ML022750295

Technical Specification Pages for Amendment No. 40, Irradiated Up to 2304 Tritium Producing Burnable Absborber Rods in the Reactor Core

ML022750295
Person / Time
Site:	Watts Bar
Issue date:	09/23/2002
From:	Padovan L NRC/NRR/DLPM/LPD2
To:	Scalice J Tennessee Valley Authority
References
TAC MB1884
Download: ML022750295 (17)

Text

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is fully open.

SR 3.5.1.2 Verify borated water volume in each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator is Ž 7630 gallons and 5 8000 gallons.

SR 3.5.1.3 Verify nitrogen cover pressure in each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator is Z 610 psig and

660 psig.

SR 3.5.1.4 Verify boron concentration in each 31 days accumulator is Ž 3500 ppm and 5 3800 ppm. AND

_____NOTE-----

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of

75 gallons, that is not the result of addition from the refueling water storage tank (continued)

Watts Bar-Unit 1 3.5-2 Amendment 7, 21, 40

RWSTs 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 ----------------NOTE--------------------

Only required to be performed when ambient air temperature is < 60OF or

> 105 0 F.

Verify RWST borated water temperature is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2 60OF and : 105 0 F.

SR 3.5.4.2 Verify RWST borated water volume is 7 days

Ž 370,000 gallons.

I SR 3.5.4.3 Verify RWST boron concentration is Z 3600 ppm and 5 3800 ppm.

7 days Watts Bar-Unit I 3.5-10 Amendment 7, 40

Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored shall be in accordance with Specification 4.3.1.1.

I APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 --------- NOTE-------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move Immediately the noncomplying fuel assembly.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the Prior to initial enrichment and burnup of the storing the fuel assembly is in accordance with fuel assembly.

Specification 4.3.1.1.

Watts Bar-Unit 1 3.7-31 Amendment 6, 40

THIS PAGE INTENTIONAL LEFT BLANK Watts Bar-Unit 1 3.7-32 Amendment 6, 4Q__

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site and Exclusion Area Boundaries The site and exclusion area boundaries shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2 (within the 3-mile circle).

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or Zirlo fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2 ) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications.

of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. For Unit 1, Watts Bar is authorized to place a maximum of 2304 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.

4.2.2 Control Rod Assemblies The reactor core shall contain 57 control rod assemblies.

The control material shall be boron carbide with silver indium cadmium tips as approved by the NRC.

(continued)

Watts Bar-Unit I 4.0-1 Amendment No. 8, 0L

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. keff : 0.95 if fully flooded with unborated water, which, includes an allowance for uncertainties as described in Sections 4.3.2.7 and 9.1 of the FSAR;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules.

d. Fuel assemblies with enrichments less than or equal to 3.80 weight percent U-235 are allowed unrestricted storage.

e. Fuel assemblies with initial enrichments greater than 3.80 weight percent and less than a maximum of 5 percent enrichment (nominally 4.95+/- 0.05 percent) may be stored in the spent fuel racks in one of four arrangements with specific limits as identified below:

1. Spent fuel assemblies may be stored in the racks without further restrictions provided the burnup of each assembly is in the acceptable domain identified in Figure 4.3-3, depending upon the specified initial enrichment.

2. New and spent fuel assemblies may be stored in a checkerboard arrangement of 2 new and 2 spent assemblies, provided that each spent fuel assembly has accumulated a minimum burnup in the acceptable domain identified in Figure 4.3-4.

3. New fuel assemblies may be stored in 4-cell arrays with I of the 4 cells remaining empty of fuel (i.e. containing only water or water with up to 75 percent by volume of non-fuel bearing material.

(continued)

Watts Bar-Unit I 4.0-2 Amendment No. 8, 40

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

4. New fuel assemblies with a minimum of 32 integral fuel burnable absorber (IFBA) rods may be stored without further restriction, provided the loading of ZrB2 in the coating of each IFBA rod is minimum of 1.25x (1.9625mg/in).

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material with no more than 75% of the water displaced.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum enrichment of 5.0 weight percent U-235 and shall be maintained with the arrangement of 120 storage locations shown in Figure 4.3-2;

b. keff

• 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR;

c. keff

• 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; and

d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

(continued)

Watts Bar-Unit 1 4.0-3 Amendment No. 6, A*0

Design Features 4.0 4.0' DESIGN FEATURES "4.3 Fuel Storage (continued) 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below Elevation 747 feet

- 1 1/2 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies in 24 flux trap rack modules.

(continued)

Watts Bar-Unit I 4.0-4 Amendment No. 6, 15 40

Design Features 4.0 I

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j Watts Bar-Unit 1 4.0-7 Amendment No. 6, 40

ECCS - Operating B 3.5.2 7.0 6.5 6.0 5.5 5.0 4.5 4.0 3.5 3.0 2.5 2.0 1.5 1;0 0.5 0.0 Initial Enrichment wt% U-235 Figure 4.3-3 Minimum Required Burnup for Unrestricted Storage of Spent Fuel of Various Initial Enrichments (continued)-

Watts Bar-Unit 1 4.0-9 Amendment No. 'AL-

ECCS - Operating B 3.5.2

! Bv ui

-1 ,

BBURNUP I

NACCEPTABLE I

DOMAIN I ,4 I

I I-UNACCEPTABLE 2 6'. / -BURNUP DOMAIN r 4 0-2.4 01/

/irrm

,:* .4* "4'i'"'iir a8 a

Initial Enrichment wt% U-235 Figure 4.3-4 Minimum Required Burnup for 2x2 Checkerboard Arrangement of 2 Spent Fuel Assemblies with 2 New Fuel Assemblies of 5% Enrichment (Maximum)

(continued)

Watts Bar-Unit I 4.0-10 Amendment No. _Z

ECCS - Operating B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS - Operating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a. Loss of coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater; and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold leg recirculation, and hot leg recirculation.- In the injection phase, water is taken from the refueling-water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation.

After approximately 3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, the ECCS flow is shifted to the hot leg recirculation phase to provide a backflush, which would reduce the boiling in the top of the core and any resulting boron precipitation.

The ECCS consists of three separate subsystems: centrifugal charging (high head), safety injection (SI) (intermediate head), and residual heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains.

The ECCS accumulators and the RWST are also part of the (continued)

Watts Bar-Unit 1 B 3.5-10 Amendment No. Q1__

RWST B 3.5.4 BASES APPLICABLE required volume is a small fraction of the available volume.

SAFETY ANALYSES The deliverable volume limit is set by the LOCA and (continued) containment analyses. For the RWST, the deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting.

The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as 27 seconds, with offsite power available, or 37 seconds without offsite power.

For a large break LOCA analysis, the minimum water volume limit of 370,000 gallons and the lower boron concentration I limit of 3600 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality.

large break LOCA is the limiting case since the safety The analysis assumes that all control rods are out of the-core.

The upper limit on boron concentration of 3800 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident.

(continued)

Watts Bar-Unit I B 3.5-26 Amendment 7, AQ0

Hydrogen Recombiners B 3.6.7 BASES APPLICABLE Hydrogen may accumulate in containment following a LOCA as a SAFETY result of:

ANALYSES I (continued) a. A metal steam reaction between the zirconium fuel rod cladding, TPBAR zirconium internals, and the reactor coolant;

b. Radiolytic decomposition of water in the Reactor Coolant System (RCS) and the containment sump;

c. Hydrogen in the RCS at the time of LOCA (i.e.,

hydrogen dissolved in the reactor coolant, hydrogen gas in pressurizer vapor space, and tritium contained in TPBARs); or

d. Corrosion of metals exposed to containment spray and Emergency Core Cooling System solutions.

To evaluate the potential for hydrogen accumulation in containment following a LOCA, the hydrogen generation as a function of time following the initiation of the accident is calculated. Conservative assumptions recommended by Reference 3 are used to maximize the amount of hydrogen calculated.

Based on the conservative assumptions used to calculate the hydrogen concentration versus time after a LOCA, the hydrogen concentration in the primary concainment would reach 4.0 v/o in about 3 days if no recombiner was functioning (Ref. 5). Initiating the hydrogen recombiners within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a DBA will maintain the hydrogen concentration in the primary containment below flammability limits.

The hydrogen recombiners are designed such that, with the conservatively calculated hydrogen generation rates discussed above, a single recombiner is capable of limiting the peak hydrogen concentration in containment to less than 4.0 v/o (Ref. 4).

The hydrogen recombiners satisfy Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit I B 3.6-44 Amendment No. 4-0

Spent Fuel Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel pool contains flux trap rack modules with 1386 storage positions and are designed to accommodate fuel I

with enrichment as high as 3.8 weight percent U-235 without restrictions. Storage of fuel assemblies with enrichment between 3.8 and 5.0 weight percent requires either fuel burnup in accordance with Specification 4.3.1.1 or placement in storage locations which have face adjacent storage cells I

containing either water or fuel assemblies with accumulated burnup of at least 20.0 MWD/KgU in accordance with I

Specification 4.3.1.1.

The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keg of 0.95 be evaluated in the absence of soluble boron. Hence, the design is based on the use of unborated water, which maintains the storage racks in a subcritical condition during normal operation with the racks fully loaded. The double contingency principle discussed in ANSI N-16.1-1975, and the April 1978 NRC letter (Reference 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, an (continued)

Watts Bar-Unit 1 B 3.72-7S Amendment 6, 4)

Spent Fuel Assembly Storage B 3.7.15 BASES -

BACKGROUND abnormal scenario could be associated with the improper (continued) loading of a relatively high enrichment, low exposure fuel assembly. This could potentially increase the criticality of the storage racks. To mitigate these postulated criticality-related events, boron is dissolved in the pool water. Safe operation of the spent fuel storage design with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with the accompanying LCO. Prior to movement of an assembly in the pool, it is necessary to perform SR 3.9.9.1.

APPLICABLE The hypothetical events can only take place during or as a SAFETY result of the movement of an assembly. For these ANALYSES occurrences, the presence of soluble boron in the spent fuel storage pool, (controlled by LCO 3.9.9, "Spent Fuel Pool Boron Concentration,") prevents criticality in the storage racks. By closely controlling the movement of each-assembly and by checking the location of each assembly after I movement, the time period for potential occurrences may be limited to a small fraction of the total operating time.

During the remaining time-period with no potential for such events, the operation may be under the auspices of the accompanying LCO.

The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of the NRC Policy Statement.

The restrictions on the placement of fuel assemblies within the spent fuel pool in accordance'with Specification 4.3.1.1 in the accompanying LCO, ensures the kff will always remain 50.95, assuming the pool to be flooded with unborated water.

This LCO applies whenever any fuel assembly is stored in the spent fuel storage pool.

(continued)

Watts Bar-Unit I B 3.72-76 Amendment 6, 40

Spent Fuel Assembly Storage B 3.7.15 BASES (continued)

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

If unable to move irradiated fuel assemblies while in Mode 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in Mode 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies stored in the spent fuel storage pool is not in accordance with Specification 4.3.1.1, the immediate action is to initiate action to make the necessary fuel assembly movements to bring the configuration into compliance with Specification 4.3.1.1.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that'the initial enrichment and burnup of the fuel assembly is in accordance with Specification 4.3.1.1 in the accompanying LCO. I REFERENCES 1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,'1978, NRC letter (Section 1.2) and implied in the proposed-revision to' Regulatory Guide 1.13 (Section 1.4, Appendix A).

Watts Bar-Unit 1 B 3.7-77 Amendment 6, LQ0.