ML18026A537

Forwards FSAR Change Delineating Use of RHR Fuel Pool Cooling Mode of Operation to Mitigate Loss of Normal Spent Fuel Pool Cooling Sys in Response to Seismic Event,Per Commitment Made Via 941228 Ltr

ML18026A537
Person / Time
Site:	Susquehanna
Issue date:	02/21/1995
From:	Byram R PENNSYLVANIA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
PLA-4273, NUDOCS 9502270258
Download: ML18026A537 (56)

Text

PR.I(3RI EY (ACCELERATED RIDS P ROCESSIX 1

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9502270258 DOC.DATE: 95/02/21 NOTARIZED: NO DOCKET N FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION BYRAM,R.G. Pennsylvania Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards FSAR change delineating use of RHR fuel pool cooling mode of operation to mitigate loss of normal spent fuel pool cooling sys in response to seismic event,per commitment made DISTRIBUTION CODE: AOOID TITLE:

NOTES:

OR via 941228 ltr.

COPIES RECEIVED:LTR Submittal: General Distribution l ENCL 3 SIZE: 5 05000387 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PDl-2 LA 1 1 PD1-2 PD 1 1 POSLUSNY,C 1 1 INTERNAL: ACRS 6 6 1 1 NRR/DRCH/HICB NRR/DSSA/SRXB 1

1 1

1

/D /

NUDOCS-ABSTRACT 1

1 1

1 OGC/HDS2 1 0 EXTERNAL: NOAC NRC PDR 1 1 NOTES: 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE O'ASTE! CONTACTTIIE DOCL'ifENTCONTROL DESk, ROOAI PI-37 (EXT. 504-0033 ) TO ELI XIINiATEYOUR NAiILFROif DISTRIBUTION LISTS I'OR DOCI.'4 IEN'I'S YOL'ON"I'L'I'.D!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 17

~

Pennsylvania Power 0

8 Light Company Two North Ninth Street ~ Allentown, PA 18101-1179 ~ 610/774-5151 FEB 21 1995 Robert G. Byram Senior Vice President Nuclear 610/774-7502 Fax: 610/774-5019 U.S. Nuclear Regulatory Commission Attn.: Document Contr'ol Desk Mail Station P 1-137 Washington, D. C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION FSAR CHANGE: RHR FUEL POOL COOLING Docket Nos. 50-387 and 50-388

Reference:

PLA-4230, RG. Byram to UShfRC, "Loss ofSpent Fuel Pool Cooling from Seismic Event/Use ofRHR Fuel Pool Cooling Mode", dated December 28, 1994.

Dear Sir:

Via the referenced letter, PP&L committed to provide an FSAR change delineating the use of the RHR Fuel Pool Cooling mode of operation to mitigate the loss of normal spent fuel pool cooling system in response to a seismic event. A copy of the FSAR change is attached for your use and information.

Although this change has been reviewed and approved internally, PP&L is treating this change as preliminary pending issuance of the final NRC Safety Evaluation on Spent Fuel Pool Cooling issues.

Upon issuance of that document, we will resolve any discrepancies and formally issue the FSAR change per our normal procedures.

Ifyou have any questions on the attachment, please contact Mr. J.M. Kenny at (610) 774-7904.

Very truly yours, i

R. G. yra Attachment CC: NRC Region I Ms. M. Banerjee, NRC Sr. Resident Inspector - SSES Mr. C. Poslusny, Jr., NRC Sr. Project Manager - OWFN Mr. J. Shea, NRC Project Manager - OWFN 950227 I DR ADOCK 05000>8 PDP

NOTE: Page numbers in parenthesis ( I indicate a spill-over from previous pages. They do not coincide with the text on that page in the FSAR.

...9502270258

SSES-FSAR average life expectancy many times the residence time of a fuel loading.

1.2.2.3.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structure; the steam separators and dryers; the jet pumps; the control rod, guide tubes; distribution lines for the feedwater, core spray, and standby liquid control; the incore instrumentation; and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, the feedwater lines, the control rod drive housings, and the ECCS lines.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure is 1020 psia in the steam space above the separators. The vessel is fabricated of carbon steel and is clad internally with stainless steel (except for the top head which is not clad).

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers, located in the upper portion of the reactor vessel. The steam is then directed to the turbine through four main steamlines. Each steamline is provided with two isolation valves in series, one on each side of the primary containment barrier.

1.2.2.3.3 Reactor Recirculation S stem The Reactor Recirculation System pumps reactor coolant through the core to remove the energy generated in the fuel. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. Each loop has one motor-driven recirculation pump. Recirculation pump speed can be varied to allow some control of reactor power level through the effects of coolant flow rate on moderator void content.

1.2.2.3.4 Residual Heat Removal S stem The Residual Heat Removal System (RHRS) consists of pumps, heat exchangers and piping that fulfill the followingfunctions:

a ~ Removal of decay heat during and after plant shutdown.

b. Rapid injection of water into the reactor vessel following a loss of coolant accident, at a rate sufficient to reflood the core maintain fuel cladding below the limits contained in 10 CFR 50.46. This is discussed in Subsection 1.2.2.4.

Rev. 47, 06/94 1.2-14

SSES-FSAR c ~ Removal of heat from the primary containment following a loss-of-coolant accident (LOCA) to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the water inside the primary containment. The redundancy of the equipment provided for the containment is further extended by a separate part of the RHRS which sprays cooling water into the drywell. ,This latter capability is discussed in Subsection 1.2.2.4.12.

d. Provide for cooling of the spent fuel pool(s) following a seismic event which results in a loss of normal spent fuel pool cooling, in conjunction with normal shutdown of both units.

1.2.2.3.5 Reactor Water Cleanu S stem RWCU A Reactor Water Cleanup System, which includes a filter demineralizer, is provided to clean up the reactor cooling water, to reduce the amounts of activated corrosion products in the water, and to remove reactor coolant from the nuclear system under controlled conditions.

1.2.2.4 Safet Related S stems Safety related systems provide actions necessary to assure safe shutdown, to protect the integrity of radioactive material barriers, and/or to prevent the release of radioactive material in excess allowable dose limits. These systems may be components, groups of components, systems, or groups of systems. Engineered Safety Feature (ESF) systems are included in this category. ESF systems have a sole function of mitigating the consequences of design basis accidents.

1.2.2.4.1 Reactor Protection S stem The Reactor Protection System initiates a rapid, automatic shutdown (scram) of the reactor. This action is taken in time to prevent excessive fuel cladding temperatures and any nuclear system process barrier damage following abnormal operational transients. The Reactor Protection System overrides all operator actions and process controls.

Rev. 46, 06/93 1.2-15

SSES-FSAR 1.2.2.4.2 Neutron-Monitorin S stem Not all of the Neutron Monitoring System qualifies as a nuclear safety system; only those portions that provide high neutron flux signals to the Reactor Protection System are safety related. The intermediate range monitors (XRM) and average power range monitors (APRM), which monitor neutron flux via in-core detectors, signal the Reactor Protection System to scram in time to prevent excessive fuel clad temperatures as a result of abnormal operational transients.

Rev. 46, 06/93 1.2-(16)

SSES-FSAR 3.1.2.1.5 Sharing of Structures, Systems, and Components Criterion 5 Criterion Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

Desi n Conformance Although Susquehanna SES Units 1 and 2 share certain structures, systems, and components, sharing them does not significantly impair performance of their safety functions.

The following safety related structures are shared between both units:

Control Structure Diesel Generator Buildings ESSW Pumphouse Spray Pond Spent Fuel Pools The safety related structures are designed to remain functional during and, following the most severe natural phenomena.

Therefore sharing these structures will not impair their ability to perform their safety functions.

Seismic Category I 'tructures which house safety related systems and equipment are discussed in Section 3.8.

The shared systems which are important. to safety are discussed below; a more detailed discussion - may be found in the referenced Subsections:

a) Emergency Service Water System (ESWS) 9.2.5 b) Diesel Generators 8.3.1.4 c) Ultimate Heat Sink '(Spray Pond) 9.2.5 SE 9.2.6 d) Offsite Power Supplies 8.2 e) Unit 1 AC Distribution System 8.3.1 f) Residual Heat Removal 5.4.7.1.1 '

(Fuel Pool Cooling Mode)

Rev. 47, 06/94 3.1-6

, ~

1 SSES-FSAR Emer enc Service Water S stem ESWS The ESWS is designed to a) Supply cooling water to the RHR pumps and their associated room coolers during the several non-emergency modes of RHR pump operation such as normal shutdown, and hot standby.

b) Supply cooling water to the various diesel generator heat exchangers, RHR pumps, room coolers, RBCCW and TBCCW heat exchangers during emergency shutdown conditions such as a LOCA.

c) Supply cooling water to the RHR pumps and their associated room coolers during a seismic event that results in a loss of the non-seismic Category I Fuel Pool Cooling System. During this event, ESWS would also supply water to the spent fuel pools to make-up for evaporative losses and proper level needed to fill the spent fuel pools to the support the RHRFPC mode should the normal make-up source be unavailable.

The ESWS pumps are located in the ESWS pumphouse with the RHRSW pumps. The ESWS pumphouse is designed as Seismic Category I and the ESWS consists of two redundant loops (denoted A and B) each capable of providing 100 percent of the cooling water required by all the ESF equipment of both Units 1 and 2 simultaneously.

The system is designed so that no single active or passive component failure will prevent it from achieving its safety related objective.

The system starts automatically on a diesel start signal.

For additional discussion, see Subsection 9.2.5.

Diesel Generators Diesel Generators A, B, C and D are housed in a Seismic Category I structure. They are separated from each other by concrete walls which provide missile protection. Additionally, a spare diesel generator (Diesel Generator 'E') is provided which can be manually realigned as a replacement for any one of the other four diesel generators. Thus, any one of the other diesel generators (A, B, C or D) can be removed from service for extended maintenance and the Diesel Generator 'E'an be substituted so that there are four operable diesel generators.

Diesel Generator 'E's housed in its own Seismic Category I structure which also provides missile protection. Loss of one of the four aligned diesel generators will not impair the capability to safely shutdown both units, since this can be Rev. 47, 06/94 3 0 1 7

SSES-FSAR done with three diesel generators. For additional discussion, see Subsection 8.3.1.4.

For descriptions of the Diesel Generator Fuel Oil System, Cooling Water System, Air Starting System, Lube Oil System, and the Intake and Exhaust Systems see Subsections 9.5.4, 9.5.5, 9 '.6, 9.5.7, and 9.5.8 respectively.

For missile protection see Subsection 3.5. Separation is discussed in Sections 3.12 and 8.3.

Ultimate Heat Sink S ra Pond The spray pond provides the water for both the ESWS system and the RHRSW systems. It is the ultimate heat sink for both Units 1 and 2. The return lines from the ESWS and the RHRSW are combined and the total quantity of water from both these systems is discharged through spray networks, which dissipate the heat back to the pond. There are two redundant return loops (A and B); either one is capable of handling the full flow from the ESWS and RHRSW when shutting down two units simultaneously.

Each return loop supplies a separate spray network and each, of these networks is divided into a large one capable of dissipating the heat from the ESWS and the RHRSW from the RHR heat exchanger on one unit, and a smaller one capable of dissipating the heat from the RHR heat exchanger on the second unit.

The spray pond contains sufficient water to meet the requirements for shutting down one unit in the event of an accident and to permit the safe shutdown of the second unit for a period of thirty days without makeup.

For additional discussion see Subsections 9.2.5 and 9.2.6.

Offsite Power Su lies The two preferred offsite power supplies are shared by both units. The capacity of each offsite power supply is sufficient to operate the engineered safety features of one unit and safe shutdown loads of the other unit.

For additional discussion, see Section 8.2.

Unit 1 AC Distribution S stem The Unit 1 AC Distribution System is a shared system between both units, since the common equipment (Emergency Service Water, Standby Gas Treatment System, Control Structure HVAC, etc.) is energized only from the Unit 1 AC Distribution System.

Rev. 47, 06/94 3.1-8

SSES-FSAR There are no Unit 2 specific loads energized from the Unit 1 AC Distribution System. The capacity of the Unit 1 AC Distribution System is sufficient to operate the engineered safety features on one unit and the safe shutdown loads of the other unit.

Residual Heat Removal Fuel Pool Coolin Mode)

With the Spent Fuel Pools crosstied, one unit's RHR system can be used to cool stored spent fuel in both spent fuel pools. In the crosstied configuration, the RHRFPC mode of one unit will draw suction from that unit's skimmer surge tank and return the cooled flow to the bottom of the unit's fuel pool. No direct flow to or from the opposite unit's fuel pool will be accomplished. With the pools crosstied and RHRFPC in operation on one of the units adequate cooling of both pools will be achieved. For further discussions see Subsections 5.4.7.1.1.6, 5.7.2.1,c, 9.1.3.1c, and 9.1.3.3.

3.1.2.2 Protection by Multiple Fission Product Barriers Qrou II 3.1.2.2.1 Reactor Desi n Criterion 10 Criterion The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not Rev. 47, 06/94

SSES-FSAR

3) In-service Inspection 5.2

4) Reactor Vessel and Appurtenances 5.4

5) Reactor Recirculation System 5.4 3.1.2.4.4 Reactor Coolant Makeup Criterion 33 Criterion A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps,,

and valves used to maintain coolant inventory during normal reactor operation.

Desi n Conformance The plant is designed to provide ample reactor coolant makeup for protection against small leaks in the RCPB for anticipated operational occurrences and postulated accident conditions. The design of these systems meets the requirements of Criterion 33.

For further discussion, see the following sections:

Reactor Coolant Pressure Boundary Leakage Detection Systems 5.2

2) Reactor Core Isolation Cooling System 5.4

3) Emergency Core Cooling System 6.3

4) Reactor Vessel Instrumentation and Control 7.6

5) Makeup Demineralizer System 9.2

6) Condensate Storage and Transfer System 9' 3.1.2.4.5 Residual Heat Removal Criterion 34 Criterion A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Rev. 46, 06/93 3.1-39

SSES-FSAR Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Desi n Conformance RHR system provides the means to remove decay heat and residual heat from the nuclear system so that refueling and nuclear system servicing can be performed.

Major RHR system equipment consists of two heat exchangers and four main system pumps. The equipment is connected by associated valves and piping, and the controls and instrumentation are provided for proper system operation.

Two independent loops are located in separate protected areas.

The RHR system is designed for four modes of operation:

a) Shutdown cooling b) Suppression pool cooling (also containment spray) c) Low pressure coolant injection.

d) Fuel Pool Cooling Both normal ac power and the auxiliary onsite power system provide adequate power to operate all the auxiliary loads necessary for, plant operation. The power sources for the plant auxiliary power system are sufficient in number, and of such electrical and physical independence that no single probable event could interrupt all auxiliary power at one time.

The plant auxiliary buses supplying power to engineered safety features and reactor protection systems and auxiliaries required for safe shutdown are connected by appropriate switching to the four aligned standby diesel-driven generators located in the plant. Each power source, up to the point of its connection to the auxiliary power buses, is capable of complete and rapid isolation from any other source.

Loads important to plant operation and safety are split and diversified between switchgear sections, and means are provided for detection and isolation of system faults.

The plant layout is designed to effect physical separation of essential bus sections, standby generators, switchgear, interconnections, feeders, power centers, motor control Rev. 46, 06/93 3.1-40

SSES-FSAR centers, and other system components.

Four standby diesel generators (A, B, C, and D) and a spare diesel generator (E), which can be manually realigned as a replacement for any one of the other four diesel generators are provided. These diesel generators supply a source of electrical power which is self-contained within the plant and is not dependent on external sources of supply. The standby generators produce ac power at a voltage and frequency compatible with the normal bus requirements for essential equipment within the plant. The standby diesel generator system is highly reliable. Any three of the five generators are adequate to start and carry the essential loads required for a safe and orderly shutdown.

The RHR system is adequate to remove residual heat from the reactor core to ensure fuel and RCPB design limits are not exceeded. Redundant reactor coolant circulation paths are available to and from the vessel and RHR system. Use of RHR in the Fuel Pool Cooling mode will not adversely impact the ability of RHR to perform Reactor Core Cooling functions as discussed in Subsections 5.4.7.1.1.6, 5.4.7.2.6c, 9.1.3.1c, and 9.1.3.3. Redundant onsite electric power systems are provided.

The design of the RHR system, including its power supply, meets the requirements of Criterion 34.

For further discussion, see the following sections:

1) Residual Heat Removal System 5.4

2) Emergency Core Cooling Systems 6.3

3) Emergency Core Cooling Systems Instrumentation and Control 7'

4) Auxiliary Power System 8.3

5) Standby ac Power Supply and Distribution 8.3

6) Station Service Water 9.2

7) Accident Analysis 15.0 3.1.2.4.6 Emergency Core Cooling Criterion 35 Criterion A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Rev. 46, 06/93 3.1-41

SSES-FSAR Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Rev. 46, 06/93 3. X- (42)

SSES-FSAR Desi n Conformance The emergency safeguard service water system, which comprises both the Emergency Service Water system and the Residual Heat Removal Service Water system, provides cooling water for the removal of excess heat from all structures, systems, and components which are necessary to maintain safety during all abnormal and accident conditions. These include the standby diesel generators, the RHR pump oil coolers and seal water coolers, the core spray pump room unit coolers, RCIC pump room unit coolers, the HPCI pump room unit coolers, the RHR heat exchangers, RHR pump room unit coolers, emergency switchgear and load center room coolers and the control structure chiller.

It also provides water to the RHR pumps and above mentioned room unit coolers during a seismic event to support operation of the RHR Fuel Pool Cooling (RHRFPC) mode. Make-up water to the Spent Fuel Pool (SFP) is provided during a seismic event in order to make-up for evaporative losses and filling of the SFP in support of RHRFPC. RHRSW provides the cooling water to the RHR heat exchangers for the RHRFPC mode.

The engineered safeguard service water system is designed to Seismic Category I requirements. Redundant safety related components served by the engineered safeguard service water system are supplied through redundant supply headers and returned through redundant discharge or return lines. Electric power for operation of redundant safety related components of this system is supplied from separate independent offsite and redundant onsite standby power sources. No single failure renders these systems incapable of performing their safety functions.

Referenced Subsections are as follows:

1) ac Power Systems 8.3.1

2) Service Water System 9.2.1

3) Engineered Service Water System 9.2.5

4) RHR Service Water System 9.2.6

5) Ultimate Heat Sink 9.2.7 3.1.2.4.16 Inspection of Cooling Water System Criterion 45 Criterion The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

Rev. 46, 06/93 3.1-51

SSES-FSAR Desi n Conformance The engineered safeguard service water and the RBCCW systems are designed to permit appropriate periodic inspection in order to ensure the integrity of system components.

Rev. 46, 06/93 3.1- (52)

SSES-FSAR Desi n Conformance New Fuel Stora e New fuel is placed in dry storage in the new fuel storage vault that is located inside the reactor building. The storage vault within the reactor building provides *adequate shielding for radiation protection. Storage racks preclude accidental criticality (see Subsection 3.1.2.6.3). The new fuel storage racks do not require any special inspection and testing for nuclear safety purposes. However, the racks are accessible for periodic inspection.

S ent Fuel Handlin and Stora e Irradiated fuel is stored submerged in the spent fuel storage pool located in the reactor building. Fuel pool water is circulated through the fuel pool cooling and cleanup system to maintain fuel pool water temperature, purity, water clarity, and water level. Storage racks preclude accidental criticality (see Subsection 3.1.2.6.3).

Reliable decay heat removal is provided by the fuel pool cooling and cleanup system. The pool water is circulated through the system with suction taken from the pool and is discharged through diffusers at the bottom of the fuel pool.

Pool water temperature is maintained below 125'F when removing the maximum normal heat load (MNHL) from the pool with the service water temperature at its maximum design value. The RHR system with its substantially larger heat removal" capacity can be used as a backup for fuel pool cooling when heat loads larger than the capability of the fuel pool cooling systems are in the spent fuel pools.

RHR also provides reliable decay heat removal to the spent fuel pools if the normal fuel pool cooling system is lost due to a Seismic event. Operation of the RHR Fuel Pool Cooling (RHRFPC) mode will provide seismic Category I, Class 1E cooling to the spent fuel pools so that boiling of the spent fuel pools does not occur as a result of a seismic event. ESW provides Seismic Category I, Class 1E make-up in support of RHRFPC.

High and low level switches indicate pool water level changes in the main control room. Fission product concentration in the pool water is minimized by use of the filters and demineralizer. This minimizes the release from the pool to the reactor building.

Rev. 46, 06/93 3.1-60

SSES-FSAR The reactor building ventilation system and the secondary containment are designed to limit the release of radioactive materials to the environs and ensure that offsite doses are less than the limiting values specified in 10CFR100 during operation and all accident conditions.

No special tests are required, because at least one pump and heat exchanger are continuously in operation while fuel is stored in the pool. Duplicate units are operated periodically to handle high heat loads or to replace a unit for servicing.

Routine visual inspection of the system components, instrumentation, and trouble alarms are adequate to verify system operability. Testing of the RHRFPC mode is accomplished through routine testing of the pumps and heat exchangers in support of other modes of RHR. The valves supporting the RHRFPC mode are routinely stroked to confirm proper operation of the valves for their RHRFPC mission.

Rev. 46, 06/93 3. 1- (61)

SSES-FSAR fuel. These interlocks preclude any load suspended from this crane from tipping over on the stored fuel in the event of a crane failure. The 5 ton auxiliary hook suspended from the same crane trolley is prevented from passing over stored fuel when fuel handling is not in progress by administrative controls'here are no planned transfe'rs of loads heavier than a new fuel element over the stored fuel.

(3)

Reference:

Position C.8. A Seismic Category I makeup water supply from each emergency service water loop is permanently connected to,each spent fuel pool by two independent Seismic Category I piping routes.

The make-up is provided for filling the -spent fuel pool to the proper level to support operation of the RHR fuel pool cooling mode, and to provide for make-up from evaporative losses during cooling by RHR.

The make-up rate is sized based on boiling so as to be conservative. The normal makeup system to the fuel pool is not Seismic Category I.

Re ulato Guide 1.14 REACTOR COOLANT PUMP FLY-WHEEL INTEGRITY Revision 1 Au ust 1975 Not applicable.

Re ulator Guide 1.15 - TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES Revision 1 December 28 1972 Testing of reinforcing bars for Category I concrete structures is in compliance with this regulatory guide.

Re ulator Guide 1.16 - REPORTING OF OPERATING INFORMATION-APPENDIX A TECHNICAL SPECIFICATIONS Revision 4 Au ust 1975 In lieu of the positions stated in this Regulatory Guide, the reporting of operating information for the Susquehanna SES complies with Technical Specifications and 10CFR50.73.

Re ulator Guide 1.17 PROTECTION OF NUCLEAR POWER PLANTS AGAINST INDUSTRIAL SABOTAGE June 1973 In lieu of the positions stated in this regulatory guide, the protection of Susquehanna SES against industrial sabotage complies with 10CFR73.

Rev. 46, 06/93 3.13-6

SSES-FSAR

Reference:

Position C.l.d and C.l.g. The normal spent fuel pool cooling system is non-seismic Category I. If a seismic event would occur cooling of the spent fuel is achieved by use of the RHR Fuel Pool Cooling (RHRFPC) mode as described in sections 5.4.7.1.1 ', 5.4.7.2.6c, 9.1.3.1, and 9.1.3.3. Either or both of two Seismic Category I ESW makeup water supplies to each pool can provide make-up in support of the RHRFPC mode.

Additionally, ESW is capable of supplying make-up for the boiling spent fuel pool analysis as described in Appendix 9A.

Reference:

Position C.l.e. The Main Steam System (MSS) beyond the outer isolation valves up to and including the turbine stop valves and all branch lines 2 1/2 in. in diameter and larger, up to and including the first valve (including their restraints) are not classified Seismic Category I; because portions of the pipe are routed in a non-Seismic Category I building (the Turbine Building). However, the turbine building has been Subsection 3.7b.2 '.

designed to withstand an SSE as stated in Further description of the turbine building is given in Subsection 3.8.4.1; applicable load combinations are given in Table 3.8-10. The subject piping is designed in accordance with ASME Section III, Class 2 requirements for the OBE and SSE as described in Subsection 10.3.3.

Reference:

Position C.l.h. The component cooling water portions of the reactor recirculation pumps are not Seismic Class I since they do not involve a safety function.

Reference:

Paragraph C.2 of the Regulatory Guide.

Items which would otherwise be classified non-seismic category I, "but whose failure could reduce the functioning" of items important to safety "to an unacceptable safety level" are to be "designed and constructed so that the SSE would not cause such failure." In addition, Paragraph C.4 of the guide requires that the "pertinent quality assurance requirement of Appendix B to 10 CFR Part 50 should be applied to the safety requirements" of such items. Both of'hese positions are considered to be adequately met by applying the following practices to such items:

06/93 3.13-10

SSES-FSAR (a) Design and design control for such items are carried out in the same manner as that for items directly important to safety. This includes the performance of appropriate design reviews.

Rev. 46, 06/93 3. 13- (11)

SSES-FSAR TABLE 3.2-1 Continued) Page 9 .

Principal Quality Construc-Source Group tion Quality FSAR of Loca- Classi- Safety Codes and Seismic Assurance Section ~Su 1 tion fication Class Standards Categonr Reenirement Cmmmnte Principal Components (34*) (1)* (2)* (3)* (4)* (5)* (6)* (7)*

• Under Reactor Vessel Service E ui nt 9.1.4 Equipment handling platform GE Other X CRD handling equipment GE Other X Fuel Pool Cooli 5 Cleanu S stem 9.1.3 Heat exchangers Other III-3.

TB1A C Pumps Other I I I-3, Skimaer surge tanks Other III-3 Filter demineralizer vessels Other VI I I-1 19.31 Resin and precoat tanks Other API-650 Cooling loop piping and valves Other III-3 46,55 downstream of valve 1-53-001. 2-53-001 RHR intertie piping and valves Other I I 1-3 Emergency service water makeup piping and valves Other III-3 I Other piping and valves Other B31.1.0 NA 19.31.56 Coolin loo ipin u stream of valve l-53-Ii0(. 2-5-II01 from skirmer surge tank Other III-3 I Radioactive Waste Mana ement Li uid Waste Nana ement S stems 11.2 Centrifugal pumps R/RW/T 0 Other III-3 31.22 Atmospheric Tanks RW/T 0 Other VI I I-1/ 31.22 III-3 Rev. 47, 06/94

• Refer to the General Notes at the end of this table.

SSES-FSAR TABLE 3.2-1 SSES DESIGN CRITERIA

SUMMARY

(Continued) Page 52

54) The diesel generator jacket water coolers (OE507B and OE507D) utilize an ASME Section VIII replacement tube bundle in accordance with the guidance of NRC Generic Letter 89-09.

55) The following manually operated valves provide a fillable volume for use of the RHRFPC mode.

The following manually operated valves, which are in the seismically analyzed sections of pipe, require a capability to be closed following a seismic event. These valves have been analyzed to demonstrate that they will be capable of closure following a seismic event:

Spent Fuel Pool to 153018A/B (253018A/B), Fuel Pool Gate Drain to 153038 (253038), and Reactor Well Diffuser to 153030A/B (25303OA/B).

The following manually operated valves, which are in seismically analyzed sections of pipe, have a post seismic event function to remain in the closed position:

Reactor Well Drain to 153031 (253031), Reactor Well Drain to 153032 (253032), Reactor Well Drain to 153062 (253062), Dryer Separator Pool Drain to 153040 (253040), Dryer Separator Pool Drain to 153041 (253041), Cask Pit Gate Drain to 153050 (253050), Cask Pit Drain to 153054 (253054), Cask Pit Drain to 053084 &, 253800, and Cask Pit Diffuser to 053025.

56) The portions of piping between the surge tank up to and including valves HV15308 (25308), 153076 (253076), and 153064A/B (253064A/B) 'have been analyzed to show that they will remain intact following a seismic event. These valves have been analyzed to demonstrate that they will be capable of closure (or remaining closed) following a seismic event.

Closure of these valves is necessary to provide a fillable volume for use of the RHRFPC mode. The Skimmer Surge Tank drain line valves, 153065A (253065A), are normally closed and assumed to remain closed during a seismic event.

Rev. 47, 06/94

SSES-FSAR the capacity of a single RHR heat exchanger and related service water capability. Figure 5.4-12 shows the minimum time required to reduce vessel coolant temperature to 212'F using one RHR heat exchanger and allowing 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for flushing.

5.4.7.1.1.2 Low Pressure Coolant In'ection LPCI Mode The, functional design bases for the,LPCI mode is to pump a total of 21,300 gpm of water per loop using the separate pump loops from the suppression pool into the core region of the vessel, when the vessel pressure is 20 psid over drywell pressure. Injection flow commences at 280 psid vessel pressure above drywell pressure.

The initiating signals are: vessel level 1.0 feet above the active core or drywell pressure greater than or equal to 1.69 psig coincident with a low reactor pressure. The pumps will attain rated speed in 27 seconds and injection valves fully open in 40 seconds'.4.7.1.1.3 Su ression Pool Coolin Mode The functional design basis for the suppression pool cooling mode is that it shall have the capacity to ensure that the bulk suppression pool temperature immediately after a blowdown shall not exceed 207'F.

5.4.7.1.1.4 Containment S ra Coolin Mode The functional design basis for the containment spray cooling mode is that there should be two redundant means to spray into the drywell and suppression pool vapor space to reduce internal pressure to below design limits.

5.4.7.1.1.5 Reactor Steam Condensin Mode This section has been intentionally deleted.

5.4.7.1.1.6 Fuel Pool Coolin Mode The functional design basis for the fuel pool cooling mode is as follows:

a) The RHRFPC mode is designed and operated to provide cooling such that the fuel pool will be maintained at or below 125 F when the Emergency Heat Load (EHL) is Rev. 46, 06/93 5.4-33

SSES-FSAR resident in an isolated fuel pool. The EHL can be removed with a RHRSW inlet temperature of 91'F with only one RHR pump and heat exchange. For crosstied fuel pools, one RHR pump and heat exchanger in one unit in combination with the normal Fuel Pool Cooling system from the adjacent unit is sufficient to maintain the fuel pools at or below 125'F with the EHL resident in one fuel pool and fuel at the scheduled offload rate in the other fuel pool. This function is described in Sections 9.1.3.b and 9.1.3.2.

b) The RHRFPC mode is designed and operated to provide sufficient cooling to prevent fuel pool boiling in the event that a seismic event causes an extended loss of both units'ormal fuel pool cooling systems. This capability exists for both crosstied and isolated fuel pools.

When one RHR pump is operated in the RHRFPC mode, the spent fuel pool level must be raised to a minimum level above the weirs in order to support the design flowrate for this mode.

Additional details describing this mode of RHR are contained in Sections 5.4.7.2.6c, 9.1.3.1c, 9.1.3.2, and 9.1.3.3.

5.4.7.1.2 Design Basis for Isolation of RHR System from Reactor Coolant S stem The low pressure portions, of the RHR system, are isolated from full reactor pressure whenever the primary system pressure is above the RHR system design pressure. See Subsection 5.4.7.1.3 for further details. In addition, automatic isolation may occur for reasons of vessel water inventory retention which is unrelated to line pressure rates. (See Subsection 5.2.5 for an explanation of the Leak Detection System and the isolation signals.) Reactor Coolant pressure boundary valves are subject to inservice inspection leakage testing requirements as provided in 10CFR50.55a (see Subsection 3.9.6).

The RHR pumps are protected against damage from a closed discharge valve by means of automatic minimum flow valves, which open on low main line flow and close on high main line flow.

5.4.7.1.3 Desi n Basis For Pressure Relief Ca acit The relief valves in the RHR system are sized on one of three bases:

(1) Thermal relief only Rev. 46, 06/93 5.4-34

SSES-FSAR (2) Valve bypass leakage only (3) Control valve failure and the subsequent uncontrolled flow which results.

Transients are treated by items (1) and (3); item (2) above has resulted from an excessive leak past 'isolation valves. F055 ARB shall be sized to maintain upstream piping at 450 psig and 10 percent accumulation with F051 and F052 fully open and a reactor pressure equal to the lowest Nuclear Boiler safety/relief valve spring set point. F097 shall be sized to maintain upstream pressure at 180 psig and 10 percent accumulation with both PCV F053 A&B failed open. F030 A, B, C, and D, F025 A and B, F029, F126, and F087 shall be set at the design pressure specified in the process data drawing plus 10 percent accumulation.

Redundant interlocks prevent opening valves to the low pressure suction piping when the reactor pressure is above the shutdown range. These same interlocks initiate valve closure on increasing reactor pressure.

In addition a high pressure check valve will close to prevent reverse flow from the reactor if the pressure should increase.

Relief valves in the discharge piping are sized to account for leakage past the check valve.

5.4.7.1.4 Design Basis With Respect to General Desi n Criteria 5 The RHR system for each unit does not share equipment or structures with the other nuclear unit except for the Spent Fuel Pools as discussed in Subsection 9.1.3.3. They also share the common Emergency Service Water System. Sharing of this system with respect to General Design Criteria 5 is discussed in Section 3.1.2.1.5.

Rev. 46, 06/93 5.4- (35)

SSES-FSAR perform flushing will cause injection of non-reactor grade water into the reactor pressure vessel but will not affect performance of the RHR shutdown cooling system. At the end of this nominal flush, the testable check bypass valve may be opened in the shutdown return line and vessel water is-permitted to enter the upper portion'.,of the chosen loop to prewarm i'ffluent is directed to radwaste and a temperature element is used to control effluent temperature. The testable check bypass valve is closed and vessel suction valves are opened to allow prewarming of the lower half of the shutdown loop with effluent directed to radwaste as before. The radwaste effluent valves are closed, the heat exchanger bypass valves opened (the exchanger valves were closed after the initial cold water flush), then the pump starts at a regulated flow through return valve F017. After waiting several minutes to permit loop internal stability to be established the service water pump is started, the service water valves are opened, the heat exchanger inlet and outlet valves are opened and cooldown of the vessel is in progress. Cooldown rate is subsequently controlled via valves F017 (total flow) and F048 (heat exchanger bypass flow). All operations are performed from the control room except for opening and closing of local flush water valves.

The manual actions required for the most limiting failure are discussed in Subsection 5.4.7.1.5.

b. Steam Condensin This section has been intentionally deleted.

C. Fuel Pool Coolin Mode Operation of RHR in the fuel pool cooling mode requires manual actions to be performed both in the control room and locally. The system will also be required to be filled and vented, which will require the manipulation of various small manual valves. The filling operation may also include operation of the ESW system in the event the normal fill systems are unavailable.

actions are described in and controlled by plant These procedures.

5.4.7.3 Performance Evaluation Thermal performance of the RHR heat exchangers is based on the residual heat generated at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after rod insertion, a 125'F vessel outlet (exchanger inlet) temperature, and the flow of two loops in operation. Because shutdown is usually a controlled operation, maximum service water temperature less Rev. 46, 06/93 5.4-39

SSES-FSAR 10'F is used as the service water inlet temperature. These are nominal design conditions; if the service water temperature is higher, the exchanger capabilities are reduced and the shutdown time may be longer and vice versa.

5.4.7.3.1 Shutdown With All Com onents Available No typical curve is included here to show vessel cooldown temperatures versus time due to the infinite variety of such curves that may be due to: (1) clean steam systems that may allow the main condenser to be used as the heat sink when nuclear Rev. 46, 06/93 5.4- (40)

SSES-FSAR f) The plates will be washed in a mild abrasive and detergent solution, then rinsed in clean water and/or acetone. The plates will be dried in a 175'F oven for a4 hours, followed by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in a 300'F oven and 4 additional hours in a 500'F oven. The plate weight will be determined, at room temperature, following each drying i.'nterval. Drying may be discontinued when no further weight loss occurs.

g) Each plate will be weighed and determine weight change.

h) Reperform step ge.

i) All data will be recorded, including pH values, for future comparison.

9.1.3 SPENT FUEL POOL COOLINQ AND CLEANUP SYSTEM 9.1.3.1 Desi n Bases The Fuel Pool Cooling and Cleanup System (FPCCS) is designed and operated with the following considerations:

a) Maintaining the fuel pool water temperature below 125'F. The heat load which served as the basis for the FPCCS design is based upon filling the pool with 2840 fuel assemblies from normal refueling discharges and transferred to the fuel pool within 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> after shutdown. Tables 9.1-2a and 9.1-2b show the originally assumed discharge schedule and heat load. Table 9.1-2e shows an updated discharge schedule.

b) During an emergency heat load (EHL) condition, one RHR pump and heat exchanger are available for fuel pool cooling. The EHL condition occurs when the spent fuel racks of one spent fuel pool contain 2850 fuel assemblies including a full core discharged to the pool within 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> after shutdown (control rods inserted). Tables 9.1-2c and 9.1-2d show the discharge schedule and heat load that was assumed for the system's design for this condition for Units 1 and 2.

Table 9.1-2f shows an updated discharge schedule. The RHR Fuel Pool Cooling (RHRFPC) Mode will maintain the isolated fuel pool water temperature, (with the heat load of 3.39 x 10'TU/hr) at or below 125'F with or without assistance from the FPCCS under normal refueling conditions. When the decay heat load of the spent fuel drops to the level for which the FPCCS is designed, the RHR system may be disengaged. For crosstied spent fuel pools, the RHRFPC mode in one unit in combination with the normal Fuel Pool Cooling System of the other unit will maintain the crosstied fuel Rev. 48, 12/94 9.1-21

SSES-FSAR pools at or below 125'F with the EHL in one pool and fuel at the normal scheduled offload rate in the other pool.

c) Following a seismic event, the normal Fuel Pool Cooling system is postulated to be unavailable due to its Non-Seismic Category I, Non-Class 1E power design. If such an event were to occur the RHR Fuel Pool Cooling (RHRFPC) mode would be used to provide cooling to the spent fuel pools to prevent boiling.

All piping and components of the RHRFPC mode are Seismic Category 1, Quality Group B or C constructed to ASME Section III standards. The RHR system is Class 1E powered and both loops have separate power supplies.

The RHRFPC system is hardpiped and requires operation of several manual valves .(which are accessible following a seismic event) to establish the flowpath.

In addition, other manual and motor operated valves must be operated in order to assure proper operation of the RHRFPC mode. Proper operation of all active components in the RHRFPC mode is confirmed on a periodic basis in accordance with plant procedures.

The RHR pump suction path for the Fuel Pool Cooling mode is shared with the Shutdown Cooling mode of RHR.

Consequently, Shutdown Cooling and Fuel Pool Cooling cannot be performed concurrently on a given unit.

However, Alternate Shutdown Cooling and Fuel Pool Cooling can be performed concurrently since different suction sources are used.

Appendix 9A contains an evaluation of a boiling spent fuel pool for a Non-Seismic Category I Fuel Pool Cooling system. Boiling of the spent fuel pool(s) would not occur during a seismic event due to use of the RHR Fuel Pool Cooling system as a backup Seismic Category I Fuel Pool Cooling system. The RHRFPC mode can be placed into service well in advance of the postulated time to boil of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (see Subsection 9.1.3.3).

d) To maintain the water clarity and quality in the pools as follows to facilitate underwater handling of fuel assemblies and to minimize fission and corrosion product buildup that pose a radiological hazard to operating personnel:

Conductivity 3 mircromho/cm at 25'C pH 5.3 7.5 at 25'C Chloride (as Cl) 0.5 ppm Rev. 48, 12/94 9.1-22

SSES-FSAR Heavy elements (Fe,Cu,Hg,Ni) 0.1 ppm Total insolubles 1 ppm 9.1.3.2 S stem Descri tion Each reactor unit is provided with its own FPCCS as shown on Figures 9.1-7 and 9.1-8.

The system cools the fuel storage pool water by transferring the decay heat of the irradiated fuel through heat exchangers to the service water system.

Water clarity and quality in the fuel storage pools, transfer canals, reactor wells, dryer-separator pools, and shipping cask pit are maintained by filtering and demineralizing.

The FPCCS consists of fuel pool cooling'umps, heat exchangers, skimmer surge tanks, filter demineralizers, associated piping, valves, and instrumentation.

E ui ment Descri tion Table 9.1-1 shows the design parameters of the FPCCS equipment.

seismic and quality group classifications of the FPCCS 'he components are listed in Section 3.2.

One skimmer surge tank for each unit collects overflow water from skimmer drain openings with adjustable weirs at the water surface elevation of each pool and well. The common shipping cask pit water overflows to both units'kimmer surge tanks.

Wave suppression scuppers along the working side of the fuel pools also drain to the skimmer surge tanks. The skimmer openings in the pool liners are protected with a wire mesh screen to prevent floating objects such as the surface breaker viewing aids from entering the surge tanks. The adjustable weir plates are set according to the required cooling flow, desired flow pattern, and water shielding needs.

The skimmer surge tank provides a suction head for the fuel pool cooling pumps and a buffer volume during transient flows in the normally closed loop FPCCS. It provides sufficient live capacity for three days'ormal evaporative loss from the fuel pool without makeup from the condensate transfer system. A removable object retention screen in the tank is accessible through the flanged tank top. Tank level indication and alarms on a control panel on the refueling floor and/or the vicinity of the fuel pool cooling pumps announce when the remote manual makeup valves must be opened or water drained from the system.

Rev. 46, 06/93 9.1-23

)

SSES-FSAR The fuel pool cooling pumps are stopped upon a low tank level signals Three fuel pool heat exchangers piped in parallel are located in the reactor building below the surge tank bottom elevation.

The shell side is subjected to the static head of the skimmer surge tank level only. This is a minimum of 5 psi lower than the tube side service water pressure, thus minimizing the possibility of radioactive contamination of the service water system (see Subsection 9.2.1) from a tube leak.

The number of heat exchangers in service depends on the decay heat load from irradiated fuel in the spent fuel pool. The common inlet and each heat exchanger outlet temperature are recorded and high temperature alarmed on a local control panel.

Three fuel pool cooling pumps piped in parallel are placed in service in conjunction with the heat exchangers. They take suction from the heat exchangers and develop sufficient head to process a partial system flow through the filter demineralizers and transfer it combined with the bypass flow to the diffuser pipes at the bottom of the pools.

The pump controls, discharge pressure indicators, flow indicator, and alarms for low flow and low discharge pressure are provided on a local control panel.

The pumps trip individually upon low NPSH. Three fuel pool filter demineralizers are piped in parallel. One fuel pool filter demineralizer is normally associated with each FPCCS with the third one in standby. The design flow per filter demineralizer is less than the total system flow. Part of the cooled water is therefore bypassing at a manually adjustable rate.

Rev. 46, 06/93 9.X-(24)

SSES-FSAR skimmer surge tanks'uring periods when the heat in the pool is greater than the capacity of the fuel pool cooling system (such that acceptable fuel pool temperatures cannot be maintained), eg, storing of a full core of irradiated fuel shortly after shutdown, the RHR system can be used to dissipate the decay heat. One RHR pump takes suction from an intertie line to the skimmer surge tank and discharges through one RHR heat exchanger to two independent diffusers at the fuel pool bottom. With the spent fuel pool(s) filled to a height approximately 7.5 inches above the weirs, the skimmer surge tank provides sufficient suction head to an RHR pump in the RHR Fuel Pool Cooling (RHRFPC) mode.

Makeup water to replenish evaporative and small leakage losses from the pools is provided from the condensate transfer storage tank into the skimmer surge tank by opening a remote manual valve.

A Seismic Category I line from each of the two emergency service water loops is connected to the RHR intertie diffuser lines of each fuel pool, allowing for emergency makeup in support of RHRFPC or during postulated boiling of the pool water as described in Appendix 9A. The manual supply valves in these emergency makeup lines are accessible from elevations below the refueling floor.

9.1.3.3 Safet Evaluation At FPCCS design conditions where the pool heat load is 12.6 MBTU/HR and service water temperature is 95'F the FPCCS will maintain the fuel pool water less than 125'F. At improved service water temperature conditions, the FPCCS can maintain the fuel pool water less than 125'F with larger heat loads in the pool. This condition occurs during refueling outages.

When this condition exists the pool is monitored to assure adequate FPCCS capacity exists. When the FPCCS cannot maintain the pool temperature less than 125'F, the RHR system in the Fuel Pool Cooling Mode (RHRFPC) can be connected to the pools to maintain pool temperatures below 125'F by the RHRFPC mode.

AT EHL conditions (33.9 MBTU/HR), RHRFPC can maintain the pool temperature below 125'F with spray pond water temperatures below Technical Specification limits. Pool configuration will be maintained during the outage sequence so that the calculated time to boil is greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

A Seismic Category I makeup is provided by a 2 in. line from each emergency service water (ESW) loop to the RHR fuel pool diffusers, thus providing redundant flow paths from a reliable Rev. 46, 06/93 9.1-27

SSES-FSAR source of water. The design makeup rate from each ESW loop is based on replenishing the postulated boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity. This provides a capacity far in excess of what would be required by the RHRFPC mode in response to a loss of normal fuel pool cooling due to a seismic event.

All piping and equipment shared with or connecting to the RHR intertie loop are Seismic Category I, Quality Group C, or equivalent, and can be isolated from any piping associated with the non-Seismic Category I Quality Group C fuel pool cooling system.

Due to its Non-Seismic Category I, Non-Class lE power design, the consequences of a seismic event are required to be analyzed for the FPC system. In response to this event, the RHRFPC mode will be used to prevent boiling from occurring; however, a non-mechanistic evaluation of boiling of both spent fuel pools is contained in Appendix 9A in order to conservatively bound the radiological consequences.

The spent fuel pools are normally maintained in a crosstied configuration during dual unit operation and refueling outages.

This assures that the time to boil following a loss of normal fuel pool cooling is a minimum of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; however, in this configuration the time to boil is typically much greater than the minimum 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> time to boil minimum would only be approached shortly after a unit is shut down for refueling. After completion of a refueling outage, when both units are at power, the time to boil is typically on the order of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. The crosstied configuration allows use of either unit's systems (normal SFP Cooling or RHRFPC) to cool the pools, thus providing fuel pool cooling redundancy. Crosstied spent fuel pools also provide redundancy for the level instrumentation in the control room. This instrumentation is designed to operate following an Operating Basis Earthquake and under boiling spent fuel pool conditions and is expected to remain functional. While not classified as Class 1E equipment, the instruments receive power from independent Class 1E power supplies that are Diesel Generator backed.

Should a seismic event occur during dual unit power operation with crosstied pools, adequate reactor core cooling will be provided and spent fuel pool boiling will be prevented. Only one loop of RHR is necessary to provide long term decay heat removal per reactor vessel. Similarly, only one loop of RHR is necessary to provide long term decay heat removal to crosstied spent fuel pools. Since either unit's RHR system can provide cooling to both units spent fuel pools with the pools crosstied, a failure of one loop of RHR in one of the units would still allow a sufficient number of loops to cool both Rev. 46, 06/93 9.1-28

I J SSES-FSAR reactors and the spent fuel pools. In this case, the unit providing spent fuel pool cooling would utilize Alternate Shutdown Cooling for long-term decay heat removal from the reactor. The other unit would utilize the normal Shutdown Cooling mode.

During specific plant evolutions, such as transfer of fuel into fuel casks, the pools will not be crosstied. These evolutions will be procedurally controlled to ensure that sufficient cooling systems are available given the plant configuration at the time of the evolution.

An evaluation of the impacts of operating the RHRFPC mode on the Ultimate Heat Sink (UHS) was performed as a separate evaluation of the minimum heat transfer case discussed in Subsections 9.2.7.3.1 and 9.2.7.3.6. The results of this evaluation indicate that the spray pond (UHS) will be maintained below the design maximum temperature under worst case accident conditions.

Additional details on the design of the RHRFPC mode are provided in Sections 5.4.7.1.1.6, 5.4.7.2.6C, and 9.1.3.1 C.

Provisions to minimize and monitor leakage from the fuel pool are described in Subsection 9.1.2.3.

Makeup for evaporative and small leakage losses from the fuel pool is normally supplied from the condensate transfer system to the skimmer surge tanks of each unit. The intermittent flow rate is approximately 50 gpm to each surge tank.

The water level in the spent fuel storage pool is maintained at a height which is sufficient to provide shielding for required building occupancy. Radioactive particulates removed from the fuel pool are collected in filter demineralizer units in shielded cells. For these reasons,,the exposure of station personnel to radiation from the spent fuel pool cooling and cleanup system is normally minimal. Further details of radiological considerations are described in Chapter 12.

An evaluation of the radiological effect of a boiling fuel pool is presented in Appendix 9A.

9.1.3.4 Ins ection and Testin Re uirements No special tests are required because at least one pump, heat exchanger, and filter demineralizer are continuously in operation while fuel is stored in the pool. The remaining components are periodically operated to handle increased heat loads during refueling.

Rev. 46, 06/93 9.1-(29)

SSES-FSAR The pool liner leak detection drain valves are periodically opened and the leak rate estimated by the volumetric method.

Gas or dye pressure testing from behind the liner plate may be performed to locate a liner plate leak.

Routine visual inspection of the system components, instrumentation, and trouble alarms is provided to verify system operability. Components and piping of the FPCCS designed per ASME Boiler and Pressure Vessel Code,Section III, Class 3 are in-service inspected as described in Section 6.6.

The system will be preoperationally tested in accordance with the requirements of Chapter 14.

Rev. 46, 06/93 9.j.- (30)

SSES-FSAR switchgear and load center room coolers, which are normally supplied by the control structure chilled water system in Unit 1 or the direct expansion (DX) cooling system in Unit 2) required during normal and emergency conditions necessary to safely shut down the plant.

The ESWS is designed to take water from the spray pond (the ultimate heat sink), pump it to the various heat exchangers and return it to the spray pond by way of a network of sprays that dissipate the heat to the atmosphere, The ESWS is required to supply cooling water to:

a) The RHR pump room unit cooler and the motor bearing oil cooler of each RHR pump during all modes of operation of the RHR system.

b) All the heat exchangers associated with the four diesel generators aligned to the system during operation and test modes, except for the governor oil coolers.

c) The room coolers for the core spray (CS) pumps, the high pressure coolant injection (HPCI) pumps, and reactor core isolation cooling (RCIC) pumps during the operation of these systems.

d) The control structure chiller, the Unit 2 emergency switchgear cooling condensing unit, reactor building closed cooling water (RBCCW) heat exchangers, and the turbine building closed cooling water heat exchanger (TBCCW) during emergency operation.

e) The spent fuel pools to provide make-up for evaporative losses during operation of the normal fuel pool cooling system or RHR Fuel Pool Cooling (RHRFPC) mode, as well as, filling the spent fuel pools in support of RHRFPC.

The ESWS is also capable of supplying make-up for postulated boiling conditions as described in Appendix 9A for a Seismic Event.

The ESWS starts automatically within approx.40-100 seconds after the diesel generators receive their start initiation signal. The ESWS can also be started manually from either the main control room or from one of the two remote shutdown panels. (i.e , ESW loop A can only be started from the Unit 2

~

remote shutdown panel and ESW loop B can only be started from the Unit 1 remote shutdown panel.)

Rev. 47, 06/94 9.2-13

~ o

~ ~ ~

~ ~

~

SSES-FSAR In order to avoid having unacceptable voltages due to the RHR or CS pumps starting simultaneously with the ESW pumps, the ESW load sequence timer is reinitialized, but only have not started before the RHR or CS pumps.

if the ESW pumps The ESWS is designed to operate during any of the following conditions:

a) Loss of offsite power Rev. 47, 06/94 9.2- (14)

a ~

~ t ~~

I

~

SSES-FSAR Page 1 of 2 TABLE 9.2-3 DEFINITION OF ESW FLOWS FOR UNITS 1 84 2 Typical Min.

ESW Loop F)ow Typical Min. '"

ESW Safe Shutdown Non.Accident Flow - 2 Loops Min. Req'd. w/1 Loop Operating and Both Min. Req'd ESW ESW Loop Flow Operating and Units Service Water No. of Users Flow Per User For DBA and 1 Service Water Not Available Component Per Loop (GPM) Loop Failed Available A(B) B(A)

U1 U2

1) Standby'" Diesel Generator 4 common 1210 (A,B,C,D) 4840 4840 4840 total Heat Exchangers 1254(E)'" (4884)'" (4SS4)'" (4884)'"

2) RHR Pump Room Unit 400 Coolers

3) RHR Pump Motor Bearing 24 24 24 Oil Cooler

4) Core Spray Pump Room 144 144 144 144 Unit Coolers

5) HPCI Pump Room Unit 10 20 20 20 20 Coolers

6) RCIC Pump Room Unit 10 20 20 20 20 Coolers

7) Control Structure Chiller 1 common 740 740 740 per loop

8) Emergency Switchgear 72 72 72

"'45 Cooling Condensing Unit

9) RBCCW heat exchangeru'400 2800

10) TBCCW heat exchanger 490

11) Makeup to Fuel Pools'" 60 120 120 TOTAL 6380 5448 6380 3898 Loop Flow (GPM) (6424)'" (5492)'" (6424)'"

Rev. 47, 06/94

t

~

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> .a

SSES-FSAR Page 2 of 2 TABLE 9.2-3 (Continued)

DEFINITION OF ESW FLOWS FOR UNITS 1 BE 2

1) On one loop only.

2) Valve in parenthesis is with any three (3) of A, B, C 5 D units in service in conjunction with "E" unit.

3) The Diesel Generator "E" flow rate shown on this table is based on the continuous duty rating of the diesel generator (5000 kw) ~

4) Both loops of ESW are aligned to the D/G's. It is preferred that one pump per loop be run during normal operations. However, in the event of a DBA and a single failure in ESW, one loop will be available to supply the design flow to the Diesel Generator.

5) This column illustrates the ESW systems ability to supply DBA flows in addition to supplying TBCCW and RBCCW with both loops operating. The actual flow rates in each loop will vary slightly because of the crosstie at the diesels (i.e. the "B" loop will pass some flow to the D/G's).

I

6) The make-up rate shown here is conservatively based on a non-mechanistic boiling spent fuel pool (see Subsection 9.1.3.1). The flowrate for make-up of evaporative losses during RHRFPC operation would be significantly less.

Rev. 47, 06/94

c >

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SSES-FSAR which is in the control room, and each pump chamber is provided with a low level submergence switch which alarms in the control room.

9.2.5.6 Pi e Crack Leaka e Detection Leakage from the ESWS can be detected by one of several methods depending on location. Leakage from piping within the ESSW Pumphouse drains into a pit which is equipped with a level switch to alarm on high water. The yard. piping from the ESSW pumphouse to the pump discharge flow elements is contained in a guard pipe which drains back to the ESSW Pumphouse and into the same pit as described above. The remaining yard piping is located in a high traffic area and the presence of a significant leak will be visually apparent.

Leak detection within the Reactor Buildings, Control Structure and Diesel Generator Buildings differs depending on the location. Seismically analyzed room flood detectors are used in the lowest elevations, such as, the RHR, Core Spray, HPCI, RCIC and TBCCW Heat Exchanger rooms. Flood detection for the rooms containing ESW lines supplying the RBCCW heat exchangers, Control Structure Chillers, Unit 2 Dx units and Fuel Pool Makeup is not feasible nor desirable, since the lines are located in upper elevations of the Reactor Building and Control Structure. In these areas, floor drains route the leakage to radwaste via either the Reactor Building or Turbine Building sumps. The excessive influent into the radwaste system will alert operators to a pipe leak.

9.2.6 RHR SERVICE WATER SYSTEM 9.2.6.1 Desi n Bases The Residual Heat Removal Service Water System (RHRSWS) has a safety related function and is an engineered safeguard system designed to supply cooling water to the residual heat removal (RHR) heat exchangers of both units.

The RHRSWS is designed to take water from the spray pond (the ultimate heat sink), pump it through the RHR heat exchanger,and return it to the spray pond by way of a spray network that dissipates the heat to the atmosphere.

The RHRSWS is designed to provide a reliable source of cooling water for all operating modes of the RHR system including heat removal under post-accident conditions, RHR Fuel Pool Cooling (RHRFPC) following a seismic event, and also to provide water to flood the reactor core or the primary containment after an accident, should it be necessary.

Rev. 47, 06/94 9.2-19

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SSES-FSAR 9.2.7 ULTIMATE HEAT SINK The ultimate heat sink has safety related functions and provides cooling water for use in the Engineered Safeguard Service Water system, described in Subsections 9.2.5 and 9.2.6, during ESSW testing, normal shutdown, and accident conditions.

9.2.7.1 Desi n Bases The ultimate heat sink is capable of providing sufficient cooling water without makeup to the spray pond for at least 30 days to (a) permit simultaneous safe shutdown and cooldown of both nuclear reactor units and maintain them in a safe shutdown condition, (b) mitigate the effects of an accident in one unit, permit safe control and cooldown of the other unit, and maintain it in a safe shutdown condition or (c) permit simultaneous safe shutdown and cooldown of both units and maintain them in safe shutdown while providing adequate cooling to both spent fuel pools following a seismic event. Continued cooling beyond 30 days is ensured by use of the makeup pumps to keep the pond at normal water level. The makeup pumps are designed to operate below the historic minimum water level of the Susquehanna River. In the event that makeup water from the makeup pumps is not available, additional provisions will be made in the 30 days available to assure continued cooling of the emergency equipment beyond 30 days. These provisions include but are not limited to: re-establishing makeup pump flow to the spray pond, emptying the cooling tower basins into the spray pond, trucking in water from neighboring water sources (such as the Susquehanna River), and providing temporary pumps and/or lines to pump water from neighboring water sources (such as the Susquehanna River, on site storage tanks, well water, etc.)'. This is in compliance with NRC Regulatory Guide 1.27 Rev. 2 as discussed in Section 3.13.

The ultimate heat sink is also capable of providing enough cooling water without makeup, for a design basis LOCA in one unit with the simultaneous shutdown of the other unit, for 30 days while assuming a concurrent SSE, single failure, and. loss of offsite power. This event is evaluated in Subsection 9.2.7.3.1.

The ultimate heat sink consists of at least one highly reliable water source with a capability to perform the safety function required above during and after any one of the following postulated design basis events:

a) The most severe natural phenomena, including the safe shutdown earthquake, tornado, flood, or drought taken individually Rev. 47, 06/94 9.2-25

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SSES-FSAR Page 1 of 1 TABLE 9.2-8 SUSQUEHANNA POND WATER ALLOWANCES Loss Description Water Allowance (x10'al) a) Evaporation due to heat dissipation duty for maximum water loss case 7.95 b) Drift from wind for maximum water loss case 1.15 c) Percolation through the pond lining 0.3 d) System charging volume Negligible e) Maximum solar evaporation losses 1.85 f) Losses resulting from wave action ul0 g) Losses resulting from sedimentation'" 1.0 h) Fuel pool makeup'" 5.0 i) A contingency for water quality considerations 2.7 Total Pond Volume Required 19.95 Total Pond Volume Provided 25.0 Based on design provisions for protection from this loss.

(2)

Negligible sedimentation is anticipated. The value given corresponds to 6 in. of pond depth, which is a conservative allowance between cleaning periods.

(3)

For additional conservatism, this value assumes boiling of the fuel pools consistent with the non-mechanistic boiling pool analysis in Appendix 9A.

Rev. 35, 07/84

SSES-FSAR APPENDIX 9A ANALYSIS FOR NON SEISMIC SPENT FUEL POOL COOLING SYSTEMS As described in Subsection 9.1.3 the Spent Fuel Pool (SFP)

Cooling Systems are designed as non-sei'smic Category I, Quality Group C systems. Consequently, the radiological consequences of a loss of spent fuel pool cooling due to a seismic event are evaluated. In order to perform this analysis it to assume the SFP will boil even though Section 9.1.3.3 is necessary establishes that the design basis of the plant for this event is to prevent boiling through the use of the RHRFPC mode.

Since the cooling systems and in close proximity it for both units are cross-connected was assumed that a seismic event causes the loss of cooling to both spent fuel pools. In addition, in order to maximize both'he heat loads and the iodine inventories in the pools, refuelings within 135 days were postulated. (Period of time between outages is nominally 180 days, thus use of 135 days is conservative.) The loss of cooling was assumed during the second refueling, just after isolation of the pools (i.e., refueling and cask pit gates installed). The RHR system is assumed to not be available for cooling the SFP even though it in response to this event. Thus, would be able to provide cooling it is assumed that the pools will boil. The analysis involved an evaluation of the time to pool boiling, the ability to maintain water level if boils, and the thyroid dose consequences at the LPZ boundary the pool due to iodine releases from the boiling pools.

The assumptions used in this analysis were consistently chosen to be conservative and bounding similar to those in Regulatory Guides for design basis accidents (e.g., Regulatory Guides 1.3, 1.25, etc.). The combination of all of these design basis assumptions occurring at the same time would be extremely unlikely, making this accident as analyzed, one of very low probability. Many of the assumptions are considered to be overly conservative. For example, operating experience with present BWR fuels (Reference 9A-1) indicates that the assumption of 700 pCi/sec (full power design basis leakage rate) is conservative for determining reactor coolant concentrations during operating conditions. This same leakage rate will be assumed for the fuel in the SFP, which is even more conservative. Even though spiking factors have yet to be observed for a temperature rise in SFPs, spiking factors have been utilized. A more realistic evaluation of this accident would result in releases of radioactivity, of magnitude below the calculated values.

if any, many orders The realistic releases would be well below the 10CFR50 Appendix I related Technical Specifications, indicating that such an incident is of little or no consequence.

Rev. 46, 06/93 9A-1

SSES-FSAR The pools will be operated in a manner which will ensure that they will not boil until at least 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after the loss of cooling. Since cooling is assumed not to be restored before the pool boils, makeup water from the Category I Emergency Service Water System is assumed to be added to the pool at a rate equal to the boiloff to keep the fuel covered with 23 feet of water at all times.

As shown in Table 9A-1, the thyroid dose consequences of the boiling pool, without operation of the Standby Gas Treatment System, are well below the guideline values of 10CFR100 and the 1.5 REM thyroid guideline of Regulatory Guide 1.29.

The following assumptions were used to calculate the heat generation and boiling rate.

l. Each fuel pool is full with 2850 fuel assemblies. The maximum expected discharge batch size of 280 assemblies was used for the most recent offload in each pool. The earlier offloads were based on 256 assembly batch sizes. To determine the heat load and thus boiling evaporation rate, sequential refuelings 129 days apart are assumed. The event is assumed to occur 6 days after the second unit is, shutdown. Six days is conservatively chosen as the minimum time to unload 280 assemblies and reinstall the fuel pool gates (thus isolating the pool). Therefore, one unit's fuel pool inventory is assumed to have decayed for 6 days.

Actual sequential refuelings occur approximately 180 days apart. The normal time to defuel 280 assemblies is 8 days.

These assumptions maximize the heat load in the recently defueled pool and thus the boiling evaporation rate. The analyses were performed for power uprate conditions.

2. The decay heat was calculated using the ANSI/ANS-5.1-1979 decay heat standard. This standard includes methodology for calculating the decay uncertainty. All values of the decay heat in this section are equal to the nominal value plus two standard deviations.

3. To determine a conservative boiling evaporation rate for purposes of this radiological evaluation, all heat generated by the fuel is assumed to be absorbed by the water in order to minimize the time to boiling. No heat is lost to the surroundings by conduction through the concrete and steel, or by evaporation. The temperature gradients from the fuel at the bottom of the pool to the cooler water at the top will create convective water and heat currents which will thoroughly mix the water, and promote an even distribution of heat rather than localized points of surface boiling.

Rev. 46, 06/93 9A-2

SSES-FSAR

4. The activity release rate from the pool depends on the evaporation rate and the iodine carryover fraction at the pool surface. The evaporation rate prior to boiling is bounded by the evaporation rate at initiation of boiling.

It is conservatively assumed that the evaporation rate prior to boiling is the same as that during boiling.

Rev. 46, 06/93

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