ML20059N956

Spent Fuel Pool Cooling Sys Enhancement. W/Two Oversize Drawings

ML20059N956
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	05/31/1990
From:	Hansen C, Hoffman J, Yasi O VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20059N954	List: BVY-90-093, Forwards Requested Design Info Needed for NRC Review of Planned Enhancements to Existing Standby Fuel Pool Cooling Subsystem ML20059N955 ML20059N956
References
EDCR-89-408, EDCR-89-408-R05, EDCR-89-408-R5, NUDOCS 9010240182
Download: ML20059N956 (60)

Text

rem! N 5 ENGINEERING E R NO.:

a DESIGN CHANGE REQUEST p

NUCLEAR SERVICES DIVISION w

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4920 YANKEE ATOMIC ELECTRIC CO.

EDCR TITLE: SPENT PUEL PCOL COOLING SYSTEM ENHANCEMENT wI e

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ENGINEERING DESIGN CHANGE REQJESV REV!EO AND AP FI I/ Outrice 6/1/90 89 LO8 Date Received Priority _

EOCR No.

Scent Fuel Pool Cooling System Enhancement Title __

Engineering Surmort Cognizant Design Department Mech. Cnst/ JF Calchern Cognizant Installation Department YMSD/CA Eancen ESDh'A Stello Plant Cognizant Engineer (PCE)

Plant Staff Review Routin_g RDL Maintenance DLP Engineering Support TAW Instrument / Control PPc

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oc RPG Quality Services Group Plant ISI/IST Coordinator Plant Services '

Plant Quality Assur mes Den RWS Training (info only)

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DISAPPROVED APPROVED PLANT MANAGER DISPOSITION.

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VYAPF 6004.01 AP 6004 Rev.12 Page i of 1 l i'

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4 FORM WE-100-2 GLOOSARY 0420, GL "E" REVISION 10 EDCR CHANGE NOTICE NUCLEAR SERVICES DIVISION EDCR NO.: 89-408 YANKEE ATOMIC ELECTRIC COMPANY CHANGE NO.:

1 PLANT: Vermont Yankee W.O.

'4920 Spent Puel Pool Cooling System Enhancement-EDCR TITLE:

X MINOR CHANGE MAJOR CHANGE - REVIEWS ATTACHED - SAFETY EVALUATION REVISED (If Applicable)

COGNIZANT ENGINEERS s'. H. Hansen YNSD:

PLANT:

M. A. Stello REASON FOR RESOLUTION OF ENGINEERING DEFICIENCY CHANGE:

X SUPPLEMENTARY INFORMATION DESIGN CHANGE FIELD CHANGE Incorporation of minor PORC approved changes and clarification.

Incorporation of supplementary information to finalize Service Water System effects from this CONTINUED ON SHEET design change.

INSTRUCTIONS FOR CRANGE/ DESCRIPTION (TYPE OR PRINT)

Remove entire text and replace with attached.

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Remove Enclosure "C" and replace with attached.

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EFFECT OF CHANGE ON INTENT, SCOPE, CAFETY EVALUATION, OR QA REQUIREMENTS:

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X NONE - MINOR CHANGE - CONCURRED WITH BY PLANT CE

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.___ DESCRIBED ABOVE PREPARED BY:

a DATE: - b,: / -9 0 A

DATE:

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Oct er.- Avv-a REVISION 1 EIX:R CHANGE NOTICE NUCI. EAR SERVICES DIVISION EDCR NO.89-408

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YANKEE ATOMIC ELECTRIC COMPANY CHANGE NO.:.

1 Continuation Sheet SHEET 2

OF' 2

INSTRUCTIONS FOR CHANGE / DESCRIPTION (Cont.)

Remove Enclosure "G" index and replace with attached.

Remove Reference (f) and replace with attached revised Reference (f).

Add attached References (qq), (rr) and (ss) to Enclosure "G".

EQi l changes are highlighted by side-bars.

Remove Enclosure "B", Attachment A, FSAR Section 10.5, " Fuel Pool Cooling System" and replace with attached.

Remove enclosure "F" drawing index and replace with attached.

I Remove Enclosure "E" Pages 1 thru 9, 40 thru 44 and replace with attached.

Remove Enclosure "D" and replace with attached.

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EDCR 89408 Spent Fuel Pool Cooling System Enhancement 1.0 Executive Summary This Engineering Design Change Request (EDCR) provides the design of the Standby Fuel Pool Cooling Subsystem (SFPCS) of the Fuel Pool Cooling Demineralizer System (FPCDS). This enhancement is being implemented in order to provide a standby system that can be remotely placed in service from the Control Room, if needed, when a high decay heat load is placed in the spent fuel pool or subsequent to a seismic event. This enhancement is to support Vermont Yankee's use of high density fuel storage racks in the spent fuel pool to increase the capacity of the spent fuel pool to 2870 spent fuel assemblies. Refer to Amendment 104 to Vermont Yankee license, reference (a), and EDCR 86412 for details of the fuel pool expansion and use of high density fuel racks. This amendment allows Vermont Yankee to use high density storage racks and place up to 2000 spent fuel assemblies (current limit) in the spent fuel pool. The remaining cells (for a total of 2870 spent fuel assemblies ) cannot be filled until the SFPCS has been installed and its proper operation demonstrated.

In letter FVY 88-17, reference (bb), Vermont Yankee provided the design criteria of the SFPCS to the NRC for review. These design criteria ensure the SFPCS satisfies Standard Review Plan 9.1.3

" Spent Fuel Pool Cooling and Cleanup System", reference (b). The SFPCS of the FPCDS is a two train, Seismic Class I, Safety Class 3 System designed to prevent a single active failure or common event from disabling both trains such that the ability of the system to remove decay heat is compromised. The essential electrical equipment contained in the SFPCS receives power from redundant Safety Class Electrical (SCE) power sources with cables physically separated. All components of the SFPCS that are in contact with fuel pool water are constructed of corrosion resistant material. This portion of the FPCDS provides cooling of the spent fuel pool by transferring the spent fuel decay heat to the Service Water System (SWS). Each train of the SFPCS contains one pump, one heat exchanger and associated valves, instrumentation and controls. The pump circulates the pool water in a closed loop, taking suction from the spent fuel pool through the heat exchanger and reto.rning it back to the pool. The heat exchanger is cooled by the seismically qualified safety class SW3.

The SFPCS is separated from the Normal Fuel Pool Cooling Subsystem (NFPCS) by a combination of motor operated valves (MOVs) and check valves. These valves provide isolation of the nonseismic NFPCS from the seismic SFPCS The components used in the SFPCS are seismically qualified and supported. Since isolation.of the seismically designed SFPCS from the nonseismic NFPCS is provided and a seismically qualified source of cooling water (the SWS) is used to cool the heat exchangers, cooling of the fuel pool following a design basis seismic event is ensured.

One train of the SFPCS is designed to maintain the fuel pool water below 150 'F after a normal refueling that completely fills the pool with spent fuel assemblies. This subsystem has the flexibility to be operated in either single train (one pump, one heat exchanger) or two train operation (two pumps, two heat exchangers or one pump two heat exchangers). This flexibility allows the SFPCS to remove the decay heat generated by a full core discharge or from normal refueling operation in the most efficient mode of operation for the condition assuming NFPCS is not available.

A positive differential pressure is maintained in the heat exchangers to protect against any possible fuel pool water leakage into the SWS. The fuel pool side of the heat exchangers has a maximum pressure equal to the static head developed by the difference in elevation between the heat exchanger and the pool surface. The minimum operating SW pressure is greater than the pressure in the fuel pool side of the heat exchanger.

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EDCR 89-408 l

l The electrical power to essential components (pumps, valves, and essential instrumentation) is configured so that no single failure of a power supply will disable the SFPCS. The power for the components in one train is supplied from a Safety Class Electrical power supply and the power for 1

the components in the second train is supplied from a different Safety Class Electrical power supply. The two isolation valves in the NFPCS are powered from separate Safety Class Electrical power supplies to ensure that the isolation of the NFPCS from the SFPCS can be accomplished when

needed, 1

l All essential electrical components, located in a harsh environment, required to ensure proper L

operation, control and monitoring of the SFPCS are environmentally qualified (EQ) under the direction of Vermont Yankee's EQ Program in accordance with 10CFR50.49.

Essential controls and monitoring for the SFPCS include pump on/off switches, service water throttle valve controls, isolation valve controls, and instrumentatica to monitor the performance of the SFPCS These are located in the Control Room to control pump operation, detect pool temperature, and control SWS flow. Components located in the Control Room are in :. ild environment and are not subject to the requirements of 10CFR50.49.

The schedule for implementing this design change is by the end of tb 1993 refueling outage. The estimated cost to complete this enhancement is approximately $ 3,000,000. A cost breakdown into construction, material, and engineering is not feasible at this time since vendor bids have not been reviewed, construction walkdown for final cost estimate has not been completed and final quantities cf materials are not defined.

Several licensing documents will requhe revision as a result of this enhancement. They are; the FSAR, EQ Program, and the ISI program. The preliminary design details of this enhancement to the FPCDS and any impact it may have on the safe operation of Vermont Yankee has been evaluated by the NRC. The results and conclusion of that safety evaluation are contained in NRC letter NVY 88 223, reference (c), which concluded that the enhancement to the FPCDS is acceptable.

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EDCR 89-408 2.0 Table of Cont #nta 1.0 Executive Summary 1

2.0 Table of Contents 3

3.0 Design Criteria 4

q 3.1 Quality Assurance 4

3.2 Technical 4

4.0 Description of Change 7

4.1 Scope and Intent 7

4.2 Operational Description 7

4.3 Impact on Design Basis 8

4.4 Implementation Considerations 15 4.5 Other Design Considerations 19 4.6 Operational and Maintenance Considerations 21 5.0 Resources 22 5.1 Organization (s) Responsible for the Detailed Design 22 5.2 Cost Estimates 22

j 5.3 Schedule 23 6.0 Open Items 23 Enclosures

,1 A

Safety Evaluation

'4 B

Affected Licensing & Controlled Documents.

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C Calculations

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D Procedures E

Procurement Information - Bill of Materials F

Drawings G

References II Review Forms l:

I Vendor Documents (later)

• A list of Calculations is provided. The calculations are on file at YNSD.'

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EDCR 89-408 3.0 Damian Critarin j

3.1 Onnlity Annurance The following documents shall be used to provide the Quality Assurance required for thia design change.

a. Yankee Operational Quality Assurance Manual (YOQAP 1 A), Rev.19.

b Y:nkee Atomic Electric Company Engineering Manual, Rev. 30.

c. Vermont Yankee Safety Classification Manual, Revision 3.

d. ANSI N45.2.11973, Cleaning of Fluid Systems and Associated Components During the Construction Phase of Nuclear Power Plants, e.

ANSI N45.2.2-1972, Packaging, Shipping, Receiving, Storage and Handling ofItems for Nuclear Power Plants,

f. ANSI N45.2.4-1972, Installatien, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Plants,

g. ANSI N45.2.61978, Qualification of Inspection, Examination, and Testing Personnel for the Construction Phase of Nuclear Power Plants.

h. ANSI N45.2.81975, Supplementary Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems for Constru, tion Phase of Nuclear Power Plants.

l 3.2 Technient Renuirements The following documents provide the technical direction for the design of the SFPCS. Unless -

otherwise specified, the code / standard issue or addenda in effect at the time of purchase shal be used.

1 3 21 Mechantent Reinted I

a. Vermont Yankee Final Safety Analysis Report (FSAR) Revision 7.

b. Vermont Yankee Safety Classification Manual, Revision 3.

c.

EBASCO BWR

" Specification for Piping, Piping Components, Hangers & Supports for Station Piping Systems", BWR QC 10, Sept.15,1968.

" Insulation", VYNP III I 1, Rev.2, December 23,1969

d. ANSI B31.11977 Power Piping Code.

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EDCR 89 408 e.

American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel j

Code".

Section XI " Rules for Inservice Inspection of Nuclear Power Plant Components" 1980 Edition including Winter 1980 Addenda.

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ANSI B16.5 " Steel Pipe Flanges, Flanged Valves, & Fittings".

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ANSI B16.9 " Factory Made Wrought Eteel Buttwelding Fittings".

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h. ANSI B16.10 */ ace to Face End Dimensions of Ferrous Valves".

i. ANSI B16.11 " Forged Steel Fittings, Socket Welding and Threaded".

J. ANSI B16.25 "Buttwelding Ends".

k. ANSI B16.34 " Valves. Flanged and Buttwelding Ends".

1. ANSI B36.10 " Welding and Seamless Wrought Steel Pipe".

m. ANSI B36.19 " Stainless Steel Pipe".

n. EPRI Report NP.5639 May,1988 " Guidelines for Piping Systems Reconciliation" (NCIG 05, Revision 1).

j o.

WRC Bulletin 316 July,1986 " Technical Position on Piping Installation Tolerances".

p.

Nuclear Regulatory Guides: RG 1.60 Rev.1, RG 1.92 Rev.1, RG 1.122 Rev.1 and ASME Code Case N411(PVRC Damping).

q. American Institute of Steel Construction (AISC) Manual of" Steel Construction" Allowable Stress Design 8th Edition.

r.

American Welding Society (AWS) " Structural Welding Code" D1.188.'

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Ebasco Specification for Vermont Yankee " Masonry" No. 5920 A 12, dated 9/19/68.

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American Concrete Institute (ACI), " Building Code Requirements for Reinforced l;

Concrete" (ACI 318-63), (Used for evaluation of original plant construction) a l

u. American Concrete Institute (ACI), " Building Code Requirements for Reinforced v.

Concrete"(ACI 318-83, Revised 1986). ( Used for design and evaluation of new structures i

l' and foundations).

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v. American Concrete Institute (ACI), " Specifications for Masonry Structures" 2

(ACI 530.188).

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w. American Concrete Institute (ACI), " Building Code Requirements for Concrete i

9 Masonry Structures"(ACI 530 88).

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NOTE: New cou; rete / masonry construction will incorporate references u., v., and w. as d,

necessary.

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EDCR 89408 1

x. USNRC Standard Review Plan 9.1.3 " Spent Fuel Pool Cooling a sd Cleanup System" Rev.1, dated July 1981.

y Uniform Building Code (UBC),1988 Edition,

z. Manufacturers Standardization Society MSS-SP 61, " Pressure Testing of Steel Valves",

aa. Manufacturers Standardization Society MSS-SP 84, " Steel Vahes. Socket Welding and Threaded Ends".

ab. ASME Section II 1977 Edition orlater, ac. ANSI B16.36 " Orifice Flanges" 1988 Edition.

3.2.2 Elnetrien1 And Inntmment and Control S1stad

a. IEEE Standard 279 1971

• Criteria for Protection Systems for Nuclear Generating Stations."

b. IEEE Standard 323 1974 "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."

c.

IEEEStandard 338 1977 " Standard Criteria for the Periodic Testing of Nuclear Power Generating Stations."

d. IEEE Standard 344 - 1975 "IEEE Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations."

e. IEEE Stanuard 383 - 1974 " Type Test of Class IE Electrical Cables, Field Splices, and Connections for Nuclear Power Generating Stations."

f. Vermont Yankee, " Ground Rules for Separation and Identification of Reactor Protection and Safeguards System. Related Electrical Equipment and Wiring",

Revision 3, June 7,1971.

g. 10CFR50.49," Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants".

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h. Vermont Yankee Demarcation Standard

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i. Vermont Yankee Mimic Color / Size Standard J. Vermont Yankee DCRDR Label Standard

k. Vermont Yankee Abbreviation standard i

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EDCR 89-408 4.0 Demerintion of Change 4.1 Scone anel Intent The intent of this design change is to address an NRC concern (relative to sufficient spent fuel cooling capautv) expressed during their review of Vermont Yankee's proposed change (ref. (a)) to increau th e capacity of the spent fuel storage pool from 2000 to 2870 fuel asseinblies. This sneern will be resolved by installing an enhancement to the existing Fuel Pool Cooling system that will meet the criteria of Standard Review Plan (SRP) 9.1.3 Rev.1 as requested by the NRC.

This design package provides the design details for the Standby Fuel Pool Cooling Subsystem (SFPCS) to be added to the Fuel Pool Cooling System. This subsystem will be used when the existing Normal Fuel Ptol Cooling Svstem (NFPCS) may not be capable of maintaining the pool temperature within the limits s ~ ulated in the Technical Specifications.

The scope of the design eacompasses all facets required in the design of a new system. These are: determining system operating and performance requirements, providing Safety Class Electrical power to the components, providinginstrumentation and control for the safe operation of the system, and the structural and mechanical design of the process piping and equipment. Each of these areas are addressed in detailin the body of this EDCR.

The following is a list of the major components and instrumentation that will be required for this design change; Two pumps Pool water level instrumentation Two heat exchangers Pool temperature instrumentation Four Motor Operated Valves Manual valves (including checks & reliefs)

Heat Exchanger DP instrumentation Piping and Fittings The design is performed in accordance with the quality and technical requirements specified in Section 3.0 of this EDCR. This will ensure that the subsystem is designed to provide the required cooling to the spent fuel pool when needed.

l 4.2 Oncrationni Denerintion i

l The SFPCS is designed to be used in the event the existing NFPCS may not be capable of l

maintaining the pool temperature within the Technical Specification limits. This could occur if a seismic event disabled the existing NFPCS or if the decay heat load in the spent fuel l

pool exceeds the heat removal capacity of the NFPCS. Indication of the pool water level and temperature are provided in the Control Room. These parameters will alert the operators to changing conditions in the fuel pool.

I If the pool temperature is approaching the Technical Specification limit, the operators will secure the NFPCS pump (s) and start the SFPCS remotely from the Control Room. The return i

side of the SFPCS is isolated from the nonseismic NFPCS by two check valves in the return l

line of the NFPCS. These isolation check valves ensure that the SFPCS flow is isolated from short circuiting through the non seismic NFPCS, When the SFPCS is in operation the pool water bypasses the filter demineralizers.

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l EDCR 89-408 The SFPCS provides cooling of the spent fuel pool by transferring the spent fuel decay heat to i

the Service Water System (SWS). Each train contains one pump, one heat exchanger, and associated valves, piping and instrumentation. The pump circulates the pool water in a closed loop, taking suction from the spent fuel pool through the heat exchanger and returning it back to the pool.

f In the event of a line break in the non seismic portion of the NFPCS, the motor operated i

isolation valves will close when the pool level reaches a predetermined level, approximately r

23.5 feet above the top of active fuel. This will prevent furtherloss of pool inventory and maintain pool level such that the SFPCS will not loose the ability to provide pool cooling.

L One pump and one heat exchanger alone are designed to provide suffici.nt flow for the l

maximum normal heat load from a normal refueling discharge. For an abnormal heat load, l

such as full core discharge, one pump and two heat exchangers, or two pumps and two heat exchangers can be placed in operation as necessary to maintain pool temperature.

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In order to protect against any possible fuel pool water leakage into the SWS, a positive differential pressure is maintained. The fuel pool side of the heat exchangers has a maximum pressure equal to the static head developed by the difference in elevation between the heat exchanger and the pool surface. The minimum operating SW pressure is greater than the pressure in the fuel pool side of the heat exchanger. Indication and alarm are l

provided in the Control Room if the positive pressure between the SW side and the fuel pool side of the heat exchanger falls to a predetermined set point.

The power and control to the pumps and valves provide the capability ofisolating the NFPCS -

l and remotely placing the SFPCS in operation. Following any scenario, such as a fire in the Reactor Building or seismic event which may disable the NFPCS, the SFPCS can be placed in operation from the Control Room, b

Indication is provided in the Control Room and locally near the Fuel Pool Cooling cubicle.

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Control Room indication for each train includes direct pool temperature, fuel pool water level, fuel pool water temperature out of the heat exchangers, pump run lights, pump discharge l

pressures, service water flow, SWS to SFPCS heat exchanger differential pressure and valve j

position lights. Local indication inclu-les fuel pool water temperature into the heat exchangers, SFPCS pump suction and discharge pressures, SFPCS heat exchanger pressures, e

L and SWS temperatures.

1 4.3 Imnnet on Denism Banaa h

The Engineering Design Bases and Section 10.5.2 of the FSAR state that the safety objective of the fuel pool cooling system is to "... maintain fuel pool water temperature at a level which c

r will prevent damage to the fuel elements, and to maintain the Reactor Building environment a

at a level which will bound the qualification of electrical equipment." In addition to the abovo safety objective the fuel pool cooling system shall monitor the pool water level and maintain t

water level above fuel sufficient to provide shielding for normal building occupancy.

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EDCR 89 408 The following design requirements, incorporated into this enhancement to the fuel pool cooling system, ensure that the above design bases are fulfilled consistent with the guidance provided in SRP 9.1.3, Rev.1 (reference (b)), and NRC Safety Evalution (reference (c)).

l a) Ability to maintain the pool temperature below 150 *F with 2,870 spent fuel assemblies in the pool, b) Ability to provide cooling during normal, abnormal, and accident conditions, c) Means to prevent loss of function resulting from single active failure or failures of non safety class components or systems, d) Ability to maintain pool water level (detect and isolate nonseismic portion of the NFPCS),

e) Fulfill design bases without impacting other existing systems (ie, impact on SW and emergency power sources), and f) Provide instrumentation and controls to initiate, monitor and shutdown the system from the Control Room.

How each of these principal criteria are incorporated into the design will be described below, a) Ability to maintain the pool tempert are below 150 *F with 2,870 spent fuel assemblies in the pool, The SFPCS is sized so that the system can remove the decay heat load generated when all.

l 2,870 cells are filled with spent fuel assemblies and a full core discharge. The heat load is calculated for two scenarios. The first is the normal maximum heat load, which is the heat generated by spent fuel when the pool is filled (2,870 assemblies) with successive regular refuelings (1/3 of the core each refueling) assuming a failure of one train of fuel pool cooling.

The second is the abnormal maximum heat load, which is similar to the normal maximum heat load except that the last 368 cells are filled by a full core discharge and no failure is assumed. The design indicated that a two train system each train having one 700 gpm pump and one 11.0 MBlu/hr tube and shell heat exchanger would be adequate to remove the heat l

from the two scenarios presented above. The calculations for the determination of the heat l

loads are documented in reference (g), and the sizing calculation for the heat exchangers in VYC 889.

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b) Ability to provide cooling during normal, abnormal, and accident conditions, l

The SFPCS is designed to provide cooling under all licensed plant conditions. This subsystem is designed Seismic Class I using the Seismic Class I SWS to remove spent fuel decay heat to the ultimate heat sink (Connecticut River). This provides another seismically qualified means to cool the fuel pool in addition to the Residual Heat Removal system in the Augmented Fuel Pool Cocling Mode.

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EDCR 89-408 Essential electrical components are environmentally qualified per Vermont Yankee's Environmental Qualification Program to ensure operability under design basis accident conditions (Enclosure "B"). Power and control to the subsystem are provided imm redundant Safety Class Electrical sources, This ensures that one train of the subsystem will receive power and control in the event of a loss of power scenario. Power and control to the pumps and valves provide the capability ofisolating the NFPCS and remotely placing the SFPCS in operation, following any scenario, including a fire in the Reactor Building or seismic event which may disable the existing fuel pool cooling system. The SFPCS can be remotely placed in operation from the Control Room as needed to maintain an acceptable pool temperature.

Therefore, since the subsystem uses a completely seismically designed piping system, redundant Safety Class Electrical power sources, environmentally qualified components, and can be remotely operated, the SFPCS has the ability to provide cooling during plant normal, abnormal, and accident conditions, c) Means to prevent loss of function resulting from single active failure or ailures of Non. Nuclear Safety class components or systems.

The power and control to the SFPCS is provided from redundant Safety C. ass Electrical buses. The "A" loop components (P19 2A, V19-220, and V70 257A) are sul plied from electrical Division II (SII) Motor Control Center (MCC) 9B. Cables for these compone nts are routed entirely in SII raceways. The "B" loop components (P19-2B, V19 221, and "70 257B) are supplied from electrical Division I (SI) MCC 8E. Cables for these components are routed in SI raceways. This design prevents a single active electrical failure from disabling the decay heat removal function of both trains of the SFPCS.

The essential instrumentation for monitoring and controlling the SFPCS is also separated by train. The instrumentation for the "A" train of the SFPCS has its power supplied from the 120 v Instrument AC power supply. The power, control and signal cables for the "A" train are routed through SII raceways. The "B" train instrumentation is powered by the Vital 120 VAC source with the power, control and signal cables routed through the SI raceways. This separation ofinstrumentation ensures that no single active failure would prevent indication, control, and monitoring of the SFPCS.

A design review and walkdown of the new and revised piping was performed to ensure that no internally generated missile or seismic interaction of plant equipment could be postulated that could affect both trains of the SFPCS. Refer to reference (dd) and (ee) for detail of the review.

A portion of the service water piping including manual valves SW 23F, SW 230 and SW 302 has been evaluated and classified nuclear Safety Class 3 (Enclosure E and Ref.(gg))

consistent with the safety classification and importance to safety of the SFPCS supply and return lines.

Flooding of the Fuel Pool Cooling System cubicle and Reactor Building was evaluated in Reference (ff) to ensure that no postulated fluid system rupture on elevatior 303 could affect t

both trains of the SFPCS and/or other essential plant equipment The physical separation of cables, controls, and equipment provides for adequate fire protection. This issue is discussed in greater detailin Section 4.5 of this EDCR.

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EDCR 89-408

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N The piping is seismically designed and supported and the components in the' system are qualified per the appropriate guidelines and/or requirements. The SFPCS is a two train system with one pump and one heat exchanger per train. Each train has separate valves to i

provide isolation of the operating train from the nonoperating train. Two isolation valves are installed in the supply line of the NFPCS and two check valves in the return line of the NFPCS. These seismically installed isolation valves (supply and return sides), ensure that M

the seismically qualified SFPCS will be isolated from the nonseismic NFPCS. The power to these valves is from separate Safety Class Electrical sources to ensure that one of the isolation valves will close if needed during a Loss of Offsite Power event concurrent with a single failure in the electrical distribution system. Since the SFPCS modincation is seismically 1

designed, uses seismically and environmentally qualified equipment, has been reviewed for missiles, fire and flooding, provides for separation of nonseismic from seismic portions and provides for separation of operating from nonoperating portions, it can be concluded that no single mechanical failure or event can disable the SFPCS.

d)

Ability to maintain pool water level (det4.ct and isolate nonseismic portion of the NFPCS),

This enhancement to the NFPCS will provide level instrumentation to constantly monitor the l

water level in the spent fuel pool. This instrumentation consista of level transmitters l

mounted at the spent fuel pool, level indicators located in the Control Room and associated 2

1 cables and power supplies required to support the measurement of fuel poollevel. This system provides for constant remote monitoring of the water levelin the fuel pool. The level measurement is powered by separate power supplies and the cables are separated by using separate raceways. The components are qualified for the environments created by the desiga e

d basis events. The level transmitters provide a signal to the two isolation valves in the supply l

line of the NFPCS (8" FPC-1B). In the event of a line break in the nonseismic NFPCS, the pool 4

l level decreases and the transmitters automatically signal the isolation valves to close at r.

)

predetermined setpoint approximately 23.5 feet above the top of active fuel. This isolates the nonseismic NFPCS from the SFPCS and maintains pool water level above the pool outlet to the SFPCS, providing proper radiation shielding and maintaining the ability of the system to i

remove decay heat.

Once the leak is isolated normal level can be restored using existing makeup paths l!

identified in Plant procedures. In addition a new seismically qualified make up sov.rce is being supplied by this enhancement. A two inch branch line from the SWS will provide an alternate makeup source, if needed, after a seismic event. The makeup is normally isolated by two closed manual valves and a spectacle flange. This arrangement is similar in design to the existing cross-tie to the Residual Heat Removal system.

L By having a redundant dedicated pool level monitoring system which isolates the-nonseismic portion of the NFPCS, it is assured that sufficient pool water level will be maintained above spent fuel to provide shielding for normal building occupancy and meet j

Technical Specification requirements for water level.

OL e) Fulfill design bases without impacting other existing systems (ie, impact on SW and ll emergency power sources).

d This enhancement has been reviewed for impact to other systems. The systems the SFPCS interfaces with directly as well as those that may be affected were examined.

L' 14 11 h

L

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EDCR 89-408 1

j The power, control and instrumentation circuitry has been designed so no single active failure of the existing circuitry or the components in the SFPCS can disable the SFPCS or the l

existing power and instrumentation circuits. This is achieved by providing separation of cables, redundant power supplies, protective devices to isolate the supplies from the SFPCS components, and using qualified safety class components.

The effects on safe shutdown equipment have been analyzed by a review of bus and diesel generator loading. This impact is negligible since the SFPCS pump loads are essentially the f

same as the NFPCS pump loads, and only one (SFPCS or NFPCS) load is applied at any time.

(

The only process piping system the SFPCS will have impact on is the SWS. The SWS provides p

the seismic source of cooling to the SFPCS heat exchangers and transfers the heat to the ultimate heat sink, the Connecticut River. Calculation VYC 855 " Fluid Flow Analysis of the

/

Service Water System at Vermont Yankee" reviewed the existing flow in the CWS to ensure j

the system could provide the additional flow required to cool the SFPCS heat exchangers with L

no detrimental effects on other Safety Class equipment. The calculation determined that the L

SWS can support cooling the SFPCS heat exchangers and still fulfill its safety function.

flowever, VYC-855 identifies a minor flow imbalance that needs to be addressed to insure sufficient post accident heat removal by the ECCS corner room RRU's (RRU 5,6,7, and 8). See l

Section 4A.3 for further discussion.

This enhancement has no other detrimental impact on existing plant systems. The piping

[

and equipment are designed so no failure of the components in the SFPCS can impact other j

plant systems. The piping in seismically supported and isolated from other plant systems by j

the location of the equipment, f) Provide instrumentation and controls to initiate, monitor and shutdown the system from the Control Room.

The SFPCS la provided with instrumentation to monitor and control the operation of the i

system. The controls and indicators are provided in the Control Room so operation from -

1 outside the Reactor Building can be accomplished. The layout of the instruments on the j

panels has been reviewed for human factors to obtain the optimum layout. In addition a

(

mimic depicting major equipment and system relationships is to be added as an additional f

aid to operators.

The parameters considered essential to ensure that the system is functioning properly are; SW flow to the SFPCS heat exchangers, l

direct temperature monitoring of the spent fuel pool water, i

f monitoring of differential pressure between SFPCS and SWS sides of heat exchanger, j

measuring spent fuel pool level for closure ofisolation valves and makeup SFPCS pump indicating lights and suction pressure trip and NFPCS isolation and SWS throttling valves operation and position.

Monitoring the SW flow to the SFPCS heat exchangers (PT 104-80NB, FI-104-80NB) and the pool temperature (TE 19-71A 1,B 1; TI 19 71NB), provides the operators with the information necessary to ensure that the SFPCS is functioning properly in maintaining the pool below 150

• F. Presently Fuel Pool temperature is recorded and alarmed in the Control Room from measurements provided from thermocouples located in the pipes of the inlet and outlet of the NFPCS pumps (P91NB). Therefore, unless pumps are operating there does not exist accurate i

indication of the Fuel Pool temperature. This modification will add instrumentation to be installed in the Fuel Pool to provide a direct indication of Fuel Pool temperature. Redundant 12

EDCR 89-408 environmentally and seismically qualiaed thermocouples will be idstalled at adjacent corners of the Fuel Pool and will be submerged to a depth of approximately 5 feet into the pool.

These thermocouples will be tri element (6 wire) type which will provide signals to safety related indicators tc be installod in the Control Room. One indicator will be installed on CAD Panel A and the other on CAD Panel B. In eddition, the thermocouples will provide input to the new ERFIS computer system and the existing recorder which presently records the existing Fuel Pool temperature. These new inputs to the recorder will be used to activate the alarm switches in the recorder which are presently actuated by the existing thermocouple.

Therefore, the annunciation of "High Fuel Pool Temperature" will be provided by the new direct readira measurement thermocouples.

Providing constant indication of the diff'erential pressure (dPT 19 76A/B, dPI.19 76A/B) between the SWS and SFPCS sides of the heat exchangers ensures fuel pool water will not enter the SWS. This indication and alarm will alert the operators of the decrease in the differential pressure. The operators will then take corrective action (reposition the Dow control valve on the SWS discharge of the SFPCS heat exchanger) to re establish the differential pressure.

The existing thermal probe instrumentation used to measure and alarm Fuel Poollevel has proven to be unreliable in the past and is difficult to calibrate. In addition, this L instrumentation only provides information at two levels in the pool (high and low). This design change will replace this existing instrumentation with a transmitter capillary seal system. The seal will be submerged into the pool to measure the head (level) of water above it.

This instrumentation will measure level from approximately 6 to 31 inches below the top of the pool. In addition to performing all the functions which are currently performed by the existing instrumentation, this new level system will be provided with indicators to be located in the Control Room on each of the CAD Panels. One indicator is provided with the Safety Class Electrical master trip unit.md a second isolated NNS indicator is added in order to provide mimic indiention. These new instrument level measuring loops will also provide a signal to close NFPCS isolation valves V19-220 and 221 in the event of an accident resulting in low low Fuel Pool level. New instrumentation will be located in the same vicinity of the instrumentation which it is replacing as shown on Drawing 5920-6340.

By monitoring the water level (LT-19 63A/B, LT-19-63A(M)/B('M)) in the fuel pool it is ensured that sufficient water level can be maintained in the pool to provide shielding for normal building occupancy and meets Technical Specification requirements. The monitors provide a signal to irolation valves in the supply line to the NFPCS to close automatically.

when the water level in the pool reaches a predetertained level. This action ensures that the pool water level will not fall below safe levels as a result of a leak in the NFPCS.

The redundant Safety Class Electrical master trip units (LT 19 63A(M),B(M)) located in the Control Room on each CAD Panel are prosided with meters to continuously monitor fuel pool level. In addition, a combination analog / digital NNS indicator is provided on each CAD.

Panel, located within the mimic arrangement. This provides ar.other means of monitoring fuel pool level. These indicators of fuel pool level are utilized to control pool make up.

l 13

.f EDCR 89-408 b

L Monitoring the suction pressure (PS 19;80NB) of the SFPCS pumps provides protection of the pumps. If the pump suction pressure falls below a predetermined pressure, the pumps will be tripped off to prevent damage to the pumps. Also the pressure switch that provides this protection must not fail and trip the pumps. For this reason the low Suction pressure circuit q

and components in the circuit are safety class and EQ qualified. The controls for the pumps g

and isolation valves are required to function properly to ensure that the system fulfills its E

design bases. The components in these circuits are also safety class and environmentally qualified.

All the essential electrical components in the instrument loops are Safety Class Electrical and mechanically SC3 to ensure that the components will function when required. The H

instrumentation is redundant and is powered from separate safety class power supplies as j

discussed above.

O L

In addition to the above, the existing local pump controls of the NFPCS have been relocated to i

the CAD Panels as described in Section 4.4 q

i The other controls and indication, not considered essential, but provided as an aid in normal system surveillance to the operators in the Control Room are; 1

• Spent Fuel Pool Level (LI 19-63NB) (Provided in addition to the indicator on LT 19 63A(M), B(M) to complete the mimic arrangement)

• SFPCS Pump Discharge pressure (PI 19 81A, B)

• SFPCS Heat Exchanger Outlet Temp, (TI 19 73A/B) i The following annunciation is also provided in the control room for operator information i

and to alert them of possible trouble with the SFPCS, l

• Fuel Pool Cooling Emergency System Trouble, This consists of;

• SFPCS Pump I4w Suction pressure Trip, 4

• SFPCS Pump Low Discharge pressure,

• SWS to SFPCS Differential pressure,

)

• Fuel Pool Temperature,

{

• Trip Card File Trouble and Gross Failure, and

• Existing Trouble Annunciation on High & Low Fuel Pool Level, g

i 0

In addition to the Control Room instrumentation, the following instruments are provided l.;

locally in the area of the SFPCS. This instrumentation is mounted on new seismically l:

designed instrument racks unless otherwise noted. Individual indication is provided for j

both the "A" and "B" loops unless designated as " common."

H 1

SFPCS Pumps P19 2NB Discharge Flow (Common)(FI 19 73)

{

SFPCS pressure at SFPCS Heat Exchangers E19-2NB (PI 19-82A, B) g SWS pressure at SFPCS Heat Exchangers E19 2NB (PI 104-93A, B)

H SFPCS Pumps P19-2NB Suction pressure (PI 19 84A, B) i SFPCS Pumps P19-2NB Discharge pressure (PI 19 83A, B)

M SFPCS Heat Exchangers E19 2NB Inlet Temperature (Common) ;

j Thermowell in pipe (TI 19 72) y SWS Temperature to E19 2NB (Common) ; Thermowell in pipe (TI 104-60) j SWS Temperature from E19 2NB (Common) ; Thermowell in pipe (TI 104-61)

U C.

l 14 1

EDCR 89-408 l

l All instruments connected to 803 SFPCS piping system's, whether they provide essential measurements or not, are classified mechanically as SC3L This assures that the proper l

Quality Assurance is provided to maintain pressure boundary integrity of the piping system at instrument connection.

t

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4.4 Imnlementation Conmiderations The midority of the installation work involved with this enhancement (running and supporting the piping, electrical cables, instrumentation lines and racks, installing the pumps, heat exchangers, etc.) can be done prior to the outage without impact to other systems.

The details of the installation (pipe location, instrument details, equipment mounting, support details, etc.) are provided by sketches and drawings found in Enclosure F of this i

EDCR. Piping insulation, where required by isometric drawings, shall be installed equivalent to the requirements of 3.2.1.c. Reference (p p), Enclosure G, outlines scope of support work.

Installation and testing of this design change shall be accomplished in accordance with the Installation and Test Procedure (I&T) developed as part of VY A.P. 6001 " Installation, Test and Special Test Procedures" incorporating the requirements of Section 3.1.

4.4.1 Electrieni and Inntrumentation and Control Considerating 4

Cable and raceways shall be installed in accordance with the attached Conduit, Trays and Grounding drawings and the Cable and Conduit lists. The routings are designed to provide physical and electrical separation to prevent a single failure from resulting in the loss of both trains. Where cable and/or conduit traverse a fire separation zone in the Reactor Building, l

the integrity of the separation zone shall be maintained.

Fire stop material shall be applied to the South end of conduits 46114 SI and 46115 SI, and to the North end of conduits 112230 SI and 112250 after cable is pulled through. Fire barriers from the Cable Vr It to the Reactor Building, and from the Cable Vault to the Control Room shall be restored rAer the cable and conduit installation. Conduits into the bottom of the CAD panels A and B shall be fire sealed within the Panels or within the plane of the Control Room floor.

Cable from the redundant instrument racks located in the Reactor Building will be routed in opposite directions in cor. luit to terminate at existing termination boxes located in the Reactor Building on the atme elevation. SII Division cable will be routed from Rack RK10AA i

to existing box TB-958-SII aad SI Division cable will be routed from RK10AB to existing box TB-954 SI. Conduit from RK10AB will be routed down to the next lower elevation and then back up to Elevation 303 to prodde separation from RK10AA conduit and to avoid crossing Divisions.

i 1

Presently TB 954 SI and TB-958-SII contain terminal blocks and cables routed from CAD Panels A and B in the Control Room. These existing terminal blocks and cables will be replaced since the documentation for these components is not sufficient to qualify them.

Cables are being routed to these existing boxes in order to utilize the existing conduit which runs to the CAD Panels in the Control Room. Conduit seals will be provided for the

)

i transmitter loops routed to these boxes and a drain hole la required at the bottom of the boxes.

In addition, a conduit seal will be added for instrument loops LT-19 63A and B at the first entrance to a tray or box originating from the transmitters.

Ib r

EDCR 89 408 Conduit required to be routed from the Cable Vault into the CAD Panels A and B can either be routed to the panels from directly underneath or off to the side and then up into the panels using elbow and flex conduit. However, conduit routed from trays in the Cable Vault up into:

the Panels should not block the Cable Vault to Reactor Building wall penetration area any more than absolutely necessary, i

Existing conduit 48972SI is presently routed from CAD Panel A down into the Cable Vault l

where it is capped. This conduit will be relabeled as 11225SSI and extended to connect to cable l

tray R449St.

Existing conduit 49082 which is routed from CAD Panel A to CAD Panel B will be relabeled 11225H.

Existing fuel pool level instrument cables C11210YSII and C11211RSI are presently routed from the fuel pool to Panel 20-22 in the Radwaste Building. When the existing level instrumentation is replaced, these cables will be removed and new cables will be routed from the CAD Panels to the same existing destination on Panel 20-22. These cables will be relabeled C1875B and C1874B. The existing means of penetrating the Radwaste Building with the existing conduit presently containing C11210YSI! and C11211RSI will be maintained.

"'he instrumentation will be located in the Control Room, or on new local instrument racks and some will be mounted directly to the process pipe.

Control Room panels, CAD Panels A and B are to be modified as shown on drawings 5920 10477 ShL 1 and 10478 Sht.1 and in accordance with the requirements included in Enclosure (F) on Sketches VYC-879 001 sheets 1 thru 5.

4 Before the panels are modified, all cover plates are to be removed and all be.'as sealed in accordance with the welding repair method included in Reference (11). After installation of equipment, the welding repair method of sealing included in reference (11) can be utilized to seal any possible remaining openings. After installation of equipment, panels can be touched up with paint as determined in the field.

Lines of demarcation as shown on drawings 592010477 Sht. I and 10478 Sht. I will be added by painting 1/16 inch wide black lines in the areas identified in accordance with Vermont Yank % Demarcation Standard and Demarcation Location guidance. Mimic materials and colors will be added in accordance with the Installation and Test Procedure and Vermont Yankee Mimic Color / Size Standard. Control Room labels are to comply to the Vermont Yankee DCRDR Label Standard and Abbreviation Standard. Nameplates are included in Enclosure (E), Section 20.0.

The NFPCS cooling pumps (P91A/B) can presently be controlled from either switches located in the Radwaste Building on CP 20 22 or the Reactor Building on RK10 and RK10A. The control switches presently located on Racks RK10 and 10A will be removed and relocated to CAD Panels A and B located in the Control Room. Thus, after implementation of this EDCR the NFPCS pumps (P91A/B) will be controlled from either Radwaste Building or Control Room. Holes left on Racks RK10 and 10A from removal of the switches will be covered and wires reconnected as shown on Drawings 5920 5298 and CWD Sh.1210,1211 included in Enclosure F.

The new instrument racks, RK10AA and RK10AB are to be located as shown on Drawing 5920-11259, with the exact location determined by the field. Rack RK10AA is to be located beside '

l existing Rack RK10A. It is desirable to maintain as much clearance as possible from the 18

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EDCR 89 408

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entrance to the fuel pool cooling pump cubical. As stated earlier the pump control switch for-pump P91A is to be removed from Rack RK10A and reinstalled on CAD Panel "B". With the removal of the pump control switch RK10A becomes an empty box containing wire terminations. Therefore, since RK10A should not need to be opened (except in the unlikely R

event of required cable terminal block maintenance), RK10AA can be located as close as j

possible to RK10A and still provide enough room to partially open the box for repair if d

necessary, l

Racks RK10AA and RK10AB are seismically designed and are to be faliricated in accordance with Drawings B 1911261, Sh. 300 & 30D in Enclosure F. The racks have been J

designed so that the equipment can be repositioned, if necessary, with YNSD Engineering approval.

O Tubing runs are to be sloped down 1/4 inch per foot in accordance with applicable drawings.

Deviation from sloping requirements should be referred to YNSD Engineering for disposition prior to implementation.

Fuel Pool level instrumentation is to be installed to replace the existing level instrumentation as shown on sketch VYC 859-001 Sht. I thru 7 included in Enclosure F. Some of the existing support steelis utilized in the support of the new level indicators. The level instruments will be removed and replaced one at a time. Satisfactory operation of each level p

instrument shall be established prior to removal of the remaining existing instrument. This p

will ensure that level indication is always provided. This practice should be followed unless alternative steps are taken to provide indication of pool level.

a The tube steel used to support the capillary seal should either be provided with markings or have a stainless steel ruler attached to it to verify pool level relative to seal submergence level. These markings should be at 1/4 inch increments as a minimum. The indications should bo visible while standing on the sides of the pool. Sketch VYI 89 408-1 in Enclosure F provides typical details for the marking of the tube steel. For calibration and setpoints refer to Sketch VYI-89-408 2, ShL 1 thru 3 in Enclosure F.

The support for the level indicator located in the Southeast corner of the pool will also support the temperature probe.

0 9

The installation of the new thermocouples installed in the fuel pool are as shown on Sketches b

VYC-859 001, Sht. I thru 7 in Enclosure F. As stated above the supporting of the probe in the L

Southeast corner will be combined with the support of the level instrumentation. Exact locations of the temperature probes will be determined by the field.'

The thermocouples utilized are direct immersion Type T assemblies and do not require a thermowell. The probes are provided with pigtails which are to be spliced to thermocouple s

i extension cable using qualified splices and Raychem heat shrink tubing. Shims are

?

required to be provided in accordance with Raychem Procedures. WCSF-050 tubing will be installed on the #18 T/C pigtails prior to installing the WCSF 115 over the entire splice. The WCSF 050 acts as a shim. The WCSF 115 must extend at least two inches beyond the ends of the splice. Raychems " Typical Kit Documentation" and "WCSF N Application Guide" should be followed during installation.

1

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Cable from the redundant instruments located on the refueling floor are to be routed in flex conduit through the cable trenches on opposite sides of the fuel pool to provide physical separation, c

4 17

-,,, ~ ~

EDCR 89406i r

4.4.2 Mechanleni and Structural Canaldarations The block wall (G 19114918) located adjacent to the fuel pool pump and heat exchanger cubicle is being partially removed to facilitate installation of the heat exchangers and pumps.

The wall will be reconstructed per Sketch VYC-893 001 Sh.1 thru 9 in Enclosure F. The.

reconstructed wall designed to provide radiological shielding (Ref. hh) has been analyzed consistent with Vermont Yankee's response to I.E. Bullet.in 80-11 and is seismically designed to withstand VY's design basis seismic event. The construction of the wall will conform to VY's construction specification for masonry walls.

The piping and supports will be installed in accordance with the support sketches and piping isometries contained in Enclosure F. The construction and installation of the piping and 3

supports will conform to the General Notes on Drawing VYS-GN.EDCR 89 408 in Enclosure F.

The foundations for the equipment (heat exchangers and pumps) will be constructed as shown on Sketches VYC-8901 through 5, Enclosure F. The construction of the foundations will be in accordance with the General Notes on the foundation sketches. NOTE: The equipment details such as size, weight and mounting details are not completely known at this time. Therefore the final foundation details will be by ECN, The floor has been reviewed for the expected loads and it is adequate for the increased loading.

4.4.3 Performnnee Tentina Considerations t

Flushing / cleaning shall be in accordance with ANSI N45.2.11973 and good construction practices. Flushing / cleaning break points for new or revised lines and equipment will be detailed in the I&T procedure.

Hydrostatic testing shall be in accordance with applicable codes. Modified sections of SWS and NFPCS can be hydro tested ASME XI and new piping that can be isolated from existing piping can be hydro tested to B31.1.

Pump performance testing will be performed using the vendors recommendations. This inplant testing should include at least 5 points on the pump curve including shut-off head and design points. The tests should also verify acceptable vibration levels and bearing -

temperatures.

Heat exchanger performance testing will be performed in accordance with plant procedures.

The largest available heat load should be used. Special consideration will be given to post -

installation verification of zero tube leaks and tube to tube sheet leaks.

1 System performance for both the SFPCS and the SWS will be performed by combined system i

function testing. Data recorded during this test may be extrapolated to verify design performance. Complete functional testing of all control logic and instrument loops shall be documented in the I&T Procedure, i

Motor Operated Valves shall be functionally tested to insure proper operation, (including stroke time, running motor load, limit and torque switch settings). Existing plant procedures may be utilized and referenced in I&T Procedure.

i

EDCR 89-408 4.5 other D alan canalderatinna I

4.5.1 Environmental and Selamle Canalderatinna The cables, MCC components, valve actuator switches, valve and pump motors, and essential instrumentation for the SFPCS located in a harsh environment are qualified for the postulated environmental conditions outlined in the W EQ Program. Enclosure B includes revisions to Volume 1 and 2 of the EQ Program Manual. This revision adds environmental qualification requirements for the SFPCS equipment and removes the NFPCS and RBCCW components from the EQ Program. The QDR revisions will be by ECN when the specific components are known.

4 Essential transmitters, thermocouples, and pressure switches are environmentally qualified in accordance with IEEE 3231974. The environmentally qualified conduit seals i

are provided where required to prevent steam and/or moisture from affecting instrument i

measurements.

All safety class equipment to be installed in the Control Room on CAD Panels A and B 89 seismically qualified. Safety Class Dixson indicators, Electroswitch control sw%es, and Rosemount Trip Card Files are seismically qualified in accordance with IEEE 3441975.

All equipment on the panels is to be seismically installed. Calculation WC 879 addresses the addition of this equipment to the Control Room panels. In addition, WC 879 includes the seismic installation requirements for power supplies to be located within the panels. It has been determined that the seismic integrity of the CAD Panels is not adversely affected by the addition of this equipment. Using the existing CAD Panels in the Control Room provides for separation between the redundant pieces of equipment.

The instrument racks in the Reactor Building (RK10AA and AB) are seismically designed and all equipment to be placed on the racks will be seismically installed. Calculation WC-883 provides the seismic analysis of the rack and installation of the equipment. Safety class transmitters and pressure switches are seismically qualified per IEEE 3441975.

Redundant racks are provided so that no single active failure can disable both instrument 1

divisions. Cables to and from these instruments are routed to conform to VY separation criteria (Section 3.2.2 (0) as shown on drawings included in Enclosure F.

New instrumentation to be added in the fuel pool to measure level and temperature is seismically qualified in accordance with IEEE 344-1975 and will be seismically supported.

Calculation WC 859 addresses the seismic installation of the pool instruments. The pool instruments are located at opposite or a$acent corners of the fuel pool as shown on Drawing G 191338 in Enclosure F. Cables routed from these instruments are located in cable trenches on opposite sides of the Fuel Pool. The cable separation is maintained to the CAD Panels in the Control Room.

All existing safety class equipment in the ECCS corner rooms, required for long term post LOCA operation, have been evaluated for a potential room temperature increase due to lower service water flow through the room coolers (reference (qq)). References (rr) and (as) document this evaluation and verify that the required equipment would perform its safety function post LOCA while subjected to potentially higher temperatures.

19

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EDCR 89-408 l

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l 4.5.2 SanarsLinn 1

l Essential redundant instrumentation is categorized as Division SI and SII. The following l

Table describes the designation of these divisions for instrument loops from the field l

instrumentation to the Control Room.

l SI Eaulement SIl Equipment

[

Instrument Rack RK 10AB Instrument Rack RK10AA l

Loop "B" Instruments Loop "A" Instruments-CAD Panel A CAD Panel B i

All non. nuclear safety equipment is suitably isolated from the safety class electrical equipment to ensure no failure of NNS electrical equipment can disable both trains of the power or controls to the SFPCS.

All circuitry and installatten of equipment resulting from modifications described in this EDCR maintain the SI and SII separation criteria at Vermont Yankee Separation between trains of safe shutdown equipment is maintained by supplying redundant components from opposite division sources (ie. "A" components P19-2A, V19 220. and V70 257A from SII, and i

"B" components P19-2B, V19 221, V10-257B from SI). Cabb routing further ensures separation of redundant equipment, as "A" component cable is routed in SII raceway, and "B" component cable is routed in SI raceway. The raceway is run from opposite sides and/or :

dif1'erent elevations of the Reactor Building up to the Fuel Pool Cooling and Demineralizer cubicle. This ensures at least one train remains operable under design basis accident

[

conditions. A common mode event, such as a fire, cannot disable the decay heat removal capability of both trains of equipment.

l 4.5.3 Fire Havard Analvnia and EFeet on Rare Shut Down Eaulement The additional fire load, consisting primarily of cable insulation and plastic instrument parts adds a negligible amount of combustible material to the affected fire area. The change-i does not impact the conclusions described in the VY Fire Hazards Analysis. The fire protection review, Reference (e), documents the fire separation adequacy of the components L

within the FPCDS cubicle.

Remote control of SFPCS is provided for plant accident conditions where the Reactor Building may be inaccessible. However, remote control of the SFPCS may be suspended due to a fire in the Control Room or Cable Vault or a Control Room evacuation. Under these conditions either the NFPCS or the SFPCS can be manually placed in operation from the Reactor Building.

The effects on safe shutdown equipment have been analyzed by a review of bus and diesel generator loading. This impact is negligible since the new pump loads are essentially the same as the existing fuel pool cooling pump loads, and only one (SFPCS or NFPCS) load is operated at any time.

f l

1 EDCR 89-408 4.5.4 Structurni Dentan con-idsIAllQAA l

The impact on the ability of the existing SWS and FPCS to withstand a design basis seismic 1

event with the SFPCS piping connected to them has been reviewed. The affected portions of l

these systems (pipe and supports) were seismically analyzed and it was concluded that the l

additio'n of the SFPCS piping will not have an adverse impact. Refer to Enclosure C for the list of calculations.

The Reactor Building floor slab where the new equipment is Ining installed has been-reviewed to ensure that the s' abs integrity is not impacted. The review was performed as calculation VYC-890 and concluded that the installation of the heat exchangers and pumps will not cause the floor to exceed allowable limits during a design basis seismic event.

i The block wall being partially removed and rebuilt to permit installation of the heat exchangers and pumps has been reanalysed. The "new" blockwall has been seismically designed to withstand a design basis seismic event. The shielding ability of the i

reconstructed wall has been reviewed and is acceptable (Reference (nn)).

ALARA design considerations have been incorporated into the design of the new piping. The piping contains a blind flange to allow cleaning of some fuel pool lines. Temporary shielding (portable water sheilds recommended) should be installed, per appropriate procedures, around the existing heat exchangers to reduce the radiation dose that the workers will receive during the installation of this design change.

4.6 Oneratinnal and Maintenance considerations The long term maintenance of the SFPCS heat exchangers is not any different from the

(

maintenance used on other tube and shell design heat exchangers at Vermont Yankee.

In order to ensure the heat exchangers sustained no damage during shipping and to provide a baseline for future testing, an eddy current examination of the tubes in accordance with plant eddy current testing procedures shall be performed on receipt.

1 Procedure, O.P. 2184 " Fuel Pool Cooling System" will require revision to incorporate' all operational aspects of the new SFPCS including the operation of the new seismically qualified make up espability available for emergency use only.

1 Procedure ON 3157 will require revision to include the new qualified pool make up capability l

from the SFPCS.

The system shall be functionally tested at least once a cycle to ensure that the entire SFPCS functions as designed. The system shall be able to pass 700 gpm of fuel pool water through each train of the system and 700 gpm of SW through each heat exchanger.

The MOV's shall be tested in accordance with Vermont Yankees surveillance program for Limitorque operators (0.P. 5220) and MOVATS Testing of MOVs (0.P.5219).

l The pumps shall be tested in accordance with ASME XI IWP 3000 " Inservice Test l

Procedures".

l l

21

l RDCR 89408 Essential instrumentation has been specified to be similar to other existing safety related instrumentation installed in the plant. Based on experience, this instrumentation has been F

proven to be reliable and to meet accuracy requirements. In addition, plant personnel has J

familiarity with it regarding operation and calibration.

Plant programs such as ISI, IST, check valves, MOV's, and heat exchanger testing will need to be reviewed to determine if they require updating, j

Other plant procedures may require revision as a result of this design change. The plant is responsible to identify any other procedures that are effected and to ensure they are updated as necessary, i

5.0 Renourcea i

5,1 Organtration(n) Rnanonsible for the Dalmilnd Denlan The following personnel at Yankee Nuclear Services Division are responsible for the design change; Cognizant Group / Engineer (CE):

Systems - Chris Hansen.

Supporting GroupsfEngineers:

Mechanical - Cliff Greeno Electrical Ralph Moschella I&C - Roger Vibert Plant Cognizant Engineer (PCE) -

Mark A. Stello 1

Implementing Cognizant Engineer (ICE) -

James F. Calchera 5.2 Coat Entima+#a Tho estimated cost to complete this enhancement is approximately $ 3,000,000. A cost breakdown into construction, material, and engineering is not feasible at this time since vendor bids have not been reviewed, construction walkdown for final cost estimate has not been completed and final quantities of materials are not defined.

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EDCR 89-408 g

5.3 haula -

Procurement:

The expected lead time for the pumps, valves, and heat exchangers is 52 weeks abr receipt of order by vendor. Due to the long lead time the specineations for the equipment have been

(

forwarded to Vermont Yankee in advance of the EDCR for early purchan requests.

(

Installation:

The installation of this design change can begin abr materials have arrived on site and the i

Installation and Test Procedures have been prepared. The installation of the pipe, electrical and instrument cables and major components can be performed during plant operation. The 4

actual connection to process systems, power supplies, control room panels, etc., should be completed during the refueling outage. The installation date is by the end the 1993 refueling outage. At this time the system must be fully functional and determined to be so by functional test.

6.0 Onen It aa t

i a

1) The actual details (drawings, vendor maintenance manuals, instrument accuracy, Qualincation Documentation Review Packages (QDR's), test requirements, etc.) of the major equipment (heat exchanger, pumps, instrumentation, and valves) will be provided by a

ECN abr selection of a vendor.

2) Revised QDR's, as required by references (rr) and (ss), will be provided by ECN prior to system operation.

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ll ENCLOSURE "A" SAFE 1Y EVALUATION EDCR 89 408 SECTION PAGE GENERAL

System Description

2 Single Active Failure Capability 3

1. Electrical

2. Mechanical

3. Conclusion SFPCS Impact on Existing Plant 4

1. Decay heat

2. Service Water flow l

3. Electrical demand

4. Structural load

5. Conclusion DETAILED SAFE 1Y EVALUATION 6

1. Accidents 6

1.1 Probability 1.2 Consequences

2. Malfuncuon of equipment -

7 2.1 Probability 2.2 Consequences

3. Possibility of different types of accidents and malfuncuons 8

3.1 Different accident.

3.2 Different malfunction

4. Margin of safety 9

l

5. Conclusions 9

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r GENERAL j

ne following sections evaluate the effect of adding the Standby Fuel Pool Coohng Subsystem (SFpCS) to the existing Vermont Yankee Spent Fuel pool Cooling System. %ts evaluation rMewed all nonnal, abnonnat and accident FSAR conditions, but focuses on Design Basis Accident (DBA) a conditions since the effects and initigation of diese accidents encornpass normal and abnormal plant conditions.

SYSTEM DESCRIFFION:

This EDCR pmvides for the addition of an Standby Fuel pool Cooling Subsystem (SITCS) to the Fuel pool Cooling and Dernineralizer System (FICDSI, thus increasing the capabilities of the Vennont Yankee Nuclear Plant to mitigate abnormal spent fuel pool heat load conditions. %In prwides aumelent heat removal capacity to preclude any impact on plant operation due to insumcierit spent fuel pool cooling.

De Vermont Yankee Fuel Pool Cooling and Deminerahrer System (FICDS) will consist of the Nonnal Fuel pool Cooling System (NFICS) and the Standby Fuel pool Cooling Subsystem (SFICS) he NFpCS provides pool water Altration and deminermhzation to maintain proper pool water clarity and cleanliness for refueling operations, and also numelent pool coohng to rnatntain pool temperatures within spectiled harJts during normal refuehngs and plant operations.

He SFTCS will be capable of prwiding pool coohng should an abnonnally high decay heat load or a design basis seismic event occur that might disable the NFICS. prior to installation of the SPICS.

the Residual lleat Removal System (RHR) in the augmented fuel pool cooling mode prwided this function. The Ri!R system augmented fuel pool cooling mode will still be available following the addition of the SFpCS: however, it will not be required for any nonnal or design basis scenerlos of fuel pool cooling.

l Re SFpCS is a two train. Setsmic Class I. Safety Class 3. Safety Class Electrical powered system designed to prevent a single active failure from disabling the design function of both trains. It is designed as a standby system that can normally be placed in operation remotely from the control room. We SFpCS cools the fuel storage paol by transferring spent fuel decay heat to the Servlee Water system. De pumps cirrulate the pool water in a closed loop, taking suction from the spent fuel storage pool through the heat exchangers and discharging it back into the fuel pool.

Each train of the SFICS (one pump and one heat exchanger) is designed to ensure that pool water temperature does not exceed the technical speelficauon limit of 1500F after a normal (1/3 core) refueling discharge. Utilizing both trains of the SFPCS prwides a decay heat removal capability for a full core discharge without exceeding the pool water technical spectAcation hmit of 150 0F.

We SFICS la qualtfled for design basis accident environments and, in place of the NITCS. will be added to the Vennont Yankee Environmental Quahlication (EQ) Ihogram.

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SINGLE ACTIVE FAILURE CAPARILITT 11 %e SFICS is designed to be single acuve failure proof electrically as detailed in Section 4.

Electrical control and power for the two pumps and four Motor Operated Valves (MOVs) and essential instmmentation is prm1ded by two separate and redundant safety related sources. Essential electrical components and cables are physicaDy separated to ehrninate the potential for common mode fa0ure of both tratne caused by a sing)e event such as a Are, missile, flood, selsmic internetton or electrical short, such that the abthty of the system to remove decay heat is compromised.

Essential electrical components are environmentaDy quallhed for normal and accident emironments to ensure that both trains of electrical and control equipment will be available for its safety related function after exposure to the design basis accident environment detailed in the Vermont Yankee EQ program.

%c addluon of new electrical equipment and cable to the plant in accordance with codes and standards and plant pmeedures ensures the plant will remain within its existing design bases, and that there witi be no adverse effect on the design bases of existing equipment.

21 %e SFICS to designed to be single active failure proof mechanleally as detailed in Section 4.

Att essential mechantral equipment including pumps, valves, heat exchangers, instrumentation, piping and pipe supports are designed and installed to ensure SPICS operation after a design basis seismic event.

InotatLon of the seismically quahfled SFICS from non seismic portions of the NFICS la provided by two MOVs in series on the supply side and by two check valves on the retum side of the NFICS.

Automatic isolauon of the NFICS occurs when pool level falle below a specifled level, thereby assuring that no fatture in NFICS can preclude continued fuel pool cooling via the SFICS. All equipment and controls for this automatic isolation are physicany separated and seismicany and environmentally qualif)ed and installed.

All cesential mechanical equipment is installed to protect against the effee.ts of internaDy generated missiles and seismic interaction from other equipment.

periodic functional and performance tests in accordance with plant procedures ensure essential equipment will meet design function and performance requirements.

%e addition of the new SFICS mechanical equipment to the plant in accordance with regulatory requirements, applicable fabrication codes and plant procedures ensures the plant will remain within its existing design bases, and that this addition will not have an adverse effect on the design bases of existing equipment.

3) Conclusion

%erefore the SFitS design meets single active failure criteria providing a basis for the following 10CFR50.59 evaluation which concludes that this modincation does not Irwolve an unresiewed safety question.

n 3

3

ff i

SFPCS DiPACT ON EXISTLIG PLANT:

1) Decay heat load, from addition al stored spent fuel assemblien, transfer:ed to the Senice Water System (SWS),

in accordance with A:nendment 10 i to Vermont Yankee's Operating 1.leense, the addition of de SPICS is in support of the use of h gh density storage racks which will increase the opent fuel pool storage capnetty from 2000 to 2870 fuel assembhes. %e additional 870 spent fuel assemblies will add about a 12% increase in deesy heat (reference (s)) over the life of the plant.

%e NFPCS transfers decay heat to the e'osed loop Resetor Building Component Cooling Water (RDCCW) system which in turn transfe. Ae heat to the Service Water System (SWS).

Since the SFICS transfers decay heat direedy to the SWS and the SWS is a once through system, any increased decay heat transferred to SWS le dincharged direedy to the ultimate heat sink (Connecticut River) and does not effect other plant equipment or systems.

2) Servlee Water System (SWS) flow demand to the SFPCS.

%e SFICS transfers fuel pool decay heat direc0y to the SWS and can be remotely placed into senice from the control room. When the SFPCS la placed in service to remove fuel pool heat, the hTICS is removed imm service reducing the heat load through the RBCCW heat exchangers to the SWS. We SWS f)ow to the SFICS heat exchanger is an increased flow demand to the SWS, because SWS flow may not be reduced to the RBCCW heat exchangers.

Wie increased SWS flow demand has been evaluated in reference (0 to determine its impact on SWS performance. Wla evaluation provided a detailed review of the addition of the SFICS and its' effect on the SWS during nortnal operation and acendent conditions. SWS performance using the SFICS during accident conditions bounds the conditions for other plant conditions plus any long term effects on the SWS.

i We safety evaluation reviewed and answered each of the questions identified in 10CFR50.59 (a) (2) to determine the safety impact on the SWS of the change relative to new or previously analyzed accidents and relative to the bases for the Technical Specifications.

• mia anktv evahintinn demanatrated that the additlan of the SFPCS. relattve to the r,tv ronctinn of the SWS. will not involve an unnvic-ci ambtv numatlan a

3) Electrical demand impact on existing plant systems and components impodant to plant safety due to the SITCS.

%e effects of safe shutdown equipment have been analysed by a review of electrical bus and diesel generatorloading (VYC 791 & VYC 836). %e SFPCS pump motors require essentially the same electricalload as the NFPCS pump motors (VYC 891). Any increased demand on this electrical supply system is negligible since, prior to placing the SPTCS into service, the NFPCS is removed from service.

We additional electrical demand for the SFICS MOV's is also negligible due to the extremely small str.e of the motors (VYC 894) compared to the electrical bus and diesel generator margins.

@ - 70 9 fr 1

4

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i

4) Structuralloading impact on the existing plant due to the SFPCS.

I

'Ihe Reactor Butiding floor slab where the SFPCS le to be installed has been analyzed M'C 890) to I

ensure that the addition of the SFICS equipment and new portions of the block wall will not increase l

the floor loading atme code allowables during a design basis selsmic event. The existing non.

seismic portion of the block wall providing radiation shlending will be renmed and replaced with a

{

seismically quallned wall M'C 893) to ensure shielding requirernents are maintained and the wall remains intact during a seistnic event.

Design and installation criterna (EDCR sections 4.3 & 4.5) for p! ping and their supports ensure that the piping will remain in its position, perform lis function and have no effect on the surrounding plant equipinent.

Design and installation criteria (EDCH sections 4.3 & 4.5) for power and control cable ensure that existing cabletrays and conduit will not be effected by the new SFPCS cable.

5) Conclusion

'therefore the SFPCS design has no adverse trirpact on the existing plant alngle active failure criteria or accident initigation capabilities and provides a basis for the following 10CFR50.59 review which concludes that this modilleation does not involve an unreviewed safety question.

WYW A

C 5

i

DETAILED RAFETY EVALLIATION

%e previous secuon supplkd the bases for the effect of the SFTCS on plant design basis. %e following Safety Evaluation documents that the addluon of the SMCS will not create an unreviewed safety question as defined in 10CFR50.50.

1.0 ACCIDENTS 1.1 increame in the Prnh.httity of Occurence of an Accident Previnnialv Evahinted in the FSAR

%e Design Basis Accidents outhned in Sect 6on 14.6 of the FSAR are:

- Control Had Drop Accident tms-of Coolant Accident Hefueling Accident

- Main Steam tane B:eak

%e FICDS SFPCS plays no role in the inttlation or direct mitigauon of any of these accidents. Thamrare. the a a or ka RFFOR to the PW daea amt !=

--- the n

• • 6*w af ;;;ss. an af anw of t'-- -- '4==ta-1.2 increase in the Cnnmannienten of an Accident IWminn Evahnated in the FSAR e

%e addition of the SFPCS will prmide addluonal spent fuel pool cooling capability beyond that of the NFPCS and also provide cooling capability after a. seismic event. Adding the SFPCS to the Vermont Yankee EQ program in place of the NMCS maintains a qualilled fuel pool cooling capability. As a result, Fuel Pool Cooling is always available during design basis accidents and abnormal conditions.

%e design and installation of the SFPCS, in accordance with EDCR 89-408 ensures that no electrical or mechanical single acuve failure will disable both trains of the SFPCS or cause a failure of a system essential to plant safety. Die ensures no impset on the radiological elTects of the design baans accidents identified in the FSAR.

Th.

th..e la na !---- --- in tha s --==- - - af an - m " =t

'; !w aval== tad la the pa AB p-neamm _^ 22 the -"seta= af the A*= " -- Penal "M Caanaan _- '

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n8ALFUNCTION OF IBQUIPMir.NT 2.1 Increame in the Prnhahilitv of Occurrence of a u=1r nctinn of Enuinment Imnnrtant to Safety u

Previouahr Evaluated by the FSAR Probabilities of occurance of malfunctions are not specifically analyzed in the FSAR but are encompassed by a single active failure afk eting many safety related pieces of equipment, such as the acttve failure of a Diesel Geneistor.

%e design and installation of the SFICS e naures that no electrical or mechanical single active faGute will disable both trah,* of est ential SFTCS equipment or cause a fadure of equipment in systems essential to plan.'s afety, he SFTCS equipment is installed utuising 5 yelcal separauon to ensure a single event such 5

as fire, electrical short or a seismic event will i ot cause the loss of essential equipment in both trains of the SFICS such that the moulty o.'the system to remove decay heat is compromised or cause the loss of equipment in bth trains of a system essential to plant safety, such as ECCS.

harefase, thaea la na le=---- la the u ' '" :- af --

= - af a - '*-- *!= af ac='

=t im _--imat ta==!= ^ F - ~=64.n t__1 the adAt^^-

af *ka RFKi and-

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2.2 increase in the conmenneneen of a Malfunctinn of Enninment Imanetant to Safety Previonshr Evaluated by the 1%I R L

Spectfle malfunctions of equipment are encompassed by a single active failure affecung many safety related pieces of equipment, such as the acuve failure of a Diesel Generator.

During the mitigation of a Design Basis Accident, plant design criterna ensure that a single active faGure of essential plant equipment or support equipment will not render both trains of any essential plant equipment or accident ndtigation equipment inoperable. By maintaining train separation of the SFICS equipment, no failure of the SFTCS equipment will adversely etTect any accident mitigation system. His ensures no impact on the radiological effects of the design basis accidents identified in the FSAR.

%e addluon of the SFPCS subsystem to the NFTCS provides addiuonal capabuity to deal with the consequences of NFPCS equipment failure, thus the consequences of a malfunction are actually reduced by the addition of the SFpCS.

hasafare. *kase la na t==---- in tha r xr----

= af a==

^!a= af eeuH =t t==- ^==* ta==8 ate na m ;- r't af the -"aeaa= af tha &&Ks.

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3.

POSSIBILITY OF DIFFE3 TENT TYPES OF ACCIDB2fTS AND MALFUNCTIONS 3,1 Ibaalh111tv of an Acektent of a DLiferent TVne than any Previoushr Evaluated in the N De addition of the SFPCS wiu prwide additional pool coohng capabilities beyond the existing pool coollng system and provide pool (cohng after a seismic event.

We discussions in the previous sections document that the additson of the SFPCS wiu not alter the basic function of pool tooling but wtB alter the way in which fuel pool cooling can be accomplit.hed. %ls change increases the capabihues of the plant to provide spent fuel pool coohng during abnormal and accident condluona evaluated in the FSAR. Existing connections to the pool are utilized preventing a dtNerent type of pool volume reduction than any previously evaluated. Based on the design, fabrication and installation requirements meeting all regulatory and fabrication code requirements, no additional types of accidents are created by the addition of the SFICS.

panna

,the--

- =a*w af an - "'-" af a 4'-

-

• tur +'- ame nn-ide evaluated la the Pa*" la not c:2 tad as a meanit af tha -"aea er ek. apKg, 3.2 Possibility of a Different Tyne of untrone tion of 5%ninment Imnaetant to hr tv than Previoushr Evaluated in the FSAR We SFPCS has no adverse eNect on equipment important to safety. De SFPCS increases the capabillues of the plant to respond to Design Ranta Accidents concurrent with a seismic event.

%e effect on Service Water System by the addluon of the SFPCS heat exchangers has been evaluated. De SFPCS heat exchangers prwide a barrier between the fuel pool water and service water. We heat exchangers' materials are corrosion resistant pins a dtNerential pressure wt!! be maintained between the fuel pool (sheU) side and service water (tube) side of the heat exchangers. Derefore, the service water pressure will be kept above the fuel pool water pressure to provide additional assurance against any leakage of fuel pool water into service water.

In addition, the service water flow demand due to the SFTCS heat exchangers has been evaluated in reference (f) showing no detnmental aNect on equipment important to safety.

Thesofare. tha

""!tv af a dif"__

.t is af r==t" e!== af annt==.nt !==----tant ta mar.-tv *ka= ane ; - h:=lw evalmatad in the pa** '- not meanit i _ the -"a+4a= af the SFPCS 1

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4 4.0 MARGIN OF SAFETT AS DEFINED IN A BASES FOR ANT TECIDf! CAL SPECIFICATION The SFPCS increanes the margin of safety by providing a new higher capacity Spent Fuel Pool Coohng Subsystem, which leven given the complete loss of the NFPCS) increases assurance of maintaining the spent fuel pool bulk water temperature below 150 0F in accordance with technical specifications.

• Ihe additional Service Water demand needed to cool the SFPCS has been analyzed in Reference (f). InittaDy utider worst case conditions, the $FpCS may slightly reduce the

)

available SWS flow to specifle accident mitigation system support equipment. However, this analysis shows no detrimental effects to the perfonnance of these systems.

Thernfare. the margrim af malmtv. as daRaad in the Tenkaimmi RamelSmatlema. la met l

reduced as a reanit af the additian af the RFM.

il

5.0 CONCLUSION

S The previous section provides a detalkd ieview of the addition of the SFPCS and its effect on the overall plant. Each of the quesuons identifled in 10CFR50.59 (a) (2) have been answered to determine the safety impact of the SFTCS relative to new or previously analyzed accidents and relat.ve tv the bases for the Technical Specifications.

Therefers. tha a-tv evalm.ea

'---- i.

ok=* the - ?-atina will mat lavelve an Unreviewed Safetr Quantiam.

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1 EDCR 89-408 Spent Fuel Pool Cooling System Enhancement i

Enclosure B Attachment A FSAR Section 10.5 " Fuel Pool Cooling System" FBAR Section 10.5, Figure 10.5-1 FSAR Section 10.6 " Service Water System" (not attached)

FSAR Section 10.6 Figure 10.6-1A f

Attachment B FSAR Figures 8.1-5a & 8.1-!5d

" Emergency 480 v Auxilary One Line j

Wiring Diagram" 1

3 4

n Attachment C Environmental Qualification (EQ) Manual changes Attachment D Safety Classification Manual changes I;

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l Attachment A FSAR Section 10.5 " Fuel Pool Cooling System" FSAR Section 10.5, Figure 10.5-1 l

FSAR Section 10.6 " Service Water System" (not attached)

FSAR Section 10.6, Figure 10.6-1A l

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ENCLOSURE O PAGE 2.

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i:il'l VYNPS FUEL POOL COOLING AND DEMINERALI2ER SYSTEM i

TABLE OF CONTENTS

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Page Title Section 10.5-1 Power Generation 0bjective........................

10.5-1 10.5.1 Safety 0bjective..................................

10.5-1 10.5.2 Power Generation Design Bases.....................

10.5-2 10.5.3 Safety Design Basis...............................

10.5-2 10.5.4 Description.......................................

10.5-10 i

10.5.5 Safety Evaluation.................................

10.5-12 10.5.6 Inspection and Testing............................

10.5.7 i

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VYNPS FUEL POOL COOLING AND DDtINERALIZER SYSTEM LIST OF FIGURES Title Figure No.

10.5-1 Fuel Pool Cooling System Fuel Pool Filter Demineralizer System 10.5-2 b

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VYNPS FUELPOOL._000LINGANDDEMINERA$12ERSYSTEM LIST OF TABLES Title Table No._

Fuel Pool Cooling and Demineralizer System - System 10.5.1 Specifications Fuel Decay !! eat - Af ter Normal Refueling or Full Core Discharged 10.5.2 to Pool - Estimated Using SRP 9.1.3 1

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10.5-111 hl4 n-

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t VYNPS 10.5 FUEL POOL COOLING AND DEMINERAL12ER SYSTEM 10.5.1 Power Generation Objective The objective of the Fuel Pool Cooling and Demineraliter System is to remove the decay heat released from the spent fuel elements. The system maintains a specified fuel pool water temperature, purity, water clarity, and water level.

10.5.2 Safety Objective The safety objective of the Fuel Pool Cooling and Demineralizer System is to remove decay heat from the stored fuel and maintain fuel pool water temperature at a level which will help maintain the Reactor Building environment within the bounding limits of the environmental qualification of electrical equipment.

10.5.3 Power Generation Design Bases 1.

The Fuel Pool Cooling and Demineralizer System shall minimize corrosion product buildup within the spent fuel pool and shall maintain proper water clarity, so that the fuel assemblies can be efficiently handled underwater.

2.

The Fuel Pool Cooling and Demineralizer System shall minimize fission product concentration in the spent fuel pool water, thereby minimizing the radioactivity which could be released from the pool to the Reactor Building environment.

3.

The Fuel Pool Cooling and Demineralizer System shall monitor fuel pool water level and maintain a water level above the fuel sufficient to provide shielding for normal building occupancy.

4.

The Fuel Pool Cooling System shall be capable of maintaining the spent fuel pool temperature below 150 F.

EDCR $7 -yv8 ECN /

(b ENCLOSURE 10.5-1 PAGE

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1 VYNPS 10.5.4 Safety Design Basis The Fuel Pool Cooling and Demineralizer System shall be designed to remove the decay heat from the fuel assemblies and maintain fuel pool water temperature at a level which will help maintain the Reactor Building environment within the bounding limits of the environmental qualification of electrical equipment.

10.5.5 Description General.

The Fuel Pool Cooling and Demineralizer System (FPCDS) consists of four heat exchangers, four pumps, two demineralizers, piping and sufficient valves for control of the design functions and required isolation capability. The Fuel Pool Cooling and Demineralizer System pumps and heat exchangers are located in the Reactor Building below the bottom elevation of the fuel pool.

The fuel pool concrete structure, metal liner, spent fuel storage racks, and the Standby Fuel Pool Cooling Subsystem of the FPCDS are designed to withstand Seismic Class I earthquake loads.

The FPCDS equipment is arranged in such a way as to provide a system with two independent means of cooling the spent fuel pool.

l Normal spent fuel pool cooling and cleanup is provided by using the Nortnal Fuel Pool Cooling Subsystem. This subsystem consist of Pumps P-9-1A and IB and Heat Exchangers E-19-1A and 1B which are arranged in two parallel trains with one train normally lined up and operating during plant operation.

This subsystem of the FPCDS is used to provide pool water filtration and j

demineralization to maintain proper pool water clarity and cleanliness for refueling operations. The Normal Fuel Pool Cooling Subsystem also provides j

sufficient pool cooling to maintain pool temperatures within specified limits during normal refuelings (nominal one-third core discharge) and plant operations.

EDCR 87-V#8 ECN. /

ENCLOSURE O 10.5-2 PAGE 7

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VYNPS 4

llowever, should an unusually high spent fuel decay heat load be placed in the pool, or a seismic event, occur, the Standby Fuel Fool Cooling Subsystem can be utilized to maintain pool temperatures within specified limits. The Standby Fuel Pool Cooling Subsystem of the FPCDS consists of Pumps P-19-2A and 2B ar.d Heat Exchangers E-19-2A and 2B which are normally lined up as two parallel Each trains in a standby mode to the Normal Fuel Pool Cooling Subsystem.

train of the Standby Fuel Pool Cooling Subsystem can be placed in service remotely.

Calculations of expected decay heat loads from normal refuelings and from a f

full core discharge both with previous cycles of spent fuel in the racks were performed in accordance with the guidance provided.in NRC Standard Review Plan 9.1.3, Revision 1, dated July 1981. The normal discharges were assumed discharged to the pool at six days and ten days following shutdown f rom normal operation. The full core discharge was assumed discharged to the poo) ten days following shutdown from normal operation for refueling.

Six days d

following shutdown for a normal refueling is derived from the guidance provided in NRC Standard Review Plan 9.1.3.

Ten days following shutdown for a normal refueling or a full core discharge is the earliest time at which the refueling cavity gates could be replaced isolating the reactor vessel from the b

spent fuel pool. The transfer of the spent fuel assemblies from the reactor vessel to the spent fuel pool is assumed to occur instantly at the six-day or ten-day time period providing a conservative fuel decay heat load in the spent fuel pool. Data from these analyses are provided in Table 10.5.2.

I Examination of this data 6 hows that while the Normal Fuel Pool Cooling Subsystem heat exchanger capacity may be exceeded for relative y short spent fuel decay times, the backup capability of the Standby Fuel Pool Cooling Subsystem of the FPCDS is more than sufficient and can be placed in service l., -

until the fuel decay heac load is reduced.

V l"

The operating temperature of the fuel pool is permitted to rise up to 25 F M

above the administrative temperature limit (125 F) when the circulation flow is temporarily interrupted or when larger than normal batches of spent fuel 1..

f j

are placed in the pool.

EDCR %*YO$

ECN l ENCLOSURE 6 PAGE @

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7 VYNPS Standby Fuel Pool Cooling Subsystem The Standby Fuel Pool Cooling Subsystem (SFPCS) of the FPCDS is shown in Figure 10.5-1.

The Standby Fuel Fool Cooling Subsystem c

he FPCDS is a two train Seismic Class I, Safety Class 3 System designed to prevent a single active failure from disabling both trains.

It is designed as a standby system that can remotely be placed in operation from the Control Room. This portion of the system cools the fuel storage pool by transferring the spent fuel decay heat (see Table 10.5.2) to the Service Water System. The pumps circulate the pool water in a closed loop, taking suction from the spent fuel storage pool through the heat exchangers and discharging it back into the fuel pool.

The standby heat exchangers are of the shell and tube design, with all parts in contact with the pool water being corrosion resistant material. These heat exchangers are each sized to maintain the fuel pool water temperature below 150 F after a normal refueling. Considering one train (one heat exchanger and one pump), this heat removal capability encompasses the normal maximum heat load from completely filling the pool with 2.870 spent fuel assemblies from the last normal discharge. The combined heat removal capability considering both trains (two heat exchangers and two pumps) operating encompasses a full core discharge heat load completely filling the pool with 2.870 spent fuel assemblies. This provides sufficient heat removal capacity to preclude any impact on plant operation due to insufficient spent fuel pool cooling.

The heat exchangers are cooled by the seismically qualified safety-related Service Water System (SWS). The design of the system places the heat exchangers on the suction side of the ptunps.

In order to protect against fuel pool water leakage into the Service Water System, a positive dif ferential pressure is maintained. The fuel pool water side of the heat exchangers has a maximum operating pressure equivalent to the static pressure head from the pool surface to the heat exchanger. The Service Water System side of the heat exchangers has a minimum operating pressure which is greater than the maximum EDCR f/-@@

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ENCLOSURE 6 PAGE 7 OF i

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VYNPS pressure on the fuel pool side of the heat exchangers. By providing a positive differential pressure under all conditions of Service Water System operation, leakage of fuel pool water to the environment is prevented. The differential pressure across each heat exchanger is monitored by a differential pressure indicator in the Control Room.

The Standby t'uel Pool Cooling Subsystem of the FPCDS includes two centrifugal pumps each with a design flow of 700 gpm. All the parts of the pump in contact with the pool water are corrosion-resistant material.

The pumps are Seismic Class I and environmentally qualified to ensure operability after exposure to a harsh environment. The pumps are located within the FPCDS cubicle in such a manner to prevent common mode failure from fire, flooding, or missiles. A low discharte Pressure alarm indicates in the Control Room, plus, the pumps are automatically tripped on a low suction pressure i

condition. One pump alone is designed to provide suf ficient flow for the maximum normal heat load from a normal refueling discharge.

For an abnormal heat load, such as full cote discharge, two pumps can be running concurrently (one in each train) (reference Table 10.5-1).

Four Motor-Operated Valves (MOVs) provide isolation from the nonseismic Normal Fuel Pool Cooling Subsystem and isolation and throttling of the servicc V:tc.

through the heat exchangers.

Each heat exchanger service water outlet MOV is

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powered by the same electrical source as its respective Standby Subsystem pump. These two MOVs V-70-257A and 257B are throttling-type valves providing service water flow control through its respective heat exchanger, and thereby controlling both pool temperature and service water to fuel pool cooling differential pressure.

The two Normal Fuel Pool Cooling Subsystem Isolation Valves V-19-220 and 221 4

are nonthrottling MOVs each powered by a different safety-related electrical power supply. These isolation valves receive a signal to close on low pool level, providing automatic pool isolation from the Normal Fuel Pool Cooling Subsystem in case of a line break in this nonseismic portion of the FPCDS.

In conjunction with the two Normal Fuel Pool Cooling Subsystem isolation MOVs in I

the supply line, there are two discharge line check valves.

These Check EDQ M N ECN /

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VYNPS Valves V-19-18 and V-19-224 provide isolation of the nonseismie Normal Fuel Pool Cooling Subsystem from the Standby Subsystem seismic portion of the system. Thus, isolation of the nonseismic portions of the Normal Fuel Pool Cooling Subsystem is assured.

Piping associated with the Service Water supply and discharEe to the heat exchangers and the fuel pool water piping will be of corrosion resistant material. The piping is designed and constructed in accordance with the requirements of ANSI B31.1-77.

Valves in the fuel pool uater piping are chosen considering their propensity not to collect corrosion products, pressure tight sealing capability, and ease of maintenance.

Indication is provided in the Control Room and/or 10,..

near tte equipment.

Control Room indication for each train includes direct pun temperature, fuel pool water temperature out of the heat exchangers, pump run lights pump discharge pressures, service water flow, SWS to ESS heat exchanger DP and valve position lights.

Local indication includes fuel pool water temperature into the heat exchangers, pump suction and discharge pressures, and heat exchanger DP.

Pool tempe ature is provided by redundant thermocouples located within the pool.

Pool level is provided by redundant transmitters located near the pool. All other transmitters and sensors are located in or near the l

Fuel lool Cooling System cubicle.

Control for the two pumps and four MOVs is provided in the Control Room.

Control Room controls include pump on/off switches, service water throttle valves control switches, and Normal Fuel Pool Cooling Subsystem isolation valves control switches. These remote controls and instrumentation are provided to detect and control pump operation, pool temperature, and system flow, thereby ensuring remote operability of the Standby Fuel Pool Cooling Subsystem of the FPCDS, should the Reactor Building be inaccessible.

Normal Fuel Pool Cooling Subsystem The Normal Fuel Pool Cooling Subsystem (NFPCS) is shown in Figure 10.5-1.

The system cools the fuel storage pool by transferring the spent fuel decay heat (see Table 10.5.2) through heat exchanger (s) to the Re ^%r *uiWS h4 EOCR81'fo8 ECN l 10.5-6 ENCLOSURE O PAGE //

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VYNPS Cooling Water System. Water purity and clarity in the storage pool, reactor well, and dryer-separator storage pit are maintained by filtering and demineralizing the pool water through filter-demineralizer(s), which is shown in Figure 10.5-2.

The system consists of two circulating pumps connected in parallel, two heat exchangers, two filter-demineralizers, and the required piping, valves and instrumentation. Each pump has a design capacity equal to a filter-demineralizer design flow rate (450 gpm) and is capable of simultaneous operation. Two filter-demineralizers are provided. The pumps circulate the pool water in a closed loop, taking suction f rom the spent fuel storage pool, circulating the water through the heat exchanger (s) and filter demineralizer(s), and returning it to the fuel pool and reactor well.

The fuel pool filter demineralizers are located in the Radwaste Building.

The pools (dryer-separator storage pit, reactor well, and fuel storage pool) are filled from the Condensate Transfer System. Make-up to the pools is supplied by the Condensate Transfer System or the Demineralized Water System.

Water is removed from the pools via the fuel pool pumps through the fuel pool filter-demineralizer units to the condensate storage tank.

Fuel pool water is continuously recirculated except during the temporary periods when the reactor well and dryer-separator pit are being drained.

The Normal Fuel Pool Cooling Subsystem is capable of removing the decay heat load of the normal discharge batch of spent fuel with sufficient decay heat reduction. The Standby Fuel Pool Cooling Subsystem can be used in lieu of the Normal Fuel Pool Cooling Subsystem to increase pool cooling in the event that a larger than normal amount of fuel is discharged into the pool or the normal fuel pool cooling heat transfer capacity is exceeded. During refueling, when the reactor well is flooded and the gates between the well and the pool are removed, the RHR System is also available to cool the fuel pool in concert with reactor vessel core cooling.

The RHR System has more than enough capacity to cool the reactor vessel core plus the entire inventory in the spent fuel pool.

EDCR N-V09 ECN /

ENCLOSURE 6 10.5-7 PAGE lA OF t:v.~rtptm-~;::.ny;wn~~r -^ mmg: v:q~ ~~~~::: gr~y:!,2::;gqu q:

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• VYNPS Two small skiminer pumps are provided which take suction f rom the top of the pool to remove surface debris. These pumps pump this water through cartridge filters then back to the pool through the service boxes located around the pools.

Pool water clarity and purity are maintained by a combination of filtering and ion exchange processes. The filter-demineralizer maintains total heavy element content (Cu, Ni, Fe, Hg, etc. ) at 0.1 ppm or less, with a pH range of 5.8 to 8.0 for compatibility with the fuel racks and other equipment.

Particulate material is removed from the circulated water by the pressure precoat filter-demineralizer unit in which a finely divided disposable filter medium is supported on permanent filter elements.

The filter medium is replaced when the pressure drop is excessive or the ion exchange resin is depleted.

Backwashing and precoating operations are manually controlled from the Radwaste Building.

The spent filter medium is flushed from the elements and transferred to the condensate phase separator tanks by backwashing with air and condensate.

The new f11ter medivo is mixed in a precoat tank and transferred as a slurry by a precoat pump to the filter where the solids deposit on the f11ter elements. The holding pump maintains circulation through the filter in the interval between the precoatu.;; operation and the return to normal system operation to hold the precoat on the elements. The pump starts automatically on loss of system flow te maintain sufficient flow through the filter media to retain it on the filter elements.

A post-strainer is provided in the ef fluent stream of the filter-demineralizer to limit the migration of the filter material. The filter holding element is capable of withstanding a differential pressure greater than the developed pump head for the system. The maximum pressure drop across the filter and associated process valves and piping should not exceed 25 psid at the time of filter media replacement. The Backwash System is used to completely remove resins and accumulated sludge from the filter demineralizers with a minimum volume of water. Backwash slurry drains to a phase separator.

The Precoat System is designed to rapidly apply a uniform precoat of filter media to the holding elements of a filter demineralizer. One centrifugal precoat pump and associated piping and valves are provided to precoat either EDCR @/~N ECN [

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VYNPS filter-demineralizer and recirculate the water to the precoat tank or suction side of the precoat pump. The filter-demineralizer units are located separately in shielded rooms.

Each room contains only the filter-demineralizer and piping. All inlet, outlet, recycle, vent, drain, and other valves are located on the outside of one shielding wall of the room, together with necessary piping and headers, instrument elements, and controls. Penetrations through shielding walls are located so as not to compromise radiation shielding requirements.

The fuel pool filter-demineralizers are also used to process liquid l

radioactive vastes. See Chapter 9 of the Vermont Yankee FSAR for details.

l l

The system instrumentation is provided for both automatic and remote manual operations.

Instrumentation and controls are provided to detect, control and record pump operation, pool temperature, and system flow. A pool Leak Detection System has been provided to monitor leakage and thus indicate pool i

integrity.

The pumps can be controlled from the Control Room or at Panel 20-22 in the Radwaste Control Room.

Pump low suction pressure automatically trips the pumps. A pump low discharge pressure alarm indicates in the Radwaste Control Room and a common trouble alarm in the Main Control Room.

The flow rate through each of the filter-demineralizers is indicated by a flow indicator on the Pump Room panel and in the Radwaste Control Room.

The flow indicators can be seen by the operators from the vicinity of the Fuel Pool Cooling System control valves.

A hi h rate of leakage through the refueling bellows assembly, drywell to B

reactor seal, or the fuel pool gates is indicated by lights on the operating floor instrument racks and is alarmed in the Main Control Room.

The filter-demineralizers are controlled from a local panel in the Radwaste Building. Differential pressure and conductivity instrumentation are provided for each filter-demineralizer unit to indicate when '- '--- ' '

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VYNPS Sultable alarms, differential pressure indicators, and flow indicators are provided to monitor the condition of the filter-demineralizers.

l 10.5.6 Safety Evaluation Maximum normal heat load in the pool will be the sum of the heat from all The previous batches plus that just discharged from the current refueling.

Normal Fuel Pool Cooling Subsystem of the Fuel Pool Cooling and Demineralizer System is used normally to maintain the pool water temperature below The Standby Fuel administrative limits during refuelings and plant operation.

Pool Cooling Subsystem is available to provide additional cooling, if needed, the pool temperature does not exceed 150 F.

to ensure that i

Maximum possible heat load would be the sum of the heat from all previous If such a situation arose, batches plus the heat from a full core discharge.

the Standby Fuel Pool Cooling Subsystem would be used to provide the cooling capacity needed under these conditions, or other high heat load conditions, to maintain the pool water temperature less than 150 F.

Also, as an additional means of cooling the spent fuel pool during refueling operations, when the fuel pool and the refueling cavity are connected and filled with water, the Residual Heat Removal (RHR) System can be utilized to provide concurrent cooling to the core and spent fuel pool by circulating the water from the core In this mode, the RHR System will be in to the pool and back to the core.

operation providing cooling to the core and can be shifted to provide concurrent reactor core and spent fuel pool cooling. The RER System has more than enough capacity to cool both the reactor core and the entire inventory of I

stored spent fuel in the spent fuel pool.

The Standby Fuel Pool Cooling Subsystem is designed to provide pool cooling j

under all licensed plant conditions.

This portion of the system is designed as Seismic Class I using the Seismic Class I Service Water System to remove spent fuel decay heat to the ultimate heat sink (Connecticut River).

Essential electrical components in this portion of the system are also environmentally qualified to ensure operability under design basis accident In addition, the equipment is located in such a manner as to conditions.

EDCR 89-'/03 ECN /

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VYNPS prevent coninon mode f ailure f rom fire, flooding, or missiles, such that the ability of the system to remove decay heat is not compromised.

Providing sufficient pool cooling and environmental qualification, assures that the spent fuel will be cooled and boiling will not occur in the spent fuel pool.

Therefore, the Reactor Building environment will not be subject to the consequences of a boiling spent fuel pool.

Leakage of potentially radioactive water from the Standby Fuel Pool Cooling Subsystem through the heat exchanger into the Service Water Systeru is prevented by providing a higher service water pressure than the Standby Fuel Pool Cooling Subsystem pressure. This dif ferential pressure ensures that leakage, if any, will go into the pool.

Indication of this differential pressure is provided in the Control Room along with the controls for initiating the emergency standby portion of the system.

Leakage of the potentially radioactive water from the Normal Fuel Pool Cooling Subsystem to the Service Water System is prevented by using an intermediate closed loop cooling system, Reactor Building Closed Cooling Water (RBCCW),

which transfers the heat from the Normal Fuel Pool Cooling Subsystem to the Service Water System. This Closed Loop System arrangement ensures that fuel pool water leakage, if any, is contained within the RBCCW System and not j

I released into the Service Water System.

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The normal fuel pool cooling flow rate is designed to be larger than that required of two complete water changes per day of the fuel pool, or one change per day of the fuel pool, reactor well, and dryer-separator pit. The Standby i

Fuel Pool Cooling Subsystem flow rate (700 gpm) is approximately 50% greater

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than the normal fuel pool cooling flow rate (450 gpm). The maximum Normal

't Fuel Pool Cooling Subsystem flow rate is twice the flow rate needed to l

maintain the specified water quality.

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An analysis has been made to determine the consequences of dropping a fully loaded spent fuel shipping cask into the fuel storage pool. The results of m

j that analysis showed that the bottom of the pool would lose its water-tight integrity, thereby making it difficult to maintain adequate shielding and e

cooling of the stored spent fuel.. To prevent any load-drop occurrence, the 10.5-11

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Reactor Building crane is designed to be single-failureproof.

(See t

Section 12.2.2.2. of the Vermont Yankee FSAR) 10.5.7 Inspection and Testing No speciai ;:ests are required of the Normal Fuel Pool' Cooling Subsystem because at least one pump, heat exchanger, and f11ter-demineralizer are normally in operation while fuel is stored in the pool.

Redundant units are l

operated periodically to handle abnormal heat loads or to replace a unit for servicing. Y.outine visual inspection of the system components, pumps, heat l

exchangers, instrumentation, and trouble alarms are adequate to verify system i

operability.

The redundant units of the Standby Fuel Pool Coold.ng Subsystem are periodically operated to ensure that the active :omponents of the subsystem can isolate and provide pool cooling by remote retnual initiation. Routine visual inspections of the system components, pumps, heat exchangers, instrumentation, and alarms are adequate to verify system operability.

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SYSTEM SPECIFICATIONS System Specification System Function Normal Fuci Pool Cooling Subsystem Total pool, well, and pit volume 81,500 ft3 Fuel storage pool volume 41,600 ft3 450 gpm System design flow Maximum flow 900 gpm 450 gpm, 225 feet TDH, 25 feet NPSH Pump characteristics llent exchanger - Capacity each 2.23 x 106 Btu / hour, FPC temperature 1250F, RBCCW temperature 1000F, RBCCW flow 350 gpm 267 square f eet, 450 gpm, 25 psi maximum Filter-demineralizer differential pressure (dirty)

Standby Fuel Pool Cooling Subsystem System design flow 70') gpm Maximum flow

'.400 gpm 700 gpm, 130 feet TDH, 20 feet NPSH Pump characteristics Heat exchanger - Capacity each 11,0 x 106 Btu / hour, FPC temperature 1500F, SW temperature 900F, SW flow 700 gpm i

EDCR 89-4'03 ECN /

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1 VYNPS TABLE 10.5.2 I

FUEL DECAY HEAT (ESTIMATED), AFTER OPERATION OF 18 MONTHS NORMAL REFUELING, 136 ASSEMBLIES DISCHARGED FULL CORE DISCHARGE, 368 ASSEMBLIES DISCHARGED Degay Heat l

(10 Btu /hr)

Normal Refueling Discharge Full Core Discharge j

Number of 6 Days 10 Days Nenber of 10 Days Cycle Bundles After After Bundles After Discharged In Pool Shutdown Shutdown In Pool Shutdown l

13 1,586 8.75 7.59 1,818 16.84 14 1,722 9.00 7.79 1,954 17.18 15 1,858 9.18 7.96 2,090 17.37 16 1,994 9.35 8.12 2,226 17.53 17 2,130 9.50 8.28 2,362 17.69 18 2,266 9.65 8.42 2,498 17.84 19 2,402 9.80 8.57 2,634 17.99 20 (1) 2,538 9.94 8.71 2.770 18.13 21 2,674 10.07 8.84 2,906 (2) 18.26 22 2,810 10.20 8.97 N/A 23 2,946 (2) 10.33 9.10 N/A N, 0TE : The decay heat from the previous cycle discharges is included in the above-estimated heat loads.

1.

Loss of full core reserve discharge capability.

2.

Exceeds capacity of reracked fuel pool.

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