR Valve Operation s'o

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A. e to Re uireme ts and Lice 'tments 10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures ... as low as is reasonably achievable."

10 CFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem. GDC 19 also requires design features for equipment outside the contxol room to permit operation in accordance with suitable procedures.

10 CFR 50.47(b)(11) states that licensees assure that "means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radi ologi cal exposures shall i nclude exposur e guidelines consistent with EPA Emergency 8'orker and Lifesaving Activity Protective Action Guides."

SSES FSAR 18.1.20 in response to NUREG-0737 Item IZ.B.2 states that "each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural controls. The design review shall determine which types of correcti,ve actions are needed for vi tal areas throughout the facility."

SSES FSAR 18.1.20.3.3.4.1 defines vital areas as those "which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident."

SSES FSAR 18. 1. 20. 3.2. 1 states that "a review was made to determine which systems could be required to operate and/or be expected to contain highly radioactive materials following a postulated" accident where substantial core damage has occurred .

SSES FSAR 18.1.20.3.2.5 states= "exposures for areas not continuously occupied (frequent and infrequent occupancy) must be determined case by case, that is, mul tiply the task duration by the area dose rate at the time of exposure."

SSES FSAR 18.1.20.3.3.3 states that "GDC 19 is also used to govern design bases for the maxi mum permi ssi bl e dosage to personnel performing any task required post-accident. These requirements translate roughly into the ob)ectives to be met in the post-accident review as given below.

July 27 I 1992 Page 7

Safety Consequences of a Boiling Spent Fuel Pool Radiation Exposure Guidelines Occupancy Dose Rate Objectives Dose Objective Continuous l5 mR/hr 5 Rem for duration Frequent 200 mR/hr 5 Rem - all activities Infrecprent 500 mR/hr 5 Rem per activity Accessway 5 R/hr Incl in above doses" SSES FSAR 18.1.20.3.4.3 states that the review results "show that the reactor building will be generally inaccessible for several days after the accident due to contained radiation sources."

SSES FSAR Figure 18.1-4 shows Room I-105 where ESW valves 153500/153501 are located to be in Rad Zone VIII with dose rates over 5000 R/hr. SSES FSAR Figure 18.1-6 shows Room I-514 where ESW valves 153090A&B and 153091A&B are located to be in Rad Zone V with dose rates between 5 and 50 R/hr. These valves must be manually opened to initiate ESW makeup to .the spent fuel pools in the loss of fuel pool cooling event.

PP&L administrative procedure NDI-6.4.3 specifies that the whole body dose for life saving actions "shall not exceed 75 rem" and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem."

10 CFR 20's ALARA provision requires plant design to minimize radiation exposure. Application of the emergency dose guidelines to a design which requires manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.

B. Concerns The ESW system is required to provide makeup to the spent fuel pools following loss of fuel pool cooling. Either a seismic event or loss of offsite power can lead to loss of fuel pool cooling. Both conditions are assumed to occur concurrent with a LOCA in the DBA for containment analyses. However, the post-LOCA dose rates in the reactor building areas where the manual valves are located are 5 to 5,000+ R/hr and will prevent these valves from being'ccessed without excessive radiation exposure to the operator. In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently.

Therefore, the manual ESW valve manipulations required for makeup to boiling spent fuel pools may not be accomplished for all postulated accident conditions.

,July 27 I 1992 Page 8

Safety Consequences of a Boiling Spent Fuel Pool Zn addition, since the boiling spent fuel pool analysis reported in SSES FSAR Appendix 9A assumed a seismic event initiated the loss of fuel pool cooling, the intentional shedding of non-Class 1E power loads in the reactor building following a LOCA without a LOOP represents either the creation of a nev kind of accident or the increased probability of a previously analyzed accident.

3 ~ Effects of ESW Makeup Water on Reactor Building Systems A. Re u ator Re uire ents 'ce s n Comm tme ts and Desi Bases 10 CFR 50 Appendix A General Design Criterion 4 states that "structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible wi th the environmental conditions associated with normal operation, maintenance, testing, and postul ated accidents, including loss-of-coolant accidents."

Standard Review Plan {NUREG-0800) 3.4.1 states that the review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result on uncontrolled release of significant radi oacti vi ty... " and that this review "also includes consideration of flooding from internal sources."

SSES FSAR 6.3.1.1.3 states that, separation barriers for ECCS "shall be constructed between the functional groups as required to assure that envi ronmental disturbances such as fire, pi pe rupture, falling obj ects, etc., affecting one functional groups will not affect the remaining groups. In addition, separation barriers shall be provided as required to assure that such disturbances do not affect both RCIC and HPCI."

SSES FSAR 9.1.3.3 states that "the design makeup rate from each ESW loop. is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of FPCCS capacity."

Minutes from Bechtel meeting on HVAC systems (February 1980) states that original requirement for SGTS was "to handle fumes from a boiling fuel pool," but that SGTS will not be able to handle this mixture since the room will become too hot. "This requirement will be deleted from the FSAR."

An internal PPEL engineering work request {EWR 830658, March 1983) noted "condensation may be expected from this July 27, 1992 Page 9

Safety Consequences of a Boiling Spent Fuel Pool evaporation which Vill this cause' vill run dovn to lover levels of the R.B.

oss of essenti al equipment, particularly electrical? 'as an evaluation been performed?" The response to these questions was "This is an inappropriate format to ask questions. Comments vere requested and none received.

Furthermore, no budget exists vi th vhi ch to fund the engineering time required to respond to these questions."

B. Concerns The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each spent fuel pool over the 30 day period. In the LOCA-LOOP condition, the reactor building HVAC system in Zone I, II and III isolation mode recirculates refueling floor air throughout all three zones. The water added to the spent fuel pools ends up in the reactor building following boil-off and overflow. The effects of this water on the safety-related structures, systems and components in the reactor buildings have not been included in design analyses.

The ECCS and RCIC room coolers are known not to be designed for latent heat effects. Dampers in the SGTS and RB-HVAC system close when the entering air temperature exceeds 165'F, while the boiling spent fuel pool was calculated to produce air temperatures of =1804F, The potential for common mode failures of multiple ECCS and safety-related systems such as the standby gas treatment system exists. Failure of one or more of these safety-related systems could increase the consequences of postulated accidents.

4o Reactor Building Heat Loads A. Re ulator Re ui ements Licensin Commitments and Desi n Bases 10 CFR 50. 49 requires that electrical equipment must be qualified to the temperature "for the most severe design basis accidents. "

10 CFR 50 Appendix A General Design Criterion 4 states that "structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the envi ronmental conditions associated with normal operation, maintenance, testing, and postul a ted accidents, including loss-of-coolant accidents."

An internal PP&L engineering work request (EWR 830658, March 1983) noted "the initial boiling rate corresponds to =3000 cfm of 2004 vater vapor at one atm. Is the equipment which vill be exposed to this atmosphere qualified for it?" The response July 27, 1992 Page 10

Safety Consequences of a Boiling Spent Fuel Pool to this question was "This, is an inappropriate format to ask questions. Comments were requested and none received.

Furthermore, no budget exists with which to fund the engineering time required to respond to these questions."

A PP&L engineering report (SEA-ME-099, December 1987) analyzed reactor building temperatures for LOCA, LOCA/LOOP and LOCA/false LOCA cases assuming spent fuel temperatures remained at 125'F, but listed as a nonconservatism that fuel pool heatup in the LOCA/LOOP case would result in higher heat loads from the RHR systems, fuel pool walls and fuel pool surface.

B. ~Co cer Secondary containment design analyses are required to account for all heat loads in the reactor building reactor building including from the boiling spent fuel pool. The existing design heat load calcs consider sensible heat from the boiling pool, but neglect latent heat. These calcs indicate little margin to equipment qualification temperature limits in many rooms for a maximum heat load in the reactor heat building of approximately 5.5x106 BTU/hr. The total design load from the spent fuel pools is 26.4x10 BTU/hr, which would add at least approximately 20.9x106 BTU/hr to the existing maximum heat load. Even the current heat loads in the spent fuel pools could increase the maximum heat load in the reactor building by =504. =-

The remaining five concerns raised in EDR G20020 involved nonconservatisms in analyses for the boiling spent fuel pool event.

DZSCUSSZON OF OPPOSING VIEWPOINT The discussions and meetings which have occurred since EDR G20020 was initiated have yielded one primary argumentsignificance.against the issues raised in EDR G20020 having nuclear safety This argument is that the licensing bases LOCA/LOOP accident for SSES does not assume a boiling spent fuel pool resulting from the event.

In order for this assumption to be valid, spent fuel pool boiling. cooling must either not be lost or must be restored prior to There are several faults in this assumption:

1) Since the FPCCS and the SW system are non-safety related systems, their components are not included in the equipment qualification program and may not survive the pressure, temperature, humidity and radiation environment in the reactox building following a postulated accident. Therefore, the July 27, 1992 Page 11

Safety Consequences of a Boil.ing Spent Fuel Pool FPCCS which is definitely lost following a LOCA/LOOP may also be lost following a LOCA without a LOOP.

2) Since the fuel pool cooling assist mode of RHR is a non-safety related function, its components are not included in the equipment qualification program and therefore may not survive the pressure, temperature, humidity and radiation environment in the reactor building following a postulated accident. In addition, the fuel pool cooling assist mode of RHR has not been utilized since the initial startup testing program and its valves were removed from the inservice inspection program several years ago and the valves may have experienced failures which have not yet been detected which would prevent their successful operation. Therefore, the fuel pool cooling assist mode of RHR may be lost following a LOCA/LOOP and a LOCA without a LOOP.

3) The fuel pool cooling assist mode of RHR requires the manual opening of valves in the reactor building which may be inaccessible following a postulated accident due to radiation levels.

4) For design the LOCA/LOOP case, it has been argued that the implicitly assumes restoration of offsite power SSES typically within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and essentia.ally always within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after event initiation. SSES FSAR Chapter 8 reports PP&L grid experience in support of these restoration times.

However, no documentation was found which states that PP&L has defined the LOOP duration for design bases events. As EDR G20020 and EDR G00005 both address, the spent fuel pool may begin boiling in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In any case, the reactor building temperature analyses for equipment qualification purposes presently counter any such credit for restoration of offsite power since non-Class 1E power loads in the reactor building may be shed =24 hours after offsite power is restored in order to satisfy room temperature limitations.

EDR G20020 identified concerns with the SSES design provisions for the boiling spent fuel pool event. The SSES design, coupled with current operating procedures, would have significant nuclear safety consequences if a loss of spent fuel pool cooling occurred.

Therefore, these concerns must be resolved for SSES. In addition, many of these concerns are applicable to other BWRs and possibly even PWRs in the United States. Therefore, these concerns must be reported to INFO/NRC in order for the adverse condition to be remedied throughout the industry.

July 27, 1992 Page 12

Attachaent 9 PPSL Memo from G. D. Miller to G. T. Jones, "Fuel Pool Cooling Deficiencies", August l8, 1992 (ET-0586)

Kote: This memo by the PPSL Supervisor, Engineering Projects provides an indication of how PP&L narrowed their scope of evaluation for the concerns in EDR G20020 to just the design of the fuel pool cooling and cleanup system.

With the exception of the instrumentation for the fuel pools, the .design of the fuel pool cooling and cleanup system has not been challenged subsequent supporting documents.

in EDR G20020 and 'ts The concerns are that the ef fects of boi ing spent fue1 pools on ~h~e systems and 1

components in the reactor building have not been adequately

. analyzed.

Page 1 0 RAND UH DATE: 8/18/92 TO: G. T. Jones A6-2 FRON: G. D. Hiller A6-3 JOB: Engineering NUHBER: ET-0586 COPIES: Distribution Technology Corres. File A6-2 Engr Tech File A6-3 FILE: A45-1A REPLY: Not applicable

SUBJECT:

Fuel Pool Cooling Deficiencies The following actions are being taken in response to concerns raised about the EOR program and the fuel pool cooling issues described in EDR G20020 and EOR G00005.

E R Process Chan es The EDR process governing procedure, EPH-gA-122, will be revised as follows:

The appeal process described in paragraph 5.4 will be revised with the first step of the appeal changed to the Engineering Review Committee. Subsequent appeal may be .to either the Hanager Nuclear Engineering, Superintendent SSES, or Hanager NSAG.

2. The management review described in paragraph 5.14 will be revised to include:

a ~ Periodic review of all invalid EDRs by the Engineering Review Committee, and

b. Periodic review of all EDRs open greater than six months by the Engineering Review Committee.

Fuel Pool Coolin Oeficienc Resolution Engineering Technology has responsibility to resolve both EDRs (EOR G00005 was previously assigned to System Engineering). These EDRs are assigned to Hark Hjaatvedt and are being worked full time by Hichael Crowthers. The EDR EPH-101B, Rev. 1

N E N 0 R A N D U N (CONTINUATION SHEET) evaluations (safety significance, operability/reportability) are being done by Jim Agnew and Joe Zola.

Our initial evaluation of these EDRs has concluded that the safety significance is minimal. This is based primarily on our understanding that the design of the fuel pool cooling system was specifically reviewed and approved by the NRC with full knowledge of the fact that the FPCCS was not a safety'elated system and that fuel pool boiling could be expected to occur under a specific combination of hypothetical conditions. However, it is not yet clear to what extent the NRC

{or the industry) considered the long-term effects of the fuel pool boiling condition. The record on this subject is confusing and further complicated by changes to the fuel design and actual outage practices which have 'not been accounted for in the FSAR analysis. Me believe that the design and procedural features which exist today provide a reasonable level of assurance that the actual safety consequences are minimized. However, procedure enhancements and additional operator training are clearly required as part of the resolution of these concerns. This evaluation will be fully documented as part of the revised EDR package.

A final evaluation of this concern is predicated on completion of a historical review of all available documentation. Thus, our plan for resolution includes:

I. Complete investigation of historical design and licensing information, including requests for information from the original design organizations (GE and Bechtel),

2. Establish the fuel pool and fuel pool cooling system design basis based on the derived design basis and current operating practice,

3. Review fuel pool designs of other boiling water reactors, 4, Complete a new analysis of the fuel pool and fuel pool cooling system based on the established design basis,

5. Prepare a point-by-point comparative description of our current operating practice and design basis against the original FSAR analysis {including the individual issues identified in the EDR),

List and assess each deviation from the or iginal analysis as described in item 5,

7. Prepare an operability evaluation accounting for each deviation as additional information becomes available,

8. Re-evaluate all issues for reportability {ongoing),

Prepare recommendations to,resolve each issue desc} ibed in item 4.

EPN-IOIC, Rev. I {51)

Page 3 E N 0 R A N D U (CONTINUATION SHEET)

We plan to keep the originators of these concerns informed of our progress as we work our way through this effort. A formal plan including schedule for the above activities is under development.

Inde endent Review b S stems Anal sis Systems Analysis (Kevin Brinckman) is in the process of conducting an independent, design review of these issues. Additionally, I have requested their review of this issue from an IPE perspective when resources become available.

Assessment of EDR Process As a followup activity I plan to request an assessment of the EDR process from an independent organization. This assessment will focus specifically on the validation and verification steps of the process. This should be conducted by NSAG.

g thtd~ii NgA has once in the past conducted t

an audit of the process. They do not iiydf'll i . Thydfi y discrepancy to be a condition adverse to quality, whereas our program recognizes the potential for discrepancies in documentation which do not constitute actual deficient conditions. Other programmatic audits have taken place on the EDR process, but none have examined the philosophy or criteria used to determine the validity of engineering issues.

EPN-101C, Rev. 1 (51)

Attachaent 10 PP&L Memo from O. C. Prevatte to G. T. Jones, "Fuel Pool Cooling Oeficiencies", August 20, 1992 (ET-0587)

MEMORANDUM TO: G. T. Jones A6-2 DATE: August 20, 1992 FROM D. C. Prevatte A6-3 FILE: A45-1A JOB: Engineering Technology cc: Distribution Corres. File A6-2 NUMBER ET-0587 REPLY: ET File A6-3

SUBJECT:

Fuel Pool Cooling Deficiencies This memo is written in response to Mr. G. D. Miller's memo ET-0586 of 8/18/92 concerning +he dis" epancies associated w-'th the fuel pool cooling system described in EDRs G20020 and G00005.

Mr. Miller's memo states that, "Our initial evaluation of the EDRs has concluded that the safety significance is minimal." I strongly disagree with this evaluation and I hereby request that the safety significance of these EDRs, particularly EDR G20020 be reevaluated for the following reasons:

1. The primary basis given for'his conclusion is that "...the design of the fuel pool cooling system was specifically reviewed and approved by the NRC with full knowledge of the fact that the FPCCS was not a safety related system and pool boiling could be expected to occur under a specific that'uel combination of hypothetical, conditions."

I consider this basis to be invalid for the following reasons:

a~ This basis appears to miss most of the main points of EDR G20020.

is not It the focuses on the non-safety related FPCCS which concern. The primary concerns are with the NRC mandated (Reg. Guide 1.13), safety-related backup cooling scheme of allowing the fuel pool to boil and providing makeup water from the safety-related ESW system. The concern is the potential inability of the operators to put this scheme into effect because of inaccessibility to the associated valves due to post-LOCA radiation levels in the reactor building, and the potential negative effects of a boiling spent fuel pool on virtually all of the safety-related systems in the reactor building, effects which have not been analyzed.

b. The "... specific combination of hypothetical conditions

..." referred to in the memo is LOCA/LOOP. This is not some off-the-wall accident scenario as the response seems to imply. This is the standard, universally recognized, NRC mandated design basis accident (DBA).

G. T. Jones August 20, 1992 Fuel Pool Cooling Deficiencies Page 2 The conditions of concern are not postulated. They are mechanistic consequences of that DBA.

2. The memo also cites as another basis, confusion concerning what was originally considered in the design, the changes to the fuel design, and outage practices which have not been accounted for in the FSAR analysis. This basis would seem to support the EDRs contentions, not refute them.

at If indeed there is confusion on these subjects, then, best, the condition is unanalyzed and therefore by definition in NRC regulations and our procedures, a safety concern.

Although this information is certainly pertinent to a historical perspective of these concerns, to determination of the magnitude of the heat loads involved, and to formulation of the corrective actions that may be effected, it is not pertinent to the ability of the plant to perform as required for the DBA conditions. The information to make this determination is clear and available today.

3 ~ Another basis cited is that "... the design and procedural features which exist today provide. a reasonable level of assurance that the actual safety consequences are minimal."

There is no elaboration on what these design and procedural features are. In conversations with Mr. Miller and others who seem to consider the EDRs as having very low safety significance, no design or procedural features have"heroic been cited. The only features that have been cited are action" of the operators, the EOPs, an EOC staff who will understand the concern and do whatever needs to be done, and a low probability of occurrence.

These are not valid features. Heroic operator action is not a valid basis for the design of a plant, nor are the EOPs (even'f they were correct in this area) which address many conditions potentially outside the plant licensing and/or design bases.

And, contrary to the memo's contention, the EOPs as they stand today are not correct. They currently tell the operator he has a minimum of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> until the fuel pool boils. Under worst case conditions, it may be less than half that time; and with the LOOP 'conditions, he has no instrumentation to tell him the condition of the fuel pool. Under DBA conditions, the operator is flying blind using nonconservative information.

Additionally, current EOPs move the plant toward the conditions of concern, not away from them. The current EOPs require deenergizing the non-1E loads in the reactor building

G.~ T.~ Jones August 20, 1992 Fuel Pool Cooling Deficiencies Page 3 at t=24 hours if reactor building temperatures are as analyzed. This, in effect, imposes a LOOP on the reactor building, thus initiating the conditions of concern.

The knowledge of the EOC staff is also not a valid feature if the conditions of .concern are not formally addressed in any official design and/or procedural documents. -Although is claimed that today's staff would understand the concerns, it there is no reason to believe this is true since the concerns aren't documented outside the EDRs and there is no training on this eventuality. And ten years from now, if the concerns are not formalized in writing, they will be even less understood.

Additionally, even if the staff does understand, if the conditions are not analyzed, which they are not, the plant could be brought to a condition where recovery is not possible in spite of their full understanding.

Low probability is also not a valid feature. This is discussed in detail further in this memo.

4 ~ The statement in the memo regarding the NRC's "... full knowledge of the fact that the FPCCS was not a safety-related system and that fuel pool boiling could be expected to occur seems to imply that if the NRC approved it as is', that makes it acceptable even if we discover discrepancies that may not, have been originally considered. I am aware of no evidence that indicates the NRC approved of our design with the understanding that: (a) the operator would be exposed to unacceptable radiation levels under design basis conditions in effecting the FSAR described fuel pool boil scheme for alternate cooling; and (b) that the boiling fuel pool might create a myriad of unanalyzed conditions in the reactor building that could threaten the operability of many of the safety-related systems in the building.

5. The memo concludes that the safety significance is "minimal."

Per procedure EPM-703, Rev. 0, Section 5.3, "...a generally signifies a documentation type of

'minimal'lassification discrepancy." Zn other words, not a real engineering concern, but'ather a documentation error that can be resolved by making editorial changes to the documents. Per this procedure, if an EDR's safety significance is classified as "minimal", it does not even have to be evaluated for operability and reportability.

These EDRs are not in any reasonable evaluation just a documentation discrepancy. They are fundamental engineering concerns raised by two engineers intimately familiar with the systems after exhaustive research. To dismiss these concerns

G. T. Jones August 20, 1992 Fuel Pool Cooling Deficiencies Page 4 by classifying them as just a documentation discrepancy is ludicrous. If concerns such as these don't even get to the stage in the process where they are required to be evaluated for operability and reportability, concerns which involve the safety of the operators and potential threat to virtually every safety-related system in the reactor building, then what does it take to trigger operability and reportability evaluations? The threshold appears to be much too high.

6. The memo's reference to "... a specific combination of hypothetical conditions ..." implies a probabalistic argument as to why the safety significance is "minimal." Indeed, in conversations with Mr. Miller and others this argument has been explicitly raised. This argument is not valid with regard to design bases for several reasons.

First, our design basis conditions of LOCA/LOOP which produce the conditions of concern are mandated by regulation. That, for design purposes, dictates a probability of 1.

Second, even for LOCA without a LOOP, our current EOPs dictate a self-imposed LOOP on the reactor building at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, again making the probability for LOCA/LOOP equal l.

Third, even for a LOOP or FPCCS system failure without a LOCA, the consequences of fuel pool boil are unanalyzed.

Fourth, EDR procedure EPM-703, Rev. 0, Section 5.3, has as a caution, capitalized, bold letters and underlined as follows, "The EDMG Evaluator must not put heavy emphasis on the perceived small probability of occurrence or the expected satisfactory outcome of analysis or reanalysis to justify continued operation with the existing discrepancy." Section 5.4 goes on to say, "SAFETY SIGNIFICANCE must be based on the potential adverse consequences of failures, even those of very low probability." Thus, by our procedures, potential consequences should be the dominant factor in evaluating safety significance, not probability. The potential consequences of the concerns raised in these EDRs, and subsequent documentation generated by Mr. Lochbaum and myself, are very grave.

The EDR process at PP&L was developed in response to a 1990 SALP inspection finding that safety significant issues were not being handled in a timely manner. Our EDR procedures are filled with words that reflect this concern; words like quickly, expeditiously, immediately, early, timely. For the step where we are today, the "screening" step, the procedures'EPM-703, Rev. 0, Section 5.2)

G. T. Jones August 20, 1992 Fuel Pool Cooling Deficiencies Page 5 intent is that EDRs be "quickly" screened after a discrepancy enters the EDR process. As of today, the official screening still has not been completed four months after the EDR entered the process, and approximately one month after Mr. Lochbaum and I personally brought these concerns to your attention. Neither the intent of the procedure nor the intent of the NRC are being fulfilled.

The plan outlined in Mr. Miller's memo for the "final evaluation" would appear to further delay the required actions.

Although all of the activities in the plan are important to understanding the problems more completely and effecting the most effective olutions, none of them are prerequisites for performing a valid "screening", and.most of them are not required to determine operability and reportability. To make these activities prerequisites for a final "screening" evaluation and then for the operability and reportability determinations, is to further delay the process unnecessarily. The information is available to make these determinations today, and they should be made immediately we are to do what's legitimately required of us.

if I have made my own operability and reportability determinations based on extensive research on these concerns. At best, the operability of the fuel pool cooling in the boil and feed mode, the fuel pool instrumentation, and much of the safety-related equipment in the reactor building is unanalyzed with regard to the effects of the boiling fuel pool on this equipment, with strong indications that, analysis would show it. as inoperable.

per 10CFR50.72 and 50.73, it is reportable.

If this is the case, I would welcome any hard, definitive, documentary information indicating that my conclusions are wrong. Both Mr. Lochbaum and I, and for that matter, many others who would like to see different conclusions, have searched for contrary evidence. To the best of my knowledge, none has been found.

k This is not to say that the plant should necessarily be shut down.

I believe that very credible arguments can be, made for a J.I.O. I therefore don't understand why there is such an apparent reluctance in the organization to acknowledge these concerns and move ahead with resolution expeditiously. Although resolution will have a cost, certainly, that cost does not necessarily have to include plant shutdown.

I therefore strongly urge that the formal screening evaluation and the evaluations of the operability and reportability of these concerns proceed without, further delay with priority over all other activities in Mr. Miller s plan, and that we expeditiously get on with the process of r'esolving these concerns.

G. T. Jones August 20, 1992 Fuel Pool Cooling Deficiencies Page 6 I. sincerely appreciate your continued personal attention in these matters, and I am at your service in addressing these concerns.

DISTRIBUTION:

J. E. Agnew A6-3 G. D. Miller A6-3 F. G. Butler A6-3 J. R. Mittenberger A6-1 R. G. Byram A6-1 M. R. Mjaatvedt A6-3 M. H. Crowthers A6-3 C. A. Myers A2-4 G. D. Gogates SSES J. G. Refling A6-3 J. S. Stefanko A9-3 G. J. Kuczynski SSES, S&A T. J. Sweeney SSES D. A. Lochbaum Enercon J. A. Zola A6-3 FUELPOOL.DCP/kbw

Attachaent 11 PP8L Memo from A. Dyszel to T. C. Dalpiaz, "U2 RI05 Fuel Pool Decay Heat Evaluation", August 21, 1992

( PL I-72230)

Note: This letter transmits interim guidance to the SSES site personnel for use dur ing an upcoming refueling outage.

This guidance is necessary because EDR G00005, initiated in September 1990, has not yet been dispositioned and the applicable discussions in FSAR Section 9.1 and Appendix 9A are no -longer accurate.

18-22-1992 15:57 P.82z84

'Uegest Z3g 3992 T. t. Oalpfaz SSES l'...- g<> ~.!>r

~ sg'gqh e

lily >(br SUSgUEHNNA STGN ELECTRIC STATION 02 RI05 FUEL POOL DECA'I HEAT BALNTIOH CCH 741687 F ILEeAI~"

w7

References:

I} PLI-8?533, "Fuel Pool Cooling Heat Loads - KOR IZ, 1991.

600005,'prfl

2) PLI-70395, 'Ul 8106 Fuel Pool Decay Heat Evaluation,"

February 7, 1992.

3} ET-0585, "Fuel pool Caalfng Oeffciencfes," 8/18/92.

This memo provides short tare relief for the open EOR, 600005, {Ref<<rence 1}

with respect to Che upcoofng N SRIO by providing NFK's evaluation af the resultfna Cise constraints for perforaance af coaaoa RHR system outage work, similar 4 the U1 RI05 evaluation Neference 2}. A long tem solutian to. this EOR is required to ensure future successful, Cfeely auiages. fn short, the U2 5RID evaluatian shows na change in ihe current outage schedule (f. ~ ., Sept. 28 for coamon RHR systei work} for na fuel failures in Qnft 8 Cycl ~ S. However, future evaluations, which will involve higher heat loads and address Che potential for fuel failure(s), will likely affect the outage schedule.

Reference 3 provides a descr)ptian of the long Cere solutfan to this problem.

3QLX FSAR Section S.A describes the radiological release results froa a loss af fuel poal cooling event. 111 'ssuaptfons used for Che FSNl analysis include 4 care reloads, in care fuel shuffling, and a aaxfeua fuel expasure of 28,500 ND/NTM. The current aperation at Susquehanna SES includes a fuel reload batch size ot aaoroxfiately W of ihe core, a maxfeul fuel discharge expasure af l0,001 i%8/OU, and a full core offload for each outage. If one ar more fuel fef lures ere suspected io have occurred during the pperating cycle just prfar ta an outage or durino fuel handling aft<<r shutdown, analyses-oust b<<

performed to assure the radio'logical release free the pastulated lass of fuel pool coalfng event are less than those presented in Che FSAR. The FSAR analyses are bounding provided that cateeon RHR systee outage work is not started until the decay heat lave'l is low enough to prevent fuel pool hailing fn less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. If na fuel failures are suspected to have occurred during ihe op<<retina cycle just prior to an outage or during fuel handling after shutdown, analyses erst be performed to assure the fuel pool water ievel can be aafntafned during a lass of fuel pool cooling event. The fuel pool

iO-22-iGQ2 i5:58 P. 03r84 water level Caa he eaiatatae4 previded that hOC Starters util e~ NO Sgatm eugene Wrh ia Qe-CN melee'lot pate te the speht fuei peel tl e.o 40 CPll) is greater than the spent fuel pool boiling rate derring the loss of fuel pool coo'l ing event.

To address this issue for the upcoming outage. Nuclear Fuels Engineering has calculated the total decay heat of the fuel in the spent fue) pools for the U2 5RIQ consistent with the approach in Reference 2. This decay heat leva) includes a total of the decay heat froo the fuel in the UI pool, U2 pool, and the full core offload. F)gure l shows the calculated spent fue) pool decay heat as a function of time after shutdown. The curve labelled 'nominal's a calculation of the decay heat based on the methodolag'y in NlRKG-OSQO. This methodology has been shown to produce higher decay heat levels than the more rigorous methodology in the AHS S.S-I919 decay heat standard. The curve labelled "maximum's also based on the methodo'logy in NUREG-0800 but accounts for uncertainty in the reactor power level (lo} and uncertainty in the decay heat methodology.

Nuclear Fuels Eng1neerino has also performed a calculation to determine. Che Cfme after the U2t3 shutdown that the spent fuel pool boiling rate during a postulated loss of spent fuel gal cooling event 1s less than the 60 GN EN makeup flow rate. Based on Che "aaximum" decay heat curve in f1gure I. a spray pond teeperature of SO'F, and open fuel pool gates. the fuel pool boil1ng rate during a postulated loss of fuel pool cooling event is less than the 60 OPN ESM makeup raCe subsequent to )4 days after reactor shutdown.

Therefore, HFE's evaluation indicates that providid fuel fa11ure does not occur during the remaining V2C5 operation or during fuel handling after shutdown, coslon RHR system outage work should not start unC11 at least 14 days after reactor shutdown. [f a fuel fa1lure occurs, a calcu1ation should be performed to determine if furCher outage restrict1ons are necessary. Note that the above calculations have been documented 1n NFE-B-NA-OhS. Rev. 2 and independently reviewed in accordance with qA procedures EN-N-301.

The )4 da restr1ction on cottisencement of coamon RHR system work does not impact the current UR IRKED outage schedule.

A. Dysxel Nuclear fuel management Pro)oct Engineer Nuclear Fuels Engineering AD/el mel311a.ad cc: J, E. Agnew SSES A. J. Rosc1oli A9-3 K. Q. Harwanko SSES R. A. Saccone SSES G. T. Jones AS-2 J. P. Spadaro AS 3 J. N. telic k AS 3 J. 5. Stefanko A9-3 C. R. Lehmann Ag-3 Zol a A5-3 G. Da Hiller AS~3 NR File A5'2

G I

IV RGURE 1 IV I

U2 5RIO SPENT FUEL POOL DEGAY HEAT fU g1 PLUS U2}

~ ~

g ~ ~ 0 ~

~ Q ~

'0

~ r b

so b DAYS AFTER HOT SHU7DQVW

~

y li

AttachIIent 12 PP8L Memo from J. M. Kenny to G. T. Jones and C. A.

Myers, "EDR on Fuel Pool Cooling", August 25, 1992 Note:. This 'confidential'emo is the first documented indication that the NRC had been informally notified of the concerns raised in EDR G20020.

~ ~ ~

QQN F1QQQTIAL August 25, 1992 G. T. Jones A6-2 C. A. Myers A2-4 EDR ON FUEL POOL COOLING On August 24, 1992 I briefed both Scott Barber and Jim Raleigh of the NRC on the status of our review of contractor originated fuel pool cooling concerns documented on an EDR. I noted that our current position was there were no immediate concerns with system operability or need for reportability under regulations identified but that our efforts were continuing to address the identified issues. I also noted that George Jones had discussed the concerns with the contractors and was personally involved in resolving the issues.

I had previously briefed Jim Raleigh in July of the fuel pool concerns and reviews being performed; Scott did bring to my attention an open inspector finding concerning the Haddam

¹ck fuel pool draindown event and subsequent efforts by NSAG on fuel pool issues. He noted there were 28 open items and that we should review these issues for status. I indicated it was my understanding Engineering would be addressing the NSAG open issues on the fuel pool as part of their effort to resolve the open EDR.

k%

M. Kenny CC: J. E. Agnew A6-3 G. D. Miller A6-3 J. R. Miltenberger A6-1 R. R. Sgarro A2-4 H. G. Stanley SSES JMK:tah FuelPool. EDR

Attachment 13 PP8L Memo from George T. Jones to Glenn D. Miller, "Fuel Pool Cooling EDR's 620020, G00005", August 27, 1992 (PLI-72267)

~ ~ ~

R pp p <<y<<p p; .,0 p <<{

August 27, 1992 Glenn D. Miller A6-3 SUSQUEHANNA STEAM ELECTRIC STATION FUEL POOL COOLING EDR'S G20020, G00005 PLI-72267 FILE'45-I A

Reference:

ET-0587, ET-0586 The reference letters show me that there still is a difference of professional opinion relative to the significance of the subject EDR's. This difference is to be resolved. This issue is to be worked expeditiously until we have resolved the outstanding questions.

I wish to point out that our belief that a design has been reviewed and approved by NRC, is not adequate justification for classification of the significance of an issue. The issue must stand on it's own merits.

I am further concerned that our actual configuration and method of operation apparently differs from that described in the FSAR.. The FSAR is our licensed bases and any deviation from that description is required to have a thorough and complete documented evaluation on file.

There were twenty-eight open items resulting from NSAG Review of Fuel Pool Cooling. These need to be included in this revie'w.

I wish to have the schedule for resolution of this issue accelerated and the classification of the significance of this issue reevaluated. I am expecting at least daily updates of our progress.

If you have questions, please call me.

>AM George . ones Attachment CC: J. E. Agnew A6-3 w/a D. C. Prevatte A6-3 w/a F. G. Butler A6-3 w/a J. R. Miltenberger A6-I w/a R. G. Byram A6-I w/a M. R. Hjaatvedt A6-3 w/a M. H. Crowthers A6-3 w/a C. A. Meyers A2-4 w/a G. D. Gogates SSES w/a J. G. Ref 1 ing A9-3 w/a G. J. Kuczynski SSES, SM w/a T. J. Sweeney SSES w/a D. A. Lochbaum Enercon w/a J. A. Zola A6-3 w/a S. M. Hauseman A6-2 w/a Nuc. Rec. Files A6-2 w/o c:Xvp51%docs%2002M5.sgt 8/26/92

Attachaent 14 PP8L Memo from Glenn 0. Miller to George T. Jones, "Fuel Pool Cooling EDRs 620020, G00005", August 31, 1992 (PLI-72297)

~ ~

August 31, 1992 George T. Jones A6-2 SUSQUEHANNA STEAH ELECTRIC STATION FUEL POOL COOLING EDRs G20020, G00005 PL I-72297 FILE A45-I

Reference:

PLI-72267, ET-0587, ET-0586 In response to your letter PLI-72267 we are continuing to work to resolve the issues in the referenced EDRs. As I explained to you previously I met with Hr. Prevatte and discussed his concerns at length on August 21. We acknowledged our differences and agreed to continue working toward resolution.

I want to reemphasize to you that we are not using the prior review and approval of our system design by the NRC as a basis for the classification of the safety significance of this issue. Statements which I made in ET-0586 relative to the safety significance were intended to summarize the EDMG position on screening.

The reference to the NRC is only a statement of the fact that our design chilosophy at the time of licensing was reviewed and approved by the NRC. This position in fact constitutes our licensing basis.

The FSAR contains references to analyses regarding the Fuel Pool Cooling and Cleanup System and the Fuel Pool. It does not describe the exact manner in which we operate'the plant. Our current fuel design and fuel cycle deviates from the FSAR description. This is the subject of G00005. Our outage practices differ from the description in the FSAR. The current outage practice is the subject of a periodic analysis done by Nuclear Fuels for each refueling outage.

We have reviewed the twenty-eight "open items" from the NSAG review. Twenty-six of the twenty-eight items were resolved to the satisfaction of NSAG. The remaining two items refer to the need for the level and temperature indication to be available in the control room via a PHS (computer display) format and to add reflash capability for alarms from panel OC211. These modifications are on the books but not being actively worked to the best of my knowledge.

EDMG completed a revision to the EDR screening for G20020 on Friday August 28, 1992 and requested comments. The significance was evaluated as minimal. Based on my review of this screening document it is unacceptable as written and I have requested it be revised. We are proceeding to review the issue for reportability regardless of the final significance level from the screening review.

Work on resolving the issue is assigned to Hark Hjaatvedt. Michael Crowthers has been working on this issue since July 20, 1992. I have also assigned Dave Kostelnik as of today. We are planning to involve Bechtel and GE. A schedule is under development.

August 31, 1992 PL I-72297 FILE A45-1A Systems Analysis'ndependent evaluation will be completed this week. We will factor their evaluation into our ongoing work. I will report on their results when available.

I will continue to keep you apprised of our progress on a daily basis.

Glenn D. Miller cc: J. E. Agnew A6-3 D. C. Prevatte A6-3 F. G. Butler A6-3 J. R. Miltenberger A6-1 R. G. Byram A6-I M. R. Mjaatvedt A6-3 M. H. Crowthers A6-3 C. A. Myers A2-4 G. D. Gogates SSES J. G. Ref ling A6-3 G. J. Kuczynski SSES J. S. Stefanko A9-3 58@RZcdg~ayyp '.- ~4'a'.~U~ *~

":iS45con'6-2 T. J. Sweeney SSES S. M. Hauseman J. A. Zola Nuclear Records A6-2

Attachment 15 PP8L Memo from Kevin M. Brinckman to George T. Jones, "Review of Fuel Pool Cooling", September 1, 1992 (PLI-72288)

Note: This engineering repor t was prepared by a PPAL engineer previously not associated with EOR G20Q20 at the request of the PP8L Manager of Nuclear Plant Engineering to provide him with an independent appraisal of the concerns raised in the EOR. This independent evaluation basically concludes that a LOCA with a loss of normal fuel pool cooling would put the operators "in a position where they would be required to make decisions on removing ECCS equipment from containment/core cooling service to cool the fuel pool" and points out that conditions.

it would involve unanalyzed This report also raises, for the first time, the concern that the hydrodynamic loads of the LOCA might damage the non-seismic, non-safety related fuel pool cooling system piping.