ML20236L626

Rev 0 to South Texas Units 1 & 2 Spent Fuel Pool Dilution Analysis

ML20236L626
Person / Time
Site:	South Texas
Issue date:	02/25/1998
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20236L615	List: ML20236L617 ML20236L619 ML20236L626 ML20236L631 NOC-AE-000178, Application for Amends to Licenses NPF-76 & NPF-80, Reflecting Rev of Criticality Analyses & Rack Utilization Schemes for Regions 1 & 2 of Spent Fuel Racks as Described in STP UFSAR & TSs
References
NUDOCS 9807130097
Download: ML20236L626 (46)

Text

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I SOUTH TEXAS UNITS 1 AND 2 SPENT FUEL POOL DlLUTION ANALYSIS Prepared By: 444/'

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Verified B I 4

Rev.0 I

2/25/98 WESTINGHOUSE ELECTRIC COMPANY 9807130097 900707 PDR ADOCK 05000498 P PDR

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l Table of Contents Section Page

1.0 INTRODUCTION

1 2.0 SPENT FUEL POOL AND RELATED SYSTEM FEATURES 2 l 2.1 Spent Fuel Pool 2

, 2.2 Spent Fuel Storage Racks 3

! 2.3 Spent Fuel Pool Cooling Subsystem 3 l.

2.4 Spent Fuel Pool Cleanup Subsystem 4 l 2.5 Dilution Sources 4 2.6 Boration Sources 8 2.7 Spent Fuel Pool Instrumentation 9 l

2.8 Administrative Controls 10 l

2.9 Piping 11 l 2.10 Loss of Offsite Power impact 11 2.0 SPENT FUEL POOL DILUTION EVALUATION 13 3.1 Calculation of Boron Dilution Times and Volumes 13 3.2 Evaluation of Boron Dilution Events 15 3.3 Summary of Dilution Events 19 l

4.0 CONCLUSION

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5.0 REFERENCES

23 FIGURES FIGURE 1 - Spent Fuel Pool and Related Systems 24 FIGURE 2 - Spent Fuel Pool Plan View 25 FIGURE 3 - Spent Fuel Pool Fluid Mixing 25 l

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1.0 INTRODUCTION

A boron dilution analysis has been completed for crediting boron in the South Texas spent fuel rack criticality analysis. The boron dilution analysis includes an evaluation of the following plant specific features:

Dilution Sources Boration Sources Instrumentation Administrative Procedures

- Piping Loss of Offsite Power impact Boron Dilution initiating Events Boron Dilution Times and Volumes The boron dilution analysis was performed to ensure that sufficient time, administrative procedures, and instrumentation are available to detect and mitigate the dilution before the spent fuel rack criticality analysis 0.95 k, design basis is exceeded. The design basis assumes normal plant operations and I

fuel movement. No other accidents such as misloading a fuel assembly are assumed to occur during the dilution accident.

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I 2.0 SPENT FUEL POOL AND RELATED SYSTEM FEATURES 1

This section provides background information on the spent fuel pool and its related systems ana features. A one-line diagram of the spent fuel pool related systems is provided as Figure 1. A spent

fuel pool is provided for each of the two Units at South Texas. For the purposes of this evaluation. the spent fuel pool and its related systems are sufficiently similar between the two Units that they will be treated as identical. Any significant differences will be identified, so that this report will be bounding for both Units.

2.1 Spent Fuel Pool The design purpose of the spent fuel pool is to provide for the safe storage of irradiated fuel assemblies. The pool is filled with borated water. The water removes decay heat, provides shielding for personnel handling the fuel, and provides for removal of a portion of the iodine released during a fuel handling accident. Pool water evaporation takes place on a continuous basis, requiring periodic makeup. The makeup source can be unborated water, since the evaporation process does not carry )

off the boron. Evaporation actually increases the boron concentration in the pool.

The spent fuel pool is a reinforced concrete structure with a stainless steel liner. The water tight liner has dedicated drain lines (channels) to collect and detect liner leakage. The pool structure is designed to meet seismic requirements. The pool is approximately 45 feet deep. The top of the pit is located on l

the 68' elevation of the fuel handling building. The bottom of the pit is at the 21' 11" elevauan.

On the floor elevation there is a 2 to 3' curb surrounding the pool. The curb, in addition to an open floor drain, minimizes any pool dilution source from the floor elevation level.

As shown in Figure 2, a transfer canal lies adjacent to the pool and connects to the reactor refueling water cavity during refueling operations. The pool and the transfer canal are connected by fuel transfer 1

slots that can be closed by pneumatically sealed gates. The transfer canal is normally filled. However, '

l the accidental opening of the gates, if the canal were dry, would lower the water level to 39'-9", %aving approximately 1'-4" of water above the top of the active fuel. The elevation of the top of the gates, when installed, is approximately just below the floor level of the spent fuel pool area. The removable 2

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1 gates are designed to support the full height of water remaining on the pool side after the canal side is completely drained.

l The gates between the pool and the transfer canal are normally removed. However, no credit is taken for the transfer canal water volume. The volume of the poolis approximately 486,000 gallons to the low level alarm elevation of 66'-0". The majority of the water volume displaced by objects in the pool is due to the spent fuel assemblies. The maximum number of assembly locations is 1969. Since it is conservative to ascume all sites are usable, the volume of all 1969 assemblies (57,900 gallons) is subtracted from the total pool volume. The racks themss ves occupy a relatively small volume (6000 gallons), but they are subtracted as well. When the above volumes are subtracted from the pool volume, the remaining water volume is conservatively rounded down to 420,000 gallons at the low level alarm setpoint elevation of 66'-O'.

2.2 Spent Fuel Storage Racks l

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The spent fuel racks are designed to support and protect the spent fuel assemblies under normal and credible accident conditions. Their design ensures the ability to withstand combinations of dead loads, live loads (fuel assemblies), and safe shutdown earthquake loads.

2.3 Spent Fuel Pool Cooling Subsystem The spent fuel pool cooling subsystem is designed to remove the heat generated by stored spent fuel elements from the spent fuel pool. System design does not incorporate redundant active components i

except for the spent fuel pool pump and heat exchanger. System piping is configured so that failure of any pipeline in the cooling system does not drain the spent fuel pool below the top of the stored spent fuel assemblies.

The portion of the spent fuel pool cooling subsystem which, if it failed, could result in a significant release of pool water, is seismically designed.

Each of the two trains of the cooling subsystem consists of a pump, a heat exchanger, valves, piping and instrumentation. The pump takes suction from the fuel pool at an inlet located four feet below the normal pool water level, transfers the pool water through a heat exchanger and returns it back into the 3

pool through a discharge header with sparging nozzles located on the pool bottom and at the opposite wall from the coeling system inlet. The retum line includes an anti siphoning hole at elevation 65'- 6" to limit loss of poolinventory in the event that the retum line breaks below the normal water level. The l heat exchangers are cooled by component cooling water.

2.4 Spent Fuel Pool Cleanup Subsystem The spent fuel pool cleanup subsystem is designed to maintain water clarity and to control borated water chemistry. The cleanup subsystom is connected to the spent fuel pool cooling system. About 250 gpm of the spent fuel pool cooling pump (s) discharge flow can be diverted to the cleanup loop, which includes the spent fuel pool domineralizers and filters. The filters remove particulate from the spent fuel pool water and the spent fuel pool domineralizer removes ionic impurities.

The refueling water purification loop also uses the spent fuel pool domineralizer and filters to clean up the refueling water storage tank after refueling operations. The flow rate in the loop is limited to 200 gpm administratively to accommodate the design flow of the spent fuel pool domineralizer.

The spent fuel pool has a surface skimmer system designed to provide optical clarity by removing surface debris. The system consists of two surface skimmers, a single strainer, a single pump and one filter. The skimmer pump is a centrifugal pump with a 100 gpm capacity. The pump discharge flow passes through the filter to remove particulate. It returns to the spent fuel pool.

2.5 Dilution Sources 2.5.1 Boron Recycle System (BRS)

The BRS connects to the spent fuel pool at two locations. The first connection is a 3' line from the outlet of the spent fuel pool heat exchangers to the BRS recycle holdup tanks. This connection is j normally isolated and is used to transfer water from the spent fuel pool to the BRS recycle holdup tanks. The isolation is by one manual valve.

l There is no check valve between the BRS recycle holdup tanks and this connection to the spent fuel pool cooling system. However, it is not credible that water would back up from the tank to the spent 4

fuel pool cooling system. In the situation where the BRS recycle holdup tank is misaligned to the spent ,

fuel pool through this connection, water from the spent fuel pool cooling system would flow by gravity to the tank due to the elevation difference. Thus, this path would only result in the loss of water from the poolif the normally closed valve were to fail or be left open. The holdup tanks also have a high level alarm, which annunciated in the radwaste control room.

The second connection between the spent fuel pool and the BRS is from the BRS recycle evaporator feed pump discharge header to the cleanup loop piping. This is a I:.,rmally isolated 3' line that is an .

additional source of makeup water to the pool. The maximum rate of addition is approximately 255 gpm, assuming both feed pumps are operating in parallel. Two normally closed and one lock-closed manual valves are used to isolate this connection.

The recycle evaporator feed pumps can take suction from either of the two BRS recycle holdup tanks.

However, by procedure, only one holdup tank is aligned at a time. Manual valve manipulations are required to switch the pump suction to another tank. Each BRS recycle holdup tank has a total volume of approximately 84,000 gallons and can have a boron concentration from 0 to 3000 ppm.

2.5.2 Reactor Makeup Water System The reactor makeup water system (RMWS) includes one reactor makeup water storage tank (RMWST) and two RMW pumps per Unit. During normal operation, one RMW pump is running on recirculation to provide RMW on demand to multiple users. Each RMWST contains approximately 153,050 gallons of non-borated, reactor grade, deionized water. Makeup to the tank is provided automatically from the demineralized water system on a low tank level signal.

The RMWS connects to the spent fuel pool cooling system directly in the retum hesder to the pool.

Uting the direct connection, the contents of the RMWST can be transferred directly to the spent fuel i pool via the RMW pumps. The direct connection is normally isohted from the RMWS by a closed manual valve. The flow rate through this path is estimated to be 245 gpm, assuming both RMW i pumps are operating. The direct connection is used as the normal water supply to the spent fuel pool and is a source of makeup water in case of a loss of spent fuel pool inventory.

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2.5.3 Demineralized Water System The demineralized water system includes a demineralized water storage tank and three transfer pumps. The storage tank capacity is 961,700 gallons. A flow path from the pump discharge header feeds directly into the spent fuel pool cooling system return header to the pool. The maximum flow from the 2"line makeup line to the spent fuel poolis estimated to be 190 gpm, assuming two transfer pumps are operating. An indirect path for demineralized water to the spent fuel poolis through a 1" line to the sluice path for the spent fuel pool and other system domineralizere. The demineralized is isolated from the cleanup loop by one manual valve. If this valve were left open, demineralized water could be transferred into the spent fuel pool. The flow from this pathway is estimated to be 55 gpm, i assuming two transfer pumps are operating.

The demineralized water supply to the upender hydraulic unit in the spent fuel pool area is provided by 3/8" tubing. If this line were to break, its flow rate into the spent fuel pool would be bounded by the flow l available through the larger lines as described above. Therefore, this 3/8" path is not considered i

further in this analysis.

2.5.4 Component Cooling Water System Component cooling water is the cooling medium for the spent fuel pool cooling system heat exchangers. There is no direct connection between the component cooling system and the spent fuel pool cooling system. If, however, a leak were to develop in a heat exchanger that is in service, the connection would be made. The component cooling system normally operates at a pressure very close to that of the spent fuel pool cooling system. Therefore,it is just as likely that a breach in a spent fuel pool cooling system heat exchanger tube would result in non-borated component cooling water entering the spent fuel pool cooling system as spent ftel pool water entering the component cooling water system.

The flow rate of any leakage of component cooling water into the spent fuel pool cooling system would be very low due to the small difference in operating pressures between the two systems. Even if there was significant leakage from the component cooling water system to the spent fuel pool, the impact on the spent fuel pool boron concentration would be limited to the loss of component cooling water surge tank volume that would initiate alarms and control room indications to alert the control room operators.

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A low surge tank level alarm would alert the control room operators of a component cooling water system leak. If this alarm were to fail and leakage from the component cooling water system to the spent fuel pool cooling system were to continue undetected, the component cooling water surge tank would be periodically refilled with water from the demineralized water system. The resulting dilution from the demineralized water system would be bounded by the dilution events discussed in Section 2.5.3.

Because a spent fuel pool heat exchanger leak is bounded by other analyzed events, it is not considered further in this analysis, i 2.5.5 Drain Systems The equipment or floor drain systems connect directly to the spent fuel pool cooling system and skimmer system at the drain connections for the spent fuel pool pumps, heat exchangers (tube side),

filters, domineralizers, demineralized filters, the skimmer pump, and skimmer filter. Each connection has a normally closed isolation valve. Backflow through these paths is not considered credible, because the situation would cause water to back up through floor drains in a number of locations before getting into the spent fuel pool cooling system.

2.5.6 Fire Protection System l

in the case of a loss of spent fuel pool inventory, two local fire hose station are a potential makeup source. These stations are capable of providing a total flow of approximately 350 gpm of non-borated water. Any planned addition of fire system water to the spent fuel pool would be under the control of an approved procedure and the effect of the addition of the non-borated water from the fire system on the spent fuel pool boron concentration would be addressed.

There is a 6" fire protection hose supply piping header located under the hose stations outside the spent fuel pool area. If this line were to break, a significant amount of water would, if not isolated by operator action, be released into the area outside and beneath the spent fuel pool area. The fire protection system contains instrumentation which would alarm in the control room should this type of flow develop in the fire protection system. Thus, the break of any of the fire protection hose supply piping is not considered further in this analysis.

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2.5.7 Spent Fuel Pool Domineralizers The two spent fuel pool demineralizers each have a maximum capacity of 75 ft' of 1:1 equivalent mixed bed resin. This implies a volume ratio of 60%/40% anion to cation resin. If we assume the beds were loaded with 100% anion, it would bound the capacity to remove boron when it is first aligned to the system. Each demineralized would be operated at 250 gpm maximum flow rate. Dilution of the spent fuel pool resulting from operation of the demineralized will not result in a change in the spent fuel pool inventory.

2.5.8 Dilution Source and Flow Rate Summary Based on the evaluation of potential spent fuel pool dilution sources summarized above, the following dilution sources were determined to be capable of providing a significant amount of non-borated water to the spent fuel pool. The potential for these sources to dilute the spent fuel pool boron concentration  !

will be evaluated in Section 3.0.

APPROXIMATE SOURCE FLOW RATE (GPM)

Boron Recycle System

- Holdup Tank to cleanup subsystem 255 Reactor Makeup Water System

- 2' connection to retum header 245 Demineralized Water System

- 2" connection to retum header 190

- 1" makeup to spent resin sluice header 55 Fire Protection System l

- Fire hose stations in spent fuel pool area 350 Spent Fuel Pool Domineralizers 500 i

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2.6 Boration Sources The normal source of borated water to the spent fuel poolis from the refueling water storage tank via j the refueling water purification pump. It is also possible to borate the spent fuel pool by the addition of

! dry boric acid directly to the spent fuel pool water. A discussion of each source follows:

I 2.6.1 Refueling Water Storage Tank The refueling water storage tank (RWST) connects to the spent fuel pool via the purification loop. This connection is used to purify the RWST water when the purification loop is isolated from the spent fuel pool cooling system. Normally, this connection can supply borated water to the spent fuel pool via the refueling water purification pump to the inlet to the spent fuel pool cooling system purification loop. The refueling water purification pump is powered from a non-safeguards bus power supply. It must be re-started manually following a loss of offsite power. The RWST is required by Technical Specifications to be kept at a minimum boron concentration of 2800 ppm.

Alternative!y, the low head safety injection pumps can be utilized to transfer RWST water to the spent fuel pool. Temporary connections are made to the vent and drain lines on one of the three low head safety injection pumps. The hoses are routed to the spent fuel pool, the pump is operated on miniflow, the vent and drain valves are opened, and the water is transferred to the pool.

2.6.2 Direct Addition of Boric Acid if necessary, the boron concentration ct the spent fuel pool can be increased by emptying drums of dry l boric acid directly into the spent fuel pool. However, boric acid dissolves very slowly at room temperature and requires that the spent fuel pool cooling pumps be available for mixing throughout the spent fuel pool water. (See section 3.1 for further discussion on spent fuel pool mixing.)

2.7 Spent Fuel Pool Instrurnentation instrumentation is available to monitor spent fuel pool water level and temperature. Additional instrumentation is provided to monitor the pressure and flow of the spent fuel pool cleanup system, and I

pressure, flow, and temperature of the spent fuel pool cooling system.

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The instrumentation provided to monitor the temperature of the water in the spent fuel pool is indicated locally and alarms are annunciated in the control room. The water level instrumentation indicates locally and the high and low alarms annunciate locally and in the control room. The instrumentation which monitors radiation levels in the spent fuel pool area provides high radiation alarms locally and in 1

the control room.

J A change of one foot in spent fuel pool level with the transfer canal isolated requires approximately.

10.300 gallons of water. If the pool level was raised from the low level alarm point to the high level Clarm (12"), a dilution of approximately 10,300 gallons could occur before an alarm would be received in the control room, if the spent fuel pool boron concentration were at 2500 ppm initially, a dilution using unborated water would only result in a reduction of the poot boron concentration of approximately 61 ppm.

2.8 Administrative Controla The following administrative controls are in place to control the spent fuel pool boron concentration and water inventory:

1. Procedures are available to aid in the identification and termination of dilution events.

2. The procedures for loss of inventory (other than evaporation) specify that borated makeup )

sources be used as makeup sources. The procedures specify that non-borated sources c#y be I used as a last resort. j i

3. In accordance with procedures, plant personnel perform rounds in the spent fuel pool enclosure i once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The personnel making rounds to the spent fuel pool are trained to be aware of the change in the status of the spent fuel pool. They are instructed to check the ,

, temperature and level in the pool and conditions around the pool during plant rounds.  !

4. The spent fuel pool boron concentration is verified by sample analysis every seven days per Technical Specification 3.9.13.

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Mministrative controls on the spent fuel pool boron concentration and water inventory ensure that the boron concentration is administratively controlled during both normal and accHent situations. The procedures ensure that the proper provisions, precautions and instructions are in place to control the pool boron concentration and water inventory.

2.9 Piping There are no systems (other than those listed in section 2.5.1 to 2.5.8) identified which have piping in the vicinity of the spent fuel pool which could result in a dilution of the spent fuel pool if they were to

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The fire protection, reactor makeup water, and demineralized water line stations, if damaged, could provide a source of spent fuel pool dilution. However, as discussed in Section 3.2, the physical arrangement of tne area surrounding the spent fuel pool would limit the amount of water which could flow into the spent fuel pool.

2.10 Loss of Offsite Power impact Of the dilution sources listed in Section 2.5.8, only the fire protection system is capable of providing non-borated water to the spent fuel pool during a loss of offsite power.

The loss of offsite power would affect the ability to respond to a dilution event. The spent fuel pool level instrumentation is not powered from emergency diesel generator-backed power supplies.

The refueling water purification pump is not powered from a safeguards supply and would not be available to deliver borated water from the RWST. The RWST cannot be gravity-drained to the spent l

fuel pool through the refueling water purification pump, because the spent fuel pool minimum level is above the maximum level of the RWST. The low head safety injection pumps are powered from a safeguards bus, and can be used as an indirect source of RWST water to the spent fuel pool. Finally, manual addition of dry boric acid to the pool could be used if it became necessary to increase the spent fuel pool boron concentration during a loss of offsite power.

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The safeguards power supplies for the spent fuel pool cooling pumps are backed by the diesel generators. However, the pumps are not part of the dieselloading sequence, so they must be manually aligned to the diesels and restarted following a loss of offsite power to assure good mixing in the spent fuel pool.

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l 3.0 SPENT FUEL POOL DILUTION EVALUATION

. 3.1 Calculation of Boron Dilution Times and Volumes For the purposes of evaluating spent fuel pool dilution times and volumes, the total pool volume available for dilution, as described in section 2.1, is conservatively (low) assumed to be 420,000 gallons.

1 Based on the criticality analysis (Reference 1), the soluble boron concentration required to maintain the spent fuel pool boron concentration at k, < 0.95, including uncertainties and bumup, with a 95%

probability at a 95% confidence level (95/95) is 700 ppm. This concentration assumes no fuel misloading accident.

j The spent fuel pool boron concentration is currently maintained above 2500 ppm, at approxima:aly 2800 ppm. For the purposes of calculating dilution times and volumes, the initial spent fuel pool boron concentration is conservatively (low) assumed to be 2500 ppm. The evaluations are based on the spent fuel pool boron concentration being diluted frora 2500 ppm to 700 ppm. To dilute the pool water volume of 420,000 gallons from 2500 ppm to 700 ppm would require 535,000 gallons of non-borated water, based on a feed and-bleed operation (constant volume).

l This analysis assumes thorough mixing of all the non-borated water added to the spent fuel pool with l the contents of the spent fuel pool. Refer to Figure 3. Based on the design flow of 2500 gpm per spent fuel pool cooling pump, the 420,000 gallon system volume is turned over approximately every three hours with one pump running, which is the normal alignment. It is unlikely, with cooling flow and convection from the spent fuel decay heat, that thorough mixing would not occur. However, if mixing was not adequate, it would be conceivable that a localized pocket of non-borated water could form somewhere in the spent fuel pool. This possibility is addressed by the calculation in Reference 1 which shows that the spent fuel rack K, will be less than 1.0 on a 95/95 basis with the spent fuel pool filled with non-borated water. Thus, even if a pocket of non-borated water formed in the spent fuel pool, K.,,

would not exceed 1.0 anywhere in the pool.

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l t_ ______ _ _ _ - _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - -

The time to dilute the spent fuel pool depends on the initial volume nf the pool and the postukted rate of dilution. The dilution volumes and times for the dilution scenarios discussed in Sections 3.2 and 3.3 are calculated based on the following equation:

to = In (C, /C, )V/O (Equation 1)

Where:

C, = the boron concentration of the pool volume at the beginning of the event (2500 ppm)

C = the boron endpoint concentration (700 ppm)

O = dilution rate (gallons / minute)

V = water volume of spent fuel pool (420,000 gallons) t, = time to reach C (minutes)

This equation is derived as follows:

Dilution is via feed and-bleed formula denved from a boron mass balance on the SFP:

O,,, O C, -----> M,, @ C,, ---> O, @ C,,

O,,, = flow of solution entering SFP at boron concentration C, Oy = flow of solution exiting SFP at boron concentration C, M,, = Mass of SFP solution The mass of boron entering or leaving the SPF per unit time is given by:

O x C, x p = flow (ft' solution / min.) x Boron conc. (ib.B/lb. solution) x density (Ib. solution /ft' solution)

= lb.8/ min.

The change in the SFP boron mass per unit time = Boron mass in - Boron mass out:

4 sLM, = OC , p - OC, p ji dt if the equation is divided by the SFP mass, the result is boron concentration. Also, C , = C,,

Change in SFP boron conc / unit time = Boron Conc entering - Boron Conc. leaving s[Q, = OC_ o - OC_ o dt M ,, M,,

O,,, = 0, for a constant volume system, and p is relatively constant for dilute solution.

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l dQ ,= 0o(C_. -C_J d M ,,

i p/M,, = SFP volume, and re-arranging, dQ (C,-k,=) Q,d! V, Integrating C,, from C, to C , and C, = 0 for an unborated (dilution source) solution:

In (0 - C ) - In (0 - C.) = Ot/V In (C,) - In (C ) = Ot/V in {G,1 =.Q1 (C ) V t, = In (C, /C, )V/O 3.2 Evaluation of Boron Dilution Events l The potential spent fuel pool dilution events that could occur are evaluated below:

3.2.1 Dilution From BR8 Recycle Holdup Tanks 1 l The contents of a BRS recycle holdup tank can be transferred via the recycle evaporator feed pumps  ;

to the spent fuel pool via the purification loop piping. The flow path is isolated by one locked closed

! and two normally closed valves. This connection is a designated source of makeup water in a loss of spent fuel pool inventory event. Each of the two BRS recycle holdup tanks has a total volume of approximately 84,000 gallons. The water in the tanks can have a boron concentration from 0 ppm to 3000 ppm, but is more typically in the range of 0 to 1500 ppm, consistent with the reactor coolant system. Therefore, any amount of boron in the BRS recycle holdup tank water would reduce the dilution of the spent fuel pool resulting from the transfer of BRS recycle holdup tank water to the spent fuel pool. To dilute the spent fuel pool from 2500 ppm to 700 ppm would require the addition of 535,000 gallons of unborated water. The combined contents of the two BRS recycle holdup tanks (approximately 168,000 gallons) is less than the required dilution volume.

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The BRS recycle evaporator feed pumps can take suction from either of the two BRS recycle holdup i tanks. Manual valve manipulations are required to switch the pump suction to another tank. However,

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l it is assumed for the purposes of this evaluation that the contents of both BRS recycle holdup tanks are  ;

available for a spent fuel pool dilution event. The 168,000 gallons of water contained in both BRS recycle holdup tanks is less than the 535,000 gallons necessary to dilute the spent fuel pool from 2500 ppm to 700 ppm. The path from the recycle evaporator feed pumps to the spent fuel pool via the 3' connection to the spent fuel pool cooling return header can provide approximately 255 gpm. If the open path were left unattended, it would take 40 minutes to increase the spent fuel pool level from the

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low to high alarm setpoints, and 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> to provide the 635,000 gallons required to dilute the pool from 2500 to 700 ppm boron, if sufficient water were available.

1 3.2.2 Dilution From Reactor Makeup Water Storage Tank i

The reactor makeup water system consists of one reactor makeup water storage tank and two reactor makeup water pumps per Unit. The reactor makeup water storage tank contains approximately 150,000 gallons of non-borated, reactor grade water. However, makeup to the tank from the demineralized water system is automatically provided on a low level signal. Thus, with sufficient makeup, the tank contents could be std %,$ to dilute the spent fuel pool from 2500 to 700 ppm.

The contents of the reactor makeup water storage tank can be transferred via the reactor makeup water pumps to the spent fuel pool via the cooling loop return header. This connection is normally isolated from the reactor makeup water system by a closed manual valve. It can be used as the normal makeup supply to the spent fuel pool and is a source of makeup water in case of a loss of spent fuel poolinventory event.

The path from the reactor makeup water pumps to the spent fuel pool via the 2' connection to the cpent fuel pool cooling retum header can provide approximately 245 gpm. If the open path were left unattended, it would take 42 minutes to increase the spent fuel pool level from the low to high alarm getpoints, and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron, if sufficient makeup water were available.

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3.2.3 Dilution From Domineralland Water system The demineralized water system includes a demineralized water storage tank and three transfer pumps. The non-borated contents of the demineralized water storage tank can be transferred directly to the spent fuel pool. The volume of the demineralized water storage tank (961,000 gallons) is greater than the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

The path from the demineralized water transfer pump to the spent fuel pool cooling retum header via the 2" connection can provide approximately 190 gpm. If the flow path were left unattended, it would take 54 minutes to increase the opent fuel pool levei from the low to high alarm setpoints, and 47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br /> to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

The path from the demineralized water pump to the spent fuel pool via spent resin sluice pump discharge header can provide approximately 55 gpm. If the flow path were left unattended, it would take 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to increase the spent fuel pool level from the low to high alarm setpoints, and 7 days to

~ provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

3.2.4 Dilution from Fire Protection Syr(em l

The fire protection system draws from two 300,000 gallon tanks. The combined volume of the two tanks is greater than the 535,000 gallons required to dilute the spent fuel pool from 2500 to 700 ppm boron. The path from the fire water pump to the two fire hose stations in the spent fuel pit area can provide approximately 350 gpm. If the hoses were placed in the spent fuel pool and left unattended,it l would take 29 minutes to increase the spent fuel pool level from the low to high alarm setpoints, and 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to provide the 535,000 gallons required to dilute the pool from 2500 to 700 ppm boron.

3.2.5 Dilution Resulting From Solemic Events or Random Pipe Breaks l

A seismic event could cause piping ruptures in the vicinity of the spent fuel pool in piping that is not ,

seismically qualified. For a seismic event with offsite power available, rupture of the reactor makeup and demineralized water supply lines to the spent fuel pit cooling loop will not result in a direct addition of unborated water to the spent fuel pool. If offs!!e poweris not available, the reactor makeup and demineralized water systems would not operate and thus, there would be no dilution source.

17

in the event of a break in one of the fire protection hose station supply lines which are outside the spent fuel pool enclosure but in the general area surrounding the spent fuel pool, water would approach the spent fuel pool, but would be blocked by the 2 to 3' curb surrounding the pool. In addition, there is an open stairwell and floor drains through which this water would drain to lower elevations of the fuel handling building. For the purposes of this analysis,it is conservatively assumed that a fire protection hose station line break floods the entire area to a depth of three inches. This is conservative because of the openings to the cask pool, the cask connecting channel, the cask decontamination area, the new fuel storage, and the drop area opening leading to bay doors in the building. Even before the water level reached three inches, the drop area would be capable of draining the full flow of any fire protection hose station supply line break.

Once the water depth was equalized at three inches inside the curb (pool side) and outside curb (floor area), the driving head to force additional water into the cnclosure would be significantly reduced. At ,

that point, most of the flow from the pipe break would i fpass the spent fuel pool enclosure, taking the path of least resistance around the enclosure to the drop area opening. l The total amount of water added to the spent fuel pool enclosure to raise the water level to three inches above the floor would be approximately 23,192 gallons assuming the spent fuel cool level was initially at the low level alarm setpoint. This is much less than the 535,000 gallons required to dilute the spent fuel pool from 2500 ppm to 700 ppm. While a limited amount of flow through the enclosure would continue until the line break were isolated, a fire protection system line break of this magnitude would be readily detected in the control room and break flow would be terminated long before enough water could enter the spent fuel pool enclosure to reduce the pool boron concentration to 700 ppm.

Because of the limited flow into the spent fuel pool enclosure, and because a fire protection hose station supply line break would be terminated long before the spent fuel pool boron concentration would be reduced to 700 ppm, this event is not considered a credible event and is given no furthe' consideration in this analysis, i

During a spent fuel pool walkdown, a fire protection system standpipe and sprinkler manifold piping were identified as a possible dilution source. However, the piping is located 71 feet from the pool. in the unlikely even of a double-ended break of this piping,it is postulated that a maximum flow of 4000 1 18

gpm could spray in the direction of the spent fuel pool. In this case, a prompt fire protection system low-pressure alarm (with indication in the fuel handling building) would be generated, since the system jockey pump cannot maintain such a high flow rate. Operators would start a fire protection pump to restore system pressure and would soon begin an investigation into the source of the leakage. An inspection of the spent fuel pool area should quickly locate the pipe break and the spray into the pool.

Assuming an initial spent fuel pool level at the low alarm level (66*-O'), the spray at 4,000 gpm would generate a high level alarm in approximately 2.5 minutes. In another 2.5 minutes, the top of pool level would be reached, and a significant amount of water would be spreading to the floor drain system.

Operators would stop the fire protection pump, concurrent with isolating the broken piping header. At 4000 gpm, the pool would be diluted from 2000 to 700 ppm in approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Thus, a flow rate of this magnitude from this pipe break would be quickly recognized and isolated long before it could dilute the spent fuel pool boron concentration signif'cantly.

3.2.6 Dilution From Spent Fuel Pool Domineralizer l 1

When the spent fuel pool domineralizer is first placed in service after being recharged with fresh resin, it can initially remove boron from the water passing through it. In the worst case, assuming 75 ft' of I anion resin per domineralizer, it is conservatively estimated that 58 ppm of boron could be removed from the spent fuel pool water before the resin becomes saturated. The deborating effect of the demineralizers is modeled by removing 250 gpm of borated water per train and retuming 250 gpm of deborated water per train until the ion exchange capacity is depleted. Since each domineralizer I

normally utilizes a mixed bed of anion and cation resin, less boron would actually be removed before saturation, in addition, procedures are in place to flush a new resin bed with borated water prior to l aligning it for service and verify boron saturation via sample analysis. Because of the small amount of boron removed by the 'demineralizers, it is not considered a credible dilution source for the purposes of this evaluation.

)

3.3 Summary of Dilution Events APPROXIMATE DILUTION TIME SOURCE FLOW RATE (GPM) TO ALARM TO 700 PPM Boron Recycle System

- Holdup Tank to cleanup subsystem 255 40 min. 35hr.

(limited source volume) 19 L___--________--.-_---__

Reactoridakeup Water System

-2" connection to retum header 240 43 min. 37 hr.

(limited source volume)

Demineralized Water System j

-2" connection to retum header 190 57 min. 49 hr.

- 1* makeup to spent resin sluico header 55 3 hr. 168 hr.

Fire Protection System

- Fire hose stations in spent fuel pool area 350 29 min. 25 hr.

Spent Fuel Pool Domineralizers 500 N/A(insufficient resin capacity)

The eddition of unborated water from the fire protection system hoses provides the shortest dilution time. However, operators would detect the unusually large drawdown from the fire protection system and would also notice the overflowing spent fuel pool during their rounds every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The connection from the Boron Recycle System gives the next shortest dilution time, but the fluid is i normally borated to some degree and the volume of one recycle holdup tank is less than the volume required for dilution from 2500 to 700 ppm. The next shortest dilution time and the limiting scenario for I

r:ormal operation is based on using the reactor makeup water connection to the spent fuel pool cleanup subsystem for makeup when the process isolation valve is manually opened. This connection is the normal flowpath for unborated water authorized for use under normal plant conditions by procedure.

~ Although it is fed from a tank which has a capacity less than the required volume to dilute the spent 4 fuel pool from 2500 to 700 ppm, the tank receives automatic makeup from the demineralized water storage tank which has a volume greater than the required dilution volume.

For the limiting scenario to successfully result in the dilution of the spent fuel pool from 2500 ppm to l 700 ppm, the addition of 535,000 gallons of water to the spent fuel pool over a period of 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> would have to go unnoticed. The first indication of such an event would be high level alarms in the control room from the pool level instrumentation. If the high level alarms fail, it is reasonable to expect t

l that the significant increase in pool level and even:ual pool overflow that would result from a pool I

dilution event will be readily detected by plant operators in time to take mitigative actions. A pool overflow condition would result in flooding of the fuel handling building sumps, and significant input flow rates would result in high sump level alarms. Although area radiation monitors are available, relatively clean spent fuel pool contents might not set off an alarm, in addition, it can be assumed that the operator rounds through the spent fuel pool area that occur once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> will detect the increase in the poollevel even if the alarms fail and the flooding is not detected.

20 l

l 1

i Furthermore, for any dilution scenario to successfully add 535,000 gallons of water to the spent fuel l pool, plant operators would have to fail to question or investigate the continuous makeup of water to the reactor makeup water storage tank or fire protection system for the required time period, and fail to recognize that the need for 535,000 gallons of administrative makeup to the demineralized water storage tank was unusual.

S i

21 i

4.0 CONCLUSION

S A boron dilution analysis has been completed for the spent fuel pool. As a result of this spent fuel pool boron dilution analysis, it is concluded that an unplanned or inadvertent event which would result in the dilution of the spent fuel pool boron concentration from 2500 ppm to 700 ppm is not a credible event.

This conclusion is based on the following:

e in order to dilute the spent fuel pool to the design k, of 0.95, a substantial amount of I

water (535,000 gallons) is needed. To provide this volume, an operator would have to initiate the dilution flow, then abandon monitoring of pool level, ignore tagged valves, violate administrative procedures, and ignore spent fuel pool and building sump high level alarms.

Since such a large water volume tumover is required, a spent fuel pool dilution event would be readily detected by plant personnel via alarms, flooding in the fuel handling building or by normal operator rounds through the spent fuel pool area.

e it should be noted that this boron dilution evaluation was conducted by evaluating the time and water volumes required to dilute the spent fuel pool from 2500 ppm to 700 ppm. The 700 ppm end point was utilized to ensure that K, for the spent fuel racks would remain less than or equal to 0.95. As part of the criticality analysis for the spent fuel racks (Reference 1), a calculation has been performed on a 95/95 basis to show that the spent fuel rack K, remains less than 1.0 with non-borated water in the pool.

Thus, even if the spent fuel pool were diluted to zero ppm, which would take significantly more water than evaluated above, the spent fuel would be expected to remain subcritical and the health and safety of the public would be assured.

l

\

l I

5.0 REFERENCES

1. Houston Power & Light Company, South Texas Project. Units 1 and 2 Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit, December,1997

2. HL&P documents:

Process and Instrumentation Diaorams SR219F05028, " Spent Fuel Pool Cooling & Cleanup System P&lD," rev. 22 SR219F05029, " Spent Fuel Pool Cooling & Cleanup System P&lD,* rev.15 7R189F05010 " Boron Recycle System P&lD," rev.13 SR279F05033, " Reactor Makeup Water System P&lD," rev.12 6-S-10-0-F 00009, " Demineralized Water Storage P&lD," rev.16 SS199F05034, " Demineralized Water Distribution System," rev.13 7R309F05024," Liquid Waste Processing System P&lD," rev.14 SR209F05017," Component Cooling Water System P&lD," rev.18 50279F05047, " Fire Protection System P&lD," rev.19 Pioina drawinas 7M369PBR205, sh. 7, rev. 3 7M369PBR205, sh. A01, rev. 5 7M369PBR205, sh. 4, rev. 6 7M1269PFC230, sh. 2, rev. 5 7M369PFC230 sh. 4, rev. 4 SF369PRC530, sh. 4, rev. 9 3M369PBM2f>3, sh. 3, rev. 8 3M369PRM263 sh. 4, rev 7 SM369PRM263 sh. A01, rev. 7 3F369PRM563 sh. A01, rev. 5 9Y360PDW724, sh. 24, rev. 4 9Y360PDW724 sh. 4, rev.1 9M361PDW224 sh. 3, rev. 0 9F369 POW 524 sh. 2, rev. 2 9F369PDW524 sh. A01, rev. 2 9M369PDW224 sh. A09, rev. 6 7M369PWL277 sh. A25, rev 4 7M369PWL277 sh.11, rev 3 7M369PWL277 sh. A82, rev.1 6M369PFC230 sh. A04, rev.1 7M369PFC230 sh. 06, rev. 3 Procedurga l OPOP02 FC-0001, " Spent Fuel Pool Cooling and Cleanup System," rev.16 OPOP02-BR-0001,

• Boron Recycle System Operation," Rev. 4, Section 21.0 OPOP02-RM-0001, " Reactor Makeup Water System Operat5ns," rev. 3

' OPOPO4-FC-0001, " Loss of Spent Fuel Pool Level or Cooling," rev. 7 OPCP01 ZA-0038, " Plant Chemistry Specifications," rev.10 OPOP09-AN-22M2, " Annunciator Lampbox 22M02 Response Instructions," Rev. 6 23

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25

ATTACHMENT 4 NO SIGNIFICANT HAZARDS EVALUATION l

I 1

I i

f Attachnent d NOC-AE-00178 l

l l l

ATTACHMENT 4 NO SIGNIFICANT HAZARDS EVALUATION Pursuant to 10CFR50.91, this analysis provides a determination that the proposed change to the Technical Speci6 cations described previously does not involve any signincant hazards consideration as defined in 10CFR50.92, as described below:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The presence of soluble boron in the spent fuel pool water for criticality control does not increase the probability of a fuel assembly drop accident in the spent fuel pool. The handling of the fuel assemblies in the spent fuel pool has always been performed in borated water.

The criticality analysis shows the consequences of a fuel assembly drop accident in the spent fuel pool are not affected when considering the presence of soluble boron. The rack &n remains less than or e. qual to 0.95.

There is no increase in the probability of an accident. The proposed change does allow a greater number of fuel storage configurations in the spent fuel pool. While this could increase the probability of a fuel misloading, the presence of sofHcient soluble boron in the spent fuel pool precludes criticality as a result of the misloading. Puel assembly placement will continue to be controlled pursuant to approved fuel handling procedures and will be in accordance with the Technical Speci0 cation spent fuel rack storage configuration limitations.

There is no increase in the consequences of the accidental misloading of spent fuel assemblies into the spent fuel pool racks. The criticality analyses demonstrate that the pool &n will remain less than or equal to 0.95 following an accidental misloading due to the boron concentration of the pool.

The proposed Technical SpeciGcation limitation will ensure that an adequate spent fuel pool boron concentration is maintained.

There is no increase in the probability of the loss of normal cooling to the spent fuel pool water when considering the presence of soluble boron in the pool water for suberiticality control since a high concentration of soluble boron has always been maintained in the spent fuel pool water.

Reactivity changes due to spent fuel pool temperature changes have been evaluated. The base case criticality analysis covers a " normal" spent fuel pool temperature range of 50 F to 160 F. Spent fuel pool temperature accidents are considered outside the normal temperature range extending from 32 F to 240 F. In all spent fuel pool temperature accident cases, suf6cient reactivity margin is available to the 0.95 Nalimit without requiring additional soluble boron above the base case level. Because adequate soluble boron will be maintained in the spent fuel pool water to maintain Nng 0.95, the consequences of a loss of normal cooling to the spent fuel pool will not be increased.

l Attachnent 4 NOC-AE-00178 i

j l

l _ _ _ _ _ _ ]

ATTACHMENT 4 NO SIGNIFICANT HAZARDS EVALUATION Therefore, based on the conclusions of the above analysis, the proposed changes do not involve a significant increase in the probability or consequences of an accident pmviously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident  !

from any accident previously evaluated.

Spent fuel handling accidents are not new or different types of accidents, they have been analyzed l in Section 15.7.4 of the Updated Final Safety Analysis Report.

l Criticality accidents in the spent fuel pool are not new or different types of accidents, they have been analyzed in the Updated Final Safety Analysis Report and in Criticality Analysis reports associated with specific licensing amendments for fuel enrichments up to and exceeding the nominal 4.95 weight percent U 2 that is assumed for the proposed change.

Current Technical Specifications contain limitations on the spent fuel pool boron concentration.

The actual boron concentration in the spent fuel pool has been maintained at a higher value. The proposed changes to the Technical Speci0 cations establish new boron concentration requirements for the spent fuel pool water consistent with the results of the new criticality analysis (Attachment 2).

Since soluble boron has always been maintained in the spent fuel pool water, and is currently l required by Technical Specifications, the implementation of this new requirement will have little effect on normal pool operations and maintenance. A dilution of the spent fuel pool soluble boron i has always been a possibility; however, it was shown in the spent fuel pool dilution evaluation l (Attachment 3) that there are no credible dilution events for which the spent fuel pool Kn could j increase to greater than 0.95. Therefore, the implementation of new limitations on the spent fuel pool boron concentration will not result in the possibility of a new kind of accident.

The proposed changes to Technical Specifications 3.9.13,4.9.13, and 5.6 continue to specify the requirements for the spent fuel rack storage configurations. Since the proposed spent fuel pool storage configuration limitations will be similar to the current ones, the new limitations will not have any significant effect on normal spent fuel pool operations and maintenance and will not create any possibility of a new or different kind of accident. Verifications will continue to be perfonned to ensure that the spent fuel pool loading configuration meets specified requirements.

The misloading of a fuel assembly in the required storage configuration has been evaluated. In all cases, the rack Kn remains less than or equal to 0.95. Removal of an Rod Control Cluster Assembly from a checkerboard storage configuration has been analyzed and found to be bounded by the misloading of a fuel asserably.

Attachnent 4 NOC-AE-00178

ATTACHMENT 4 NO SIGNIFICANT HAZARDS EVALUATION As discussed above, the proposed changes will not create the possibility of a new or different kind of accident. There is no significant change in plant configuration, equipment design or equipment.

Under the proposed amendment, no changes are being made to the racks themselves, any other systems, or to the physical structures of the Fuel Handling Building itself. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The Technical Specification changes proposed by this License Amendment Request and the resulting spent fuel storage operation limits will provide adequate safety margin to ensure that the stored fuel assembly array will always remain suberitical. Those limits are based on a plant specific criticality analysis (Attxhment 2) performed in accordance with Westinghouse spent fuel rack criticality analysis methodology.

While the criticality analysis utilized credit for soluble boron, storage configurations have been defined using 95/95 La calculations to ensure that the spent fuel rack Kg will be less than 1.0 with no soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide suberitical margin such that the spent fuel pool Kg is maintained less than or equal to 0.95.

The loss of substantial amounts of soluble boron from the spent fuel pool which could lead to La exceeding 0.95 has been evaluated (Attachment 3) and shown to be not credible. A safety evaluation has been performed which shows that dilution of the spent fuel pool boron concentration from 2500 ppm to 700 ppm is not credible. Also, the spent fuel rack Kg will remain less than 1.0 (with a 95/95 confidence level) with the spent fuel pool flooded with unborated water. These safety analyses demonstrate a level of safety comparable to the conservative criticality analysis methodology required by Westinghouse WCAP-14416.

Based on the above evaluation, the South Texas Project concludes that the proposed changes to the Technical Specifications involve no significant hazards consideration.

l Attachnent 4 NOC-AE-00178

i l

ATTACHMENT 5 Proposed Changes to the South Texas Project Updated Final Safety Analysis Report for Revised Spent Fuel Rack Utilization Plan l

Attachment $

NOC-AE-00178 I

STPEGS UFSAR t

l The Fuel Storage and Handling and Radioactive Waste Systems are designed to assure adequate safety under nor1nal and postulated accident conditions.

For further discussion, see the following UFSAR sections:

Residual Heat Removal System 5.4.7 Fuel Storage and Handling 9.1 Air Conditioning, Heating, Cooling, and Ventilation Systems 9.4 Radiation Protection 12.0 All Other Systems Required for Safety 7.6 Engineered Safety Features Actuation System 7.3 Radioactive Waste Management 11.0 3.1.2.6.3 Criterion 62 - Prevention of Criticality in Puel Storace and Handlinct Criticality in the Fuel Storage and Handling System shall be l prevented by physical systems or processes, preferably by use of geometrically safe configurations.

3.1.2.6.3.1 Evaluation Acainst Criterion 62 - Appropriate plant fuel ,

handling and storage facilities are provided to preclude accidental l criticality for new and spent fuel. Criticality in the new fuel storage rack is prevented by the geometrically safe configuration of the storage rack.

There is sufficient spacing between the assemblies to assure that the array when fully loaded is substantially subcritical. Criticality in the spent fuel storage rack is prevented by both the geometrically safe configuration of the storage rack and the use of administrative procedures to control the placement of burned and fresh fuelp Storage of fuel assemblies are limited by design to the loading and rack storage position.

New fuel and spent fuel storage rack design details are provided in Section 9.1. ( g,} y,g g &gg,_6/; eg The new fuel racks are designed to withstand nominal operating loads as well as SSE and Operating Basis Earthquake (OBE) seismic loads meeting Safety Class (SC) 3 and American Institute of Steel Construction (AISC) requirements. The new fuel racks are designed to withstand maximum uplift force of 5,000 pounds.

The center-to-center distance between the adjacent fuel assemblies is sufficient to ensure a K.cr 1 0.95 even if unborated water is used to fill the new fuel storage area. gThe K,,, of the spent fuel storage racks is maintained at [o owl a) less than or equal to t..,:: ~*-a if ""knv=tef untu is uccd to fill G.c spent piel atn,ana nnni 5./ h:th the ntent: ccater distance hotcreen the adjacent Dia' =ne-Elic :nd th; -- nr -d-M -tr;;i c; pr:::drec to-contre 14he placement nf hornaa and < :-h fsel, i%cLdq wpMdn uJ h/& ac'F 9 ik97 b4sif, t ven i[ Wsbor< W wu b g 4 scd h 0 R 5 Sft'd I J "Y f

l 9 ,4..m., ,, ,/ a 5 o.w c.,a 4 u <s,+.W ~ o " "' "

ted M n N N presta g of 3,3,34 Sped del fo l DgbhRevision loroa 5.

STPEGS UFSAR 4.3.2.6.1 New Fuel Storace:

New fuel is stored in 21-inch center-to-center racks in the new fuel storage facilities in a dry condition.

For the flooded condition and for the low water density optimum moderator condition (with unborated water assuming fuel of the highest anticipated enrichment exceed 0.95. of 5.0 w/o U-235), the effective multiplication factor does not In the analysis for the storage facilities, the fuel assemblies are assumed to be in their most reactive condition, namely fresh or undepleted and with no control rods or removable neutron absorbers present.

Credit is taken for the

! inherent neutron-absorbing effect of the construction materials of the racks .

Assemblies cannot be closer together than the design separation provided by ,

1 the storage facility, except in special cases such as in fuel shipping containers the design. where analyses are carried out to establish the acceptability of In the case of an accident that would ir. crease reactivity, such as an assemblydrop in the normal dry condition (k,,, s 0.70), the maximum k,,,

less than 0.95. will be i (Reference 4.3-40). This includes allowances for biases and uncertainties I 4.3.2.6.2 Soent Puol Storac_e:

i 4.3.2.6.2.1 Analysis Methodology _The criticality analysis of the rack is periormea with the AMn/ KENO Va code packages.

(Reference 4.3 35) is used to generate needed cross section data and theThe 'O AMPX c Va computer code (Reference 4.3-36) is used to determine rack reactivit .

KENO Va has been benchmarked against critical experiments that are representative of the STPEGS design. This benchmarking led to the nelusion that the calculational model is capable of determining the multip cation factor of the racks to within 0.0074 AK, with an uncertainty of .0029 AK at a 95 percent probability at the 95 percent confidence level.

i i

The effects of various mechanical and thermal variances a conservatively treated by combining the worst values applying the most imiting parameter variances together at one time in a KENO model. Thes include pool water temperature, stainless steel thickness variation, a orber panel dimensional variation and boron loading, fuel enrichment vari ion, and uranium dioxide l l pellet density. These " worst case" basic cell '

,, s are 0.9221 for Region 1 racks fuel. The forFfresh

%,, of all4.0 w/o fuel storage and are patterns 0.9267 for R ion 2 racks for fresh 1.7 w/o ss than 0.95 including biases and uncertainties, thus meeting the requireme s of GDC 62.

To accommodate higher fuel enrichment in close packed storage in both rack regions, reactivity credit for asse ly burnup and for the presence of I

Westinghouse Integral Fuel Burnab fuel assemblies is allowed. Bu up credit Absorbers (IFBA) (Reference 4.3-37) in is achieved by the use of reactivity equivalencing. Th concept of reactivity equivalencing is based upon the reactivity decreas. associated with fuel depletion. A series of reactivity calculations i assembly discharge bur performed to generate a set of enrichment-fuel ordered pairs which yield the equivalent K.,, when the fuel is stored i the racks. A separate series of calculations is performed for the gion 1 and Region 2 racks.

=

Repl a ce M %A rema.%de r 04 774.17./,.2.I M N#

4.3-29 Revision 5

1 of 2 INSERT 1 The reactivity of the spent fuel rack is analyzed in Reference 4.3-41. To previde safety I

margin in the criticality analysis of the spent fuel racks, credit is taken for the soluble boron present in the Region 1 and 2 spent fuel pcol. This parameter provides significant negative reactivity in the criticality analysis of the spent fuel rack and will be used here in conjunction with administrative controls to offset the reactivity increase when ignoring the presence of the spent fuel rack Boraflex poison panels. Soluble boron credit provides sufficient relaxation in the enriclunent limits of the spent fuel racks. Reference 4.3-42 provides the evaluation of spent fuel pool dilution. It is shown that there is no credible event which would result in a spent fuel pool dilution from the required soluble boron concentration (2500 ppm) to the minimum soluble boron concentration that assures K,g s 0.95 (700 ppm).

The design basis for preventing criticality in the spent fuel pool is:

1. the effective neutron multiplication factor, K,g, of the fuel rack array will be less than 1.00 in pure, unborated water, with a 95 percent probability at a 95 percent confidence ,

confidence level, including uncertainties; and,

2. the effective neutron multiplication factor, K,g, of the fuel rack array will be s 0.95 in the pool containing borated water, v4 a 95 percent probability at a 95 percent  ;

confidence level, including uncertainties. {

The purpose of this section is to present the storage requirements, including maximum nominal enrichments, minimum burnup values, minimum decay times, minimum IFBA content, storage configurations, and the minimum pool soluble boron concentration.

With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the various types of 17x17XL fuel do not contribute to any increase in the basic assembly reactivity. This includes small changes in guide tube and instrumentation tube dimensions. Therefore, future fuel assembly upgrades do not require a criticality analysis if the fuel diameter continues to be 0.374 inches.

l The fuel rod, guide tube and instrumentation tube claddings are modeled with zircaloy in l this analysis. 'I his is conservative with respect to the Westinghouse ZlRLOm product which is a zirconium alloy containing additional elements including nicui m. Niobium has a small absorption cross section which causes more neutron capture a 'he cladding .

regions resulting in a lower reactivity. Therefore, this analysis is conser iu vith respect to fuel assemblies containing ZIRLO cladding in fuel rods, gt ' ubes, and the instrumentation tube.

i%oppes\ insert _I doc

2 of 2 INSERT 1, con't Empty water cells may be substituted for fresh or burned fuel assemblies at any location.

Any positive reactivity effect due to the additional water in the 2x2 analytical cell is offset by the absence of uranium.

When storing fuel with an initial nominal enrichment greater than the maximum all-cell enrichment for the respective rack region, a rack K,g ofless than equal to 0.95 is ensured by the maintenance of a minimum amount of soluble boron and the use of administrative procedures to control the placement of burned and fresh fuel and RCCAs. A rack K,g of less than 1.00 in pure, unborated water, is ensured by the use of administrative procedures to control the placement of burned and fresh fuel and RCCAs. Guidance on the close packed storage of fresh and burned fuel, and fuel containing IFBA's, is provided in Section 5.6 of the Technical Specifications.

The licensing basis for the racks is met by the combination of the physical design and center-to-center spacing of the storage cells, the required presence of soluble boron, and the use of administrative procedures to guide the placement of fuel assemblies and RCCAs.

I r

1-A hoppes\ insert _1. doc

STPECS UFSAR The depletable, two-dimensional, transport theory code PHOENIX (Referen 4.3

37) is used to perform the reactivity equivalence calculations. PHOEN has been validated by comparisons with both benchmark critical experiment and experiments where the isotopic fuel composition has been examined fo owing discharge from a reactor (Reference 4.3-37).

Since the burrup history of fuel assemblies which will be dischar ed in the future is not known exactly, a reactivity uncertainty is applied to the burnup-dependent reactivities computed with PHIV. NIX. An uncer inty which increases linearly with burnup, passing through 0.01 AK at 30 00 MWD /MTU is applied to the PHOENIX calculational results in the developm t of the burnup requirements. This uncertainty is considerr.d to be very co ervative and is based on consideration of the good agreement between PHOEN predictions and measurements'(Reference isotopic buildup variances.4.3-37) and on conservative estim tes of fuc1 assembly Reactivity credit for the presence of Westinghouse In gral Fuel Burnable Absorbers (IFBA) (Reference 4.3-37) in fuel assembli is also achieved by the use of reactivity equivalencing (Reference 4.3-37) FBA's consist of neutron absorbing material applied as a thin ZrB coating the outside of the UO2 fuel pellet. As a result, the neutron absorbing aterial is a non-removable or integral part of the fuel assembly once it is manufactured.

In this case, the concept of reactivity equiv encing is based upon the reactivity decrease associated with the pres ce of IFBA's. A series of reactivity calculations, to calculate a fue assembly reference K ing, is performed using PHOENIX. The fuel assembi 3 to determine the maximum assembly reactiv ty.containing the IFBA's is depleted These results are then used to generate yield a set of enrichment-fuel the equivalent K int when theassemb y IFBA content ordered pairs which fuel s stored in the racks. IFBA credit is only used for Region 1 racks.

The fuel burnup used in the reactiv highest equivalent K err when the fu y calculation is that burnup which yields assembly depletions performed in is stored in the spent fuel racks. Fuel per assembly considered in this OENIX show that for the number of IFBA pins nalysis, the maximum reactivity for the rack geometry occurs at zero burnup.' Although the boron concentration in the IFBA pins decreases with fuel dep1 irin, the fuel assembly reactivity decreases more rapidly, resulting in a aximum fuel rack reactivity at zero burnup.

Uncertainties associated w th the IFBA dependent reactivities computed with PHOENIX equivalenceare accounted for in the development of the individual reactivity limits.

approximately 10% of For t IFBA credit applications, an uncertainty of total number of IFBA pins is accounted for in the development of the IF requirements.

The following equat on is used to develop the maximum K err for the spent fuel storage racks:

K.tr- Kwor + B..thod + Bp ,s + ([ (ks)2,,,,,e + (ks) 2 e,L,,d j l

l t

4.3-30 Revision 3 l

STPEGS UFSAR where:

K ,, = worst case KENO K,,, that includes material, mechani 1, and enrichment tolerances.

B, t ,,, =

method bias determined from benchmark critical omparisons Byn =

method bias to account for Boraflex absorber article self- l shielding ks ,n = 95/95 uncertainty in the worst case KEN rr ks-o.a = 95/95 uncertainty in the method bias.

The neutron absorber material used in both rac designs consists of Boraflex l panels between the rack cells. The racks us at STPEGS are designed to minimize both the formation of gaps and t size of gaps which are formed.

However, a conservative analysis was pe ormed considering gap formation at the midplane of the fuel assemblies. iven the margin available to the 0.95 Kcr limit, for a scenario of the d elopment of gaps in all four panels j surrounding an assembly, gaps up o 3.75 inches in length are acceptable, in '

Region 1. For a scenario of t development of gaps in only two of the four panels surrounding an asse , gaps up to'7.75 inches in length are j acceptable. In Region 2, aps up to 3.00 inches in length are acceptable for the four panel assumpti and gaps up to 6.25 inches are acceptable for the two panel case.

I A similar series f analyses indicate that the end of the Boraflex panels may shrink up to 8 5 inches in the Region 1 racks and up to 7.50 inches in the Region 2 ra .s .

A Boraf' x surveillance program is followed to detect degradation of the abso erpanelsandtoinstituteappropriatecorrectiveactionsifdegradationl 7 of he panels is indicated.

2.50 nom;+rl l 4.3.2.6.2.2 Pecion 1 Rack Desien - The Region 1 racks have a 10.95 inch-nominal center-to-center spacing with locked removable absorber assemblies l between the cells. This region is conservatively desigaed to accommodate close packed storage of unirradiated fuel enriched to ,AOC weight percent uranium-235. l The following equation is used to develop the maximum F6cc fo '

the spent fuel storage racks:

Kerr" Karn + B tn,4 +Ban + /((ks)8 ,n + (ks) 8. nag]

where:

K . ,,t = worst case KENO K.,, that includes - terial, mechanical, and enrichment tolerances, i

B.,t w, -

method bias determined fr - benchmark critical comparisons l B yn = method bias to accoun or Boraflex absorber particle self- l shie?. ling ks.,, = 95/93 uncerta . y in the worst case KENO K,c j ks nnu = 95/'35 une ainty in the method bias j Substituting c ulated values in the order listed above, the result is:

FL,, = 0 21 + 0.0074 + 0.0014 + /((0.0041)2 + (0.0029)8] = 0.9359 4.3-31 Revision 5

l I

STPEGS UFSAR Since K rr is less than 0.95 including uncertainties at e = M, probability / confidence level, the ace 1 eria for criticality is met for Region 1 close age of fuel assemblies enriched to a nominal 4.0 w -

7,50 n em i a 4 l Storage of close packed fuel with nominal enrichment of greater than eve- w/o is achievable by the use of reactivity equivalencing for burnup credit and the presence of IFBA pins, as discussed in Sectic: 1.0.;.0.2.1. I

~

R4 h<e.u 4 3 4f. j xeaccavacy equivaAencing mor ournup crealc allowc tuel with an initial nomina enrichment of greater than 4.0 w/o to be stored in a close packed array if t e fuel assembly K.,, is less than the constant Kore contour given in Section 5.

of the Technical Specifications. This minimum bcrnup curve starts at 4. w/o  !

at 0 MWD /MTU and ends at 5.0 w/o at 5400 MWD /MTU. The curve includes a l reactivity uncertainty of 0.0018 AK, consistent with the minimum burn requirement of 5400 MWD /MTU at 5.0 w/o.

Reactivity equivalencing for IFBA credit allows fuel with an ini al nominal enrichment of greater than 4.0 w/o to be stored in a close pac d array if the fuel assembly reference Kar is less than or equal to 1.445. figure reflecting this constant Ka, is given in Section 5.6 of the echnical Specifications. This curve reflects the minimum number o IFBA pins required in an assembly for close packed storage. The curve sta s at 4.0 w/o and no IFBA pins and ends at 5.0 w/o and 80 IFBA pins.

The IFBA absorber material is a zirconium diboride ZrB ) coating on the 2

outside of the fuel pellet (Reference 4.3-37) . ch IFBA pin has a nominal absorber material loading of 1.57 milligrams B ' per inch, which is the 2

l minimum standard loading offered by Westinghe se for 17x17XL/STD fuel assemblies. The IFBA B ' loading is reduced y 5% to conservatively account 2

for manufacturing tolerances and then by additional 28.5% to conservatively model a minimum absorber length of 120 ches. l Additional IFBA credit calculations re performed to examine the reactivity effects of higher IFBA linear B ' 1 dings (2.36 and 3.14 mg/in) . These 2

calculations confirm that the ass mbly reactivity remains constant provided the net B ' material per assembl is preserved. Therefore with higher IFBA B' 2

2 loadings, the required number f IFBA pins per assembly can be reduced by the ratio of the higher loading o the nominal 1.57 mg/in loading.

The IFBA requirements we developed based on the standard IFBA patterns used by Westinghouse. Howe r, since the worth of individual IFBA pins can change depending on positio within the assembly (due to normal variations in thermal flux), studies were erformed to evaluate the effect and a conservative reactivity margin as included in the development of the IFBA requirement to account for thi effect. This assures that the IFBA requirement remains valid at intermediate enrichments where standard IFBA patterns may not be available.

In addition, o conservatively account for calculational uncertainties, the IFBA requi ments also include a conservatism of approximately 10% on the l total nur er of IFBA pins at the 5.0 w/o end.

i l

Puel y also be stored in a checkerboard fashion. Two fresh fuel assemblies wit nominal enrichments of 5.0 w/o may be stored with two burned assemblies wb ch meet a burnup credit criteria. For determination of the rack riticality, the_"_ burned" assemblies are represented by fresh fuel assemblies 4.3 32 Revision 5 1

I STPEGS UFSAR

, of low enrichment. For the storage of fuel in a checkerboard pattern in Region 1 racks, the fresh fuel is assumed to be at a nominal enrichment of 5.0 w/o. The burned fuel is assumed to have a reactivity equivalent to fresh assembly with an initial nominal enrichment of 2.8 w/o. )

The equation stated above is again used to develop the maximum Eyre r the j storage of fue'. in a checkerboard pattern. Substituting calculat d values in the order listed above, the result is:

K.rt = 0.9109 + 0.0074 + 0.0014 + /[(0.0047)8 + (0.0029)8] = .9252 Since FLer is less than 0.95 including uncertainties at 95/95 probability / confidence level, the acceptance criteria or criticality is met for the Region 1 checkerboard arrangement of fresh el assemblies at a nominal enrichment of 5.0 w/o and burned fuel ass lies with reactivities i equivalent to that of fresh 2.8 w/o assemblies.

! Reactivity equivalencing for fuel burnup is 1so performed for the burned fuel assemblies in the checkerboard arrangemen . Burned assemblies with an initial nominal enrichment of greater than 2.8 o may be stored in a close packed l array if the fuel assembly K rr is les than the constant Ekre contour given in I Section 5.6 of the Technical Specif ations. This minimum burnup curve starts l at 2.0 w/o O MWD /MTU and ends at .0 w/o at 17,500 MWD /MTU. The curve l includes a reactivity uncertain of 0.0058 AK, consistent with the minimum burnup requirement of 17,500 /MTU at 5.0 w/o.  !

Empty water cells may be stituted for fresh or burned fuel assemblies at i any location. Any posi ve reactivity effect due to the additional water in I the 2x2 analytical ce is offset by the absence of uranium. This is j

illustrated by the e of empty water holes in the checkerboard pattern used i in Region 2 racks discussed in Section 4.3.2.6.2.3. In Region 2, fresh, no IFBA, 5.0 w/o a emblies may be stored in a checkerboard pattern with empty l water holes. similar checkerboard storage pattern in Region 1 is allowable l since the R ion 1 racks have a greater cell to cell spacing than Region 2 I racks. T wider assembly spacing would serve to further reduce the rack reac*,iv y.

Whe storing fuel with an initial nominal enrichment of greater than 4.0 w/o, a ack K. of less than or equal to 0.95 in the Region 1 racks is ensured by e use of administrative procedures to control the placement of burned and fresh fueld Guidance on the close packed storage of fresh and burned fuel, jr and fuel containing IFBA's, is provided in Section 5.6 of the Technical Specifications. Similar guidance is provided on the storage of fresh and burned fuel in a checkerboard pattern in Region 1 racks. Empty water cells may be substituted for fuel assemblies in all cases.

_ .2o nu H asd l 4.3.2.6.2.3 Recion 2 Rack Desion - The Region 2 racks have a 9.15 inch-nominal center-to-center spacing with fixed absorber material surrounding each cell. This region is conservatively designed to accommodate unirradiated fuel enriched to J<7' weight percent uranium-235. [ The foll -

equation is used to develop the maximum Ke rr for the s orage racks:

gn + [(ks)8... + (ks ) 8,,,cul 4.3-33 Revision 5

l l

STPEGS UFSAR l

where:

Kom  !

=

worst case KENO K.cr that includes material, mechanica and enrichment tolerances.

B.cu, =

method bias determined from benchmark critic comparisons I B yn =

method bias to account for Boraflex abso - r particle self-shielding l ks om. = 95/95 uncertainty in the worst c KENO K,cc

)

ks.,twa = 95/95 uncertainty in the met bias Substituting calculated values i e order listed above, the result is:

F,gg = 0.9267 + 0.0074 +

14 + /[.0049)8 + (0. 0029) *] = 0.9412 l

Since Kerr is ss than 0.95 including uncertainties at a 95/95 probabi for

/ confidence level, the acceptance criteria for criticality is met

-gion 2 close packed st.orage of fuel assemblies enriched to a nominal

. w/o uranium-235.

1.2Dnomid Storage of close packed fuel with nominal enrichment of greater than F -w/o is achievable by the use of reactivity equivalencing for burnup credit, as discussed in Fertirn 4.2.2.0.2.1.

R t {er e.ot d.1 4I.

Reactivity equivalencing for burnup credit allows fuel with an initial nomi enrichment of greater than 1.7 w/o to be stored in a close packed array i the fuel assembly K.,, is less than the constant K,ge contour given in Secti 5.6 of the Technical Specifications. This minimum burnup curve starts 1.7 w/o at 0 MWD /MTU and ends at 5.0 w/o at 42,000 MWD /MTU. The curve i udes a reactivity uncertainty 0.0140 AK, consistent with the minimum rnup requirement of 42,000 MWD /MTU at 5.0 w/o.

Puel may also be stored in a checkerboard fashion. fresh fuel assemblies with nominal enrichments of 5.0 w/o may be stored . a 2x2 checkerboard pattern with two empty water cells, i.e., each el assembly is separated from its nearest adjacent neighbor by an empty wa cell.

The equation stated above is again use o develop the maximum K,c, for the storage of fresh or burned fuel in eckerboard pattern. Substituting calculated values in the order li ed above, the result is:

Ke rr = 0.8634 + 0.0074 + 0.0 + /[0.0062)8 + (0.0029)8] = 0.8790 Since Kerr, is less th 0.95 including uncertainties at a 95/95 probability / confide e level, the acceptance criteria for criticality is met for the Region 2 eckerboard arrangement of empty cells and fresh fuel assemblies at nominal enrichment of 5.0 w/o uranium-235.

When st ng fuel with an initial nominal enrichment of greater than 1.7 w/o, a rac K,c, of less than or equal to 0.95 in the Region 2 racks is ensured by th use of administrative procedures to control the placement of burned and resh fuel.l Gulcance on the close packed storage of burned fuel is provided in Section 5.6 of the Technical Specifications. Similar guidance is provided on the storage of fresh and burned fuel in a checkerboard pattern in Region 2 racks. Empty water cells may be substituted for fuel assemblies in all cases, i

4.3-34 Revision 5 l

L____------_----

k W r Ce, c Q. re s<re-t

• A 4n de hd h feelim 51 6 STPECS UFSAR ef & T:ce L;cd Spec;/ic /,*n s.

4.3.2.6.2.4 Storare Configuration Interf ce Requirements - When the two storage areas meet at an interface, the type f interface is explicitly defined to ensure the rack K.rr limit is met.)JThe transition schemes /

described below will be used at the interface of two storage configuration areas in the spent fuel racks. Empty water cells may be substituted for fuel assemblies in all cases.

The interface between a close packed fuel storage area and a checkerbo ded storage area in Region 1 shall be such that either: (1) the lower re civity burned fuel assembly in the checkerboard pattern is carried into t first row of the close packed storage area of fuel; or, (2) at least one r of empty water cells separate a close packed fuel storage area and a e ckerboarded storage area. Empty water cells may be substituted for fue assemblies in all cases.

The interface between a close packed fuel storage are and a checkerboard storage area in Region 2 shall be such that either- (1) there is a one row carryover of alternating empty cells from the ch erboard area into the first row of the close packed area with the remainin cells of the row filled with Region 2 "close packed"-type assemblies; or ) at least one empty row of cells separates the checkerboard pattern ea and the close packed storage area. Empty water cells may be substit ed for fuel assemblies in all cases.

There are no restrictions on the i erface between Region 1 close packed storage areas and adjacent close acked storage areas in Region 2.

The interface between a chec >rboarded storage area in Region 1 and any Region 2 rack storage area vill b such that either: (1) the Region 1 checkerboard pattern is carried to t interfae, but the last row at the interface leaves 3

the Category 1 fuel a embly pos: cions vacant; or (2) at least one row of empty water cells i either Region 1 or Region 2 racks separate the Region 1 ,

checkerboarded s rage area and the Region 2 rack storage area. '

The interfac etween a checkerboarded storage area in Region 2 and any Region 1 r k storage area vill be such that at least one row of empty water cells i either Region 1 or Region 2 racks separate the Region 2 check oarded storage area and the Region 1 rack storage area.

I checkerboarded storage areas in both Regions 1 and 2 are adjacent, at least one row of empty water cells in either Region 1 or Region 2 racks will separate the checkerboarded storage areas in the respective racks.

4.3.2.6.2.5 Rack Utilization - Section 5.6 of the Technical Specifications describes the storage of fresh, burned, and IFBA-containing fuel in the spent fuel storage racks. Both close packed and checkerboarded storage of fuel is allowed in Region 1 and Region 2 racks, depending on the reactivity of the fuel assembly.

The reactivity characteristics of fuel assemblies which are to be placed in the spent fuel storage racks are determined and the assemblies are categorized 2

y reactivity. Alternately, if necessary, all assemblies may be treated as if each assembly is of the highest reactivity class until the actual assembly reactivity classification is determined. Section 5.6 of the Technical Specifications provides the definitions of the reactivity classifications and the allowed storage patterns.

4.3-35 sevision 3 I

l

a SMc}( M 5(d a c < d 3 N~ .

1 STPEGS UFSAR The boron concentration of t water in the spent fuel pool is maintained at or above the mP61 mum value ne ded to ensure that the rack Ein is less than or equal in ek. .v.n*

to 0.95 in the event of mirplic d 2:::=hlic: in the negica 1 ..si. oud' p;;[9- 9 1 _. *_ m nr m e4ng1. -4 rpi red highcst scactiviiq w=Lcwusy assembif .. m.e 4.3.2.6.3 In-Containment Storace of Fresh and Soent Fuel The in-containment storage racks provide for the temporary storage of both fresh and burned fuel during refueling operations. The fuel is stored in 16 inch center-to-center stainless steel racks in the in-containment storage area (ICSA). The ICSA is flooded with borated refueling water when fuel is present.

4,p g The n-thodology used to analyze the ICSA is the same as that used for the Spent Puel Storage Racks, as described in S;; tic; .2.0.0.0.1. The ICSA racks htye a 16 inch-nominal center-to-center spacing. This region is conservatively designed to accommodate close packed storage of unirradiated fuel enriched to 4.5 weight percent uranium-235. The following equation is used to develop the maximum Pin for the ICSA fuel storage racks:

K.cr = Koor.c + B th.a + / { { ks ) *,,,,e + (ks) 2.,ta ]

where:

K.,,e =

worst case KENO F4n that includes material, mechanical, and enrichment tolerances B otn , =

ke =

method bias determined from benchmark critical comparisons ks.,tu.

,,e 95/95 uncertainty in the worst case KENO Y,n

=

95/95 uncertainty in the method bias.

Substituting calculated values in the order listed above, the result is:

Fin = 0.9315 + 0.0074 + / [(0.0064)3 + (0.0029)*] = 0.945.9 Since K,n is less than 0.95 including uncertainties at a 95/95 l probability / confidence level, the acceptance criteria for criticality is met for ICSA storage of fuel assemblies enriched to a nominal 4.5 w/o uranium-235.

Storage of close packed fuel with nominal enrichment of greater than 4.5 w/o is achievable by taking credit for the presence of IFBA pins, as discussed in se-*ir- .2.2 !.2.1.

l

[14Q-W 47 4/.

Reactivity equivalencing for IFBA credit allows fuel with an initial nominal enrichment of greater than 4.5 w/o to be stored in a close packed array if the fuel assembly reference Ka, is less than or equal to 1.484. A figure reflecting this constant Ka, is given in Section 5.6 of the Technical Specifications.

This curve reflects the minimum number of IFBA pins required in an assembly for close packed storage. The curve starts at 4.5 w/o and no IFBA pins and ends at 5.0 w/o and 36 IFBA pins. Note that fuel assemblies categorized as Category 2, 3, or 4 (as defined in Technical Specification 5.1.6.2), have a lower reactivity than fuel with an initial nominal enrichment of greater than 4.5 w/o and no IFBA's. Therefore, such assemblies may be stored in a close packed array in the ICSA.

The IFBA absorber material is a zirconium diboride (ZrB ) coating on the 3 outside of the fuel pellet (Reference 4.3-4 0) . Each IFBA pin has a nominal poison material loading of 1.57 milligram B" per inch, which is the minimum 1

4.3-36 Revision 5

STPEGS UFSAR 4

9.1.2.3 Safety Evaluation. Units 1 and 2 of the STPECS are each provided with separate and independent fuel handling facilities. Flood protection of each MIB is discussed in Section 3.4.1. A detailed discussion of missile protection is provided in Section 3.5.

The applicable design codes and the various external loads and forces j considered in the design of the FHB are discussed in Section 3.8.4. Details i of the seismic design and testing are presented in Section 3.7. l,oo I Design of this storage facility in accordance with GDC 62 and RG 1.13 nsures

• j a safe condition under normal and postulated accident conditions, e N,, of the spent fuel storage racks is maintained less than or equal to . , even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administra-tive procedures to control the placement of burned and fresh fueleh wMl Ms, Under accident conditions, the 4,, is maintained well below 0.95 assuming 2200 -999 ppa borated water. The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensure that the rack K,, is less than or equal to 0.95 in the event of misplaced assembliseeAm 2.; ;hu p ? : 2 ;;.;.... .. .= :: = L 2 2 2:! et -22 2::::. Con- 5 sideration of criticality safety is discussed in Section 4.3. 3 The SFP is designed to maintain leaktight integrity. ToL6 Kin 3lC ensure such l integrity, the pool is lined with stainless steel plate, and plate welds are backed with channels to detect and locate leakage. Leakage entering these channels is directed to the Liquid Waste Processing System (LWPS) via the FHB sump. Should a leak be detected, either by a low. level alarm (setpoint: 6 in. below normal water level) or by the fuel pool liner channel leak detection method, the operator would initiate makeup to the spent fuel pool. Makeup i

capability is provided by permanently installed connections to: (1) the Demineralized Water System (DWS), (2) the Reactor Makeup Water System (RMWS),

and (3) the refueling water storage tank (RWST) in the Emergency Core Cooling System (ECCS).

A complete loss of SFP cooling is not considered a credible event since the components involved are des'igned to SC 3 seismic Category I requirements and could be powered from redundant Engineered Safety Features (ESF) power supplies. Further, G e systems providing cooling are redundant. Therefore, no single failure would result in a complete loss of fuel pool cooling'. For a more detailed discussion of SFP cooling, refer to Section 9.1.3.

9.1.3 Spent Fuel Pool Cooling and Cleanup System The Spent Fuel ~ Pool Cooling and Cleanup System (SFPCCS) is designed to remove the decay heat generated by spent fuel assemblies stored in the SFP and/or the in-Containment storage area. A second function of the system is to maintain visual clarity and purity of the spent fuel cooling water and the refueling j water.

9.1.3.1 Desien Bases. The SFPCCS design heat loads are given in Table 9.1-1. System capabilities to withstand natural phenomena and piping rupture are addressed in Chapter 3. The spent fuel pool cooling portions of the SFPCCS are designed to seismic Category I requirements, and are located in the MLB, a seismic Cate5ory I building. The spent fuel pool water purification 9.1-5 Revision 3 1

STPEGS UFSAR REFEREMCES (Continued) f Section 4.3:

4.3-28 Suich, J. E. and Honeck, H. C.,

"The HAMMER System,, Heterogeneous l Analysis by Multigroup Methods of Exponential and Reactors", DP-1064, January 1967.

{

4.3-29 Flatt, H. P. and Buller, D. C., " AIM-5, A Multigroup, One Dimensional Diffusion Equation Code", NAA-SR-4694, March 1960.

4.3-30 Moore, J. S.,

l

" Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods", WCAP-9000-L, Revision 1 (Proprietary), July 1969 and WCAP-7806, December 1971.

4.3-31 Le ame r, R. D . , e t al . , "PUO,-UO Fueled Critical Experiments",

2 WCAP-3726-1, July 1967.

4.3-32 Nodvik, R.J., 1 "Saxton Core II Fuel Performcace Evaluation", WCAP-3385-56, Part II, " Evaluation of Mass Spectrometric and Radiochemical July 1970. Analyses of Irradiated Saxton Plutonium Fuel",

4.3-33 Camden, T.M.,

l et al., "PALADON-Westinghouse Nodal Computer Code,"

WCAP-9485A (Proprietary) and WCAP 9486A (Non-Proprietary),

December 1979, and Supplement 1, September 1981.

4.3-34 Liu, Y.S.,

i et al., "ANC: A Westinghouse Advanced Nodal Code",

WCAP-10965-A (Proprietary) and WCAP-10966-A (Non-proprietary), September 1986.

4.3-35 Green, N.M. "AMPX: A Modular Code System for Generating Co ed Mult. @ p Net.<<on-Gamma Libraries from ENDFIB,"

ORNL/rM-3706, March 1976.

PELETEE 4.3- 6 Petrie, L.M. and Landers, N.F., " KENO Va An Improved Monte Carlo Criticality Program Wit pergrouping,"

NUREG/CR-0200, December .

4.3-37 Fecreau, M.W. et al, " iticality Analysis of the South Texas Units 1 & pent Fuel Racks", attachment to Letter ST-U -1101, dated May 8, 1992 from R.C. Cobb, tinghouse Commercial Nuclear Fuel Division, to D.F.

Hoppes, Houston Lighting and Power.

4.3-38 Nguyen, T. Q.,

et. al., " Qualification of the PHOENIX-P/ANC Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A, June 1988.

4.3-39 Ford, W. E., et. al., "CSRL-V:

Processed ENDF/B-V 227 Neutron Group and Pointwise Cross Section Libraries for Criticality Safety, Reactor and Shielding Studies," NUREG/CR-2306, ORNL/CSDTM-160 (1982).

4.3-49 Revision 5 l

STPEGS UFSAR REFERENCES (Continued)

Section 4.3:

4.3-40 Fecteau, M. W., et. al., " Criticality Analysis of the South Texas Units 1 & 2 Fresh and In-Containment Fuel Storage Racks",

Westinghouse Commercial Nuclear Fuel Division, July 1992.

Attachment to Letter ST-UB-HL-1132, R. C. Cobb, WCNFD, to D. F. Hoppes, HL&P, dated July 15, 1992.

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.3-50 Revision 5