ML17157C292

Summary of 930318 Meeting W/Util Re Effects Resulting from loss-of-SFP Cooling Following Loca.List of Attendees & Licensee Handouts Encl

ML17157C292
Person / Time
Site:	Susquehanna
Issue date:	04/16/1993
From:	Clark R Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9304220335
Download: ML17157C292 (50)

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 April 16, 1993 Docket Nos.

50-387 and 50-388 LICENSEE:

Pennsylvania Power and Light Company FACILITY:

Susquehanna Steam Electric Station, Units 1 and 2

SUBJECT:

MEETING

SUMMARY

, SPENT FUEL POOL COOLING UNDER SEVERE ACCIDENT CONDITIONS On March 18,

1993, a meeting was held between Pennsylvania Power and Light Company (PP&L), the licensee for Susquehanna Steam Electric Station (SSES),

Units 1

and 2 and the NRC staff to discuss the effects resulting from a loss-of-spent-fuel-pool (SFP) cooling following a loss-of-coolant-accident (LOCA).

~ac~caround By letter dated November 27,

1992, two contract engineers workin'g in PP&L's Nuclear Plant Engineering Section filed a 10 Part CFR 21 report contending that a "substantial safety hazard" exists in the design of the SSES SFP cooling system in that it fails to meet numerous regulatory requirements and that following a postulated "design basis accident (DBA) LOCA or LOCA with a loss-of-offsite-power (LOOP)", "there is the potential for meltdown of irradiated fuel outside primary containment and the failure of all safety-related systems in the reactor building".

The 10 CFR Part 21 report was supplemented by a letter dated December 14, 1992.

Based on an initial review of the Part 21 report, the staff requested additional information from the licensee by letter dated February 18, 1993.

The staff proposed that the questions be used as the agenda for a meeting with the licensee.

The proposed meeting was held March 18,

1993, and is the subject of this correspondence.

The meeting was attended by the two engineers who filed the Part 21 report, a

reporter for " Inside NRC" and a representative of the Commonwealth of Pennsylvania, A list of attendees is attached as Enclosure 1.

~Summa r The licensee's presentation closely followed the slides in Enclosure 2.

Licensin Histor In 1976, PP&L revised the Susquehanna Preliminary Safety Analysis Report (PSAR) to change the design of the SFP cooling system from a seismic to a non-seismic system.

Pool boiling was not included in the Susquehanna original licensing basis.

However, the cooling system was analyzed for pool boiling as a result of a seismic event.

The design basis cause of loss-of-fuel pool cooling is a seismic event.

Redundant seismic Category I Emergency Service Mater (ESM) connections are provided to allow for makeup of evaporative losses in the event of failure of the SFP cooling system, The plant's design pre)gin (i".,11'i; (jujij(ijI~j~'dQi'+1dI(t 1

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response to a complete loss-of-pool cooling due to a seismic event is to allow the pool to boil with inventory makeup provided from a safety-related

source, the ESW system.

The analysis for this event is documented in Appendix 9A of the Final Safety Analysis Report (FSAR).

S stem Overview The Susquehanna SFP is designed to store 2850 fuel assemblies.

The relevant design basis requirements and description are found in subsection

9. 1.3 of the FSAR.

The pool cooling system consists of three pumps and heat exchangers per unit, three filter-demineralizer units, one skimmer surge tank per unit, associated piping, valves and instrumentation.

The heat removal capacity of the system is 13.2 M BTu/hr with 95 'F service water.

The system is designed to maintain pool water temperature at or below 125 'F during the "emergency heat load" condition equivalent to a full core offload 10.5 days after shutdown following a typical fuel cycle discharge which fills the pool.

Under this postulated full core offload scenerio, the pool water temperature cannot be maintained below the design limit of 125 'F by the SFP cooling system alone.

As stated in the FSAR, under these conditions, it is necessary to connect one loop of the residual heat removal (RHR) system to the SFP.

Loss-of-Fuel Pool Coolin The plant's design response to a complete loss-of-fuel-pool cooling due to a

seismic event is to allow the fuel pool to boil, with inventory makeup provided from a safety-related

source, the ESW system.

The licensee has performed analyses to determine the time it would take for the pool to boil and the potential offsite doses, assuming the maximum normal heat load (MNHL) which they determined to be 12.6 M BTu/hr.

Operation of the Standby Gas Treatment System (SGTS) is anticipated under these conditions;

however, in the

analyses, no credit was taken for the SGTS.

The licensee's analyses determined that there would be a minimum of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> before the pool would boil with no makeup from the two refueling water pumps (which transfer condensate from the refueling water storage tank into the reactor well and dryer-separator area of the pool) or from the ESW.

Under these conditions, the licensee's analysis determined that the offsite dose is well below the 10 CFR Part 100 limit.

A considerable part of the meeting was devoted to the licensee's response to the staff's request of February 18, 1993, for additional information, The licensee's responses are summarized in the enclosed slides. 't the conclusion

" of the meeting, the staff requested that the licensee formally submit a

kt I written response, taking,into account the questions on source term, which determine the radiation levels operators would be exposed to following an accident.

Enclosures:

1.

List of Attendees 2.

Licensee's Handouts Original signed by Richard J. Clark Richard J. Clark, Senior Project Hanager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next page DISTRIBUTION tk k

'ktt RRC

& Loca PDRs*

PDI-2 Reading*

T. Hurley/F. Hiraglia, 12G-18 J.

Partlow, 12G-18 S.

Varga J.

Calvo C. Hiller R. Clark*

H. O'rien OGC E. Jordan L. Kokajko V. Ordaz J.

Lee C. HcCracken H. Fleishman ACRS(10)

V. HcCree,

EDO, 17G-21 J. White, RGN-I*

E. Wenzinger, RGN-I*

• Licensee's handouts 0

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written response, taking into account the questions on source term, which determine the radiation levels operators would be exposed to following an accident.

Enclosures:

1.

List of Attendees 2.

Licensee's Handouts R c ard J.

C ark, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next page

Pennsylvania Power

& Light Company a

Susquehanna Steam Electric Station, Units 1

& 2 CC:

Jay Silberg, Esq.

Shaw, Pittman, Potts

& Trowbridge 2300 N Street N.W.

Washington, D.C.

20037 Bryan A. Snapp, Esq.

Assistant Corporate Counsel Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. J.

M. Kenny Licensing Group Supervisor Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. Scott Barber Senior Resident Inspector U.

S. Nuclear Regulatory Commission P.O.

Box 35 Berwick, Pennsylvania 18603-0035 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.

Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III Allegheny Elec.

Cooperative, Inc.

212 Locust Street P.O.

Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I

U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1,

Box 1797 Berwick, Pennsylvania 18603 George T. Jones Manager-Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Robert G.

Byram Senior Vice President-Nuclear Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. David A. Lochbaum 80 Tuttle Road

Watchung, New Jersey 07060 Mr. Donald C. Prevatte 7924 Woodsbluff Run Fogelsville, Pennsylvania 18601

MEETING ATTENDEES MEETING BETWEEN NRC STAFF AND PP&L SPENT FUEL POOL COOLING SYSTEM MARCH 18 1993 ENCLOSURE 1

NAME C. Hiller R. Clark D.

Shum W. Dizard L, Kokajko V. Ordaz J.

Lee C. McCracken J.

White M, Fleishman 0,

Lochbaum D. Prevatte G. Jones J.

Kenny D.

Ney H. Crowthers D. Kostelnik R. Sgarro H. Mjaatvedt D. Roth G. Hiller ORGANIZATION NRC/NRR NRC/NRR NRC/NRR INSIDE NRC NRC/NRR NRC/NRR NRC/NRR, NRC/NRR NRC/RGN-I NRC/OCM/KR Self-Employed Powerdyne Corp.

PP&L PP&L PA/DER/BRP PP&L PP&L PP&L PP&L PP&L PP&L

ue oo oo in ssue FPC-NRC g

8 03-1 7-1 993

Fuel Pool Cooling issue

~ Introductory Remarks

~ Licensing History

~ System Overview

~ Design Bases

~ NRC Questions

~ Event Evaluations

~ Summary

~ Closing Remarks

Licensing History

~ PSAR Change submitted in1976

-FPC systems are non-seismic

-Pool boiling analyzed from seismic event

-Emergency Service Water provides make-up Resulting dose meets regulatory requirements

~ Documented in SSES FSAR

-NRC approval in Susquehanna SER

~ Pool boiling was not included in our original licensing basis

s em verview FPC-NRC 03-17-1 993

Spent Fuel Pool Design

~ Storage capacity for 2850 fuel assemblies

~ Seismic Category I structure

~ Subcriticality assured

~ Decay heat analyzed

~ Radiation levels ALARA

Fuel Pool Cooling System

~ Three pumps and beat excbangers per unit

~ 13.2 MBtulhr capacity @ 95 F Service Water

~ Filter demineralizers

~ Local control panels

~ RHR Fuel Pool Cooling Assist mode

~ ESW supply for emergency makeup

UNIT 2 UNIT l SPENT FUEL POOL SKIMMER TANK iJ CASK PIT:'.,"

8KIMMER TANK SPENT FUEL POOL 25307IA 253070A 2530718 25309IA 253090A 253500 251070 2530908 253501 2530708 2530918 FROM ESW FROM ESW 253001 253021 TO FUEL PQQL 153001 153021 153019A TO RHR TO CLEANUP TO SUP.

POOL QRY WELL SPRAY TO LPCI 251060 HV 251 F008 FUEL POOL HEAT EXCH.

IA IA RHR HEAT EXCH.

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REACTOR PRESSURE VESSEL SERVICE WATER IB IC IB IC FUEL POOL COOLING PUMPS

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Design Bases

~ Fuel Pool Structural Design

~ Normal and Emergency Heat Loads

~ Loss of Fuel Pool Cooling Event

~ Radiological Consequences

Fuel Pool Structural Design

~ Fuel Pool Cooling System is non-Seismic Cat I

~ All piping 8 equipment shared with or connecting to the RHR intertie loop is Seismic Cat l and can be isolated from non-seismic portion.

~ Pool and racks are Seismic Cat I

~ ESW makeup piping is Seismic Cat I

FSAR Design Basis Decay Heat Last Batch Offload to Single Fuel Pool Decay Heat (MBtu/hr),

13 Design Basis 10 10 20 30 40 Days After Shutdown 50 60

Maximum Normal Heat Load MNHL

~ Assumed fuel cycle discharge schedule

~ Fuel pool filled

~ Last quarter core offload

~ Must maintain temperature below

. 125'F

~ Heat load = 12.6 MBtu/hr Design Basis Decay Heat Heat Load (MBtu/hr) 14 12 10 8

10 12 14 Days After Shutdown

Emergency Heat Load EHL

~ Assumed fuel cycle discharge schedule

~ Fuel pool filled

~ Full core offload

~ Must maintain temperature below 125'F

~ Heat load = 32.6 MBtulhr

~ Credit for RHR system Fuel Pool Cooling Assist

~ Designed for heat loads > 12.6 MBtu/hr

Loss of Fuel Pool Cooling Event

~ Assumes functional failure of system caused by seismic event

-Analyzed to determine potential offsite doses

-No credit for SGTS

-Analyzed to determine time to boil

- ESW makeup flowrate based on MNHL For all other events, restoration of fuel pool cooling function was assumed

Time-to-Boil Analysis

~ Same fuel cycle as for MNHL

~ Fuel pool filled

~ Last quarter core offload

~ Heat load = 9.8 MBtulhr

~ Time-to boil: 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> Design Decay Heat Heat Load (MBtu/hr) 14 12 10 8

10 12 14 Days After Shutdown

Radiological Consequences FSAR

~

Purpose:

Calculate offsite dose resulting from a loss of FPC.due to a seismic event

~ Boiling is assumed to occur

~ Offsite dose below 10CFR100 limit and the 1.5 Rem thyroid guideline of RG 1.29 "Seismic Design Classification"

ues ions FPC-NRC 16 03-17-1 993

NRC uestions

~ Changes to fuel design and refueling practices

~ Event Evaluations (Outage vs. Non-Outage)

-Seismic (FPC Design Basis)

-LOOP LOCA LOCA/LOOP

-Severe Accident (Beyond Plant Design Basis)

Fuel Desi n and Refuelin Practices

~ Qx9 vs Bx8 fuel bundles

-higher burnup

-lower fuel pin failure rate

~ 18 month vs 12 month fuel cycles

-1/3 core vs 1/4 core reloads

~ Full core offload vs fuel shuffling

~ Shutdown risk assessed and managed

Comparison with Original Desi n

Last Batch Offload to Single Fuel Pool (U1-7)

Decay Heat (MBtu/hr) 14 12 10 Design Basis U1-7 10 40 20 30 Days AfferShutdown 50 60 FPC-NRC 03-1 7-1 993

Fuel Design and Refueling Practices

- Evaluation

~ All changes currently bounded by existing design basis

~ FSAR Change developed to support current refuel practices Doses bounded by previous SER evaluations

-25 hour time to boil maintained

Typical Refuel Outage

~ Shutdown risk assessed and managed

~ Outage periods impose highest heat loads

~ Configuration provides greatest capability to prevent boiling

~ Fuel Pools cross-tied via Cask Storage Pit

~ Rx Cavity 8 Storage Pit cross-tied

~ No fuel in vessel

Typical Refuel Outage Sequence

~ Outage schedule reviewed to assure FSAR Appendix 9A is bounding

~ Core fullyoffloaded 13 days after shutdown Fuel Pools monitored

~ Fuel Pool temperature maintained at less than110'F

~ Fuel Pools crosstied and connected to RX Cavity

ven va ua ions FPC-NRC 03-17-1993

Event Evaluations

~ Seismic Event (FPC Design Basis)

~ Loss of Offsite Power (LOOP)

~ LOCA LOCA/LOOP

~ Severe Accident (Beyond Plant Design Basis)

Seismic Event - Evaluation

~ Use RHR FPC Assist to prevent boiling

. Allowboiling with makeup from ES'LN

~ Restoration of HVAC or use of SGTS on Zone III only will mitigate effects of boiling

~ If boiling occurs, condensation in Zone III will drain to RB sumps

~ Can accommodate condensation for extended period in RB basement without compromising long-term cooling

REFUELING FLOOR III UNITNl

~ ~ SUCTION BLDG BECIRC DISCHARGE II UNITN2 REACTOR BLDG SGTS

Comparison To Actual Outa e

Practice

~ Analysis:

Time after shutdown 10.5 days

- Heat load was 9.8 MBtu/hr

-Predicted time-to-boil was 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />

~ Actual:

Time for U2-5RIO was 38 days Heat load was 5.65 MBtu/hr Predicted time-to-boil was 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />

~ Outages are managed to maintain design basis

Time-to-Boil T 0 =110'F Time To Boil (hours) 125 100 75 50 2 Pools 8 Well 2 Pools 1 Pool 25 0

0 5

10 15 20 25 30 35 40 Heat Load (MBtu/hr)

Volume (2 Pools 8 Well) = 172,232 cu-ft Volume (2 Pools) = 107,178 cu-ft Volume (1 Pool) = 48,690 cu-ft FPC-NRC 27 03-1 7-1 993

Typical Refuel Outage-Heat Decay Decay Heat (MBtu/hr) 30 25 20 10 10 15 20 25 30 35 40 Days After Shutdown 50 55 60 65-

Typical Refuel Outa e-Time To Boil Time To Boil (hours) 70 60 110'F 50 40 30 20 10 10 15 20 25 30 35 40 45 Days After Shutdown 50 55 60 65 FPC-NRC 29 03-17-1993

LOOP

~ FPC lost to both units

~ Expect to recover offsite power well within 25 hrs

~ Stable Grid

~ Maintain for first contingency loss of largest unit

~ Recover Grid in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

-Procedures in place and tested

~ SBO coping time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

LOOP Evaluation

~ LOOP recovery times and their associated probabilities (Ref: SSES IPE) 3 hrs - 89.20%

12 his - 97.96%

-24 hrs - 99.53%

-60 hrs - 99.92%

-75 hrs - 99.95%

~ Restore normal FPCS or RHR FPC Assist

~ Cooling restored prior to pool boiling

DBA LOCA-Fuel Response

~ Analyzed with GE SAFER/GESTR

~ PCTs - 1000'F for best-estimate case

~ PCTs - 1500 F for Appendix K evaluation

~ Results are below threshold for clad rupture

~ No fuel failures expected

~ Reactor Bldg is accessible

DBA LOCA

~ FPC lost initiallydue to Aux Load Shed

~ Expect normal FPCS to be functional following LOCA

~ Reactor Bldg accessible

~ Adequate time exists to perform operator actions

~ Cooling can be restored

~ Non-LOCA unit can provide cooling via cross-tied pools

DBA LOCA/LOOP

~ FPC lost initiallydue to Aux Load Shed

~ Expect normal FPCS to be functional following LOCA

~ Expect to restore offsite power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

~ Reactor Bldg accessible

~ Adequate time exists to perform operator actions

~ Cooling can be restored Non-LOCA unit can provide cooling via cross-tied pools

Severe Accident - Shieldin Evaluation

~ Reg Guide 1.3 source (100% core damage)

~ Assumes highest contact dose for room

~

Purpose:

To maximize dose for shielding 8 EQ design (Defense in Depth)

~ Contained source yields > 5000 Rlhr in some areas

~ No airborne contribution assumed

~ RHR FPC Assist not accessible

~ Two ESW valves on Unit 1 not accessible

Severe Accident Evaluation-Beyond Plant Design Basis

~ Severe core damage event renders FPC and RHR FPC Assist inaccessible

~ Only expect one pool to boil (ifpools isolated)

~ Makeup to isolated pool with ESLN always possible via Cask Pit connection

~ Boiloffresults in - 500K gallons per unit based on evaporation rates

~ Can accommodate condensation for extended period in RB basement without compromising long-term cooling

~ Long-term decay heat removal assured

Summa

~ Shutdown risk management

~ Reliable offsite power supply

~ Accident response methodology minimizes risk to fuel 0 - Restore normal Fuel Pool Cooling

-RHR Fuel Pool Cooling Assist Cross-connect fuel pools

- ESW makeup from either unit

Summary

~ Plant is within original plant design basis

~ Safe operation is assured

~ Fuel pool can be cooled for:

-Seismic events

-LOOP

-LOCA LOCA/LOOP

~ Can mitigate effects for Severe Accident

~ Controls strengthened

~ Procedures improved

,.;, Modifications are curreqtly under consideration

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