ML20062H498

Proposed Tech Specs Re Spent Fuel Storage During Cycle 4 Reload

ML20062H498
Person / Time
Site:	Millstone
Issue date:	11/30/1990
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20062H495	List: B13641, Revised Application for Amend to License NPF-49,changing TS Re Spent Fuel Storage During Cycle 4 Reload ML20062H498
References
CON-FIN-B-13641 NUDOCS 9012050017
Download: ML20062H498 (10)

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l Millstone-Nuclear Power Station,tinit No. 3 'i

' Proposed Revision to Technical Specifications- i

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'4 ' PDR ADOCK 05000423 .'

November-1990.

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DESIGN FEATURES 5.6 FUEL STORAGE

[.RITICAllTY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

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A k ((ted water. equivalent to less than or equal to 0.95 when flooded with unb8

b. A nominal 10.35-inch center-to-center distance between fuel assemblies placed in the storage racks.

a c. Fuel assemblies stored in Region I of the spent fuel pool may have a maximum nominal fuel enrichment of up to 5.0 weight percent U .

_ Region I is designed to permit storage of fuel in a 3-out b4 array with the 4th storage location blocked as shown in Figure 3.9-2.

d. Fuel assemblies stored in Region 11 of the spent fuel pool may have

- a maximum nominal fuel enrichment of up to 5.0 weight percent, conditional upon compliance with Figure 3.9-1 to ensure that the

_ design burnup of the fuel has been sustained.

DRAINAGE I 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 45 feet.

_ CAPACITY 5.5.3 The spent fuel storage pool contains 756 storage locations of which a maximum of 100 locations will be blocked. 4 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

I MILLSTONE - UNIT 3 5-6 Amendment No. A9 L _

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i Attachment 2 i Millstone Nuclear Power Station, Unit No..3 .

Description and Evaluation of Proposed Change to' Technical Specification i h

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Attachment 2 B13641/Page 1 Millstone Nuclear Power Station, Unit No. 3 Description and rvaluation of Proposed Chance to T ghnical Specifications rescription of Chance Section 5.6, Fuel Storage Design Feature Section 5.6.1.1.e has been deleted. This section specified that actual plant analysis was performed on a Cycle 3 specific basis due to considerations on the spent fuel pool cooling and piping systems and cool structure. The beyond Cycle 3 thermal hydraulic analysis has now been completed on these structures and components for spent fuel with a maximum enrichment of 5.05 weight percent extrapolated out to the end of life and consequently this restriction is no longer necessary.

Evaluation A. Spent Fuel Pool Bulk Temperature Analysis l

The spent ' fuel pool . bulk temperature was calculated to account for spent 'uel with a maximum enrichment of 5.05 weight percent U-235 with the number of stored fuel assemblies i the spent fuel pool increased to the projected final capacity. In addition, the decay heat load program has modeled the actual fuel bundle transfer rate from the reactor (3 bundles /hr) and has revised the time to the start of fuel movement into the pool based upon refueling cutage experience.

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The decay heat loads were calculated for a number of pool operatitg i conditions at the end of pool life. These are:

1. Normal refueling--one refueling load core (1/2 core) transferred from the reactor vessel to the pool starting 11 days after shutdown.

2. Normal plant operation--one refueling load (1/2 core) in the pool after 25 days of decay time. This conservatively models the decay heat load in the spent fuel during normal plant operation (reactor cooling system operating, turbine on line).

3. End of cycle full core offload--one full core offload after one year of power operation. Fuel transfer from the reactor starts 11 days after shutdown.

4. Emergency- full core offload--a full core offload commencing 36 days after the start of the previously refueling. This corre-sponds to an " abnormal maximum heat load" as defined in Standard Review Plan (SRP) 9.1.3 and is used as the decay heat load for E

Attachment 2 B13641/Page 2 all full core offloads occurring less than one year after a previous refueling shutdown.

The decay heat load for the older spent fuel in the pool was modeled as one refueling batch one year old, one refueling batch at 3 years old and the remaining refueling batches at 18-month intervals for cases 1, 2, and 3 above. For case 4, the older spent fuel was modeled as one refueling batch 18 months old and the remaining refueling batches at 18-month intervals over the life of the pool. The analysis is conservative because the present fuel plan is evaluating up to 24 month operating cycles beginning with Cycle 5. The fuel burnup rates used were conststent with 24-month operating cycles which adds more conservatism to the decay heat load.

A summary of the pool temperatures using these heat loads under a variety of normal and accHent conditions is found in Table 1.

It should be noted that for Millstone Unit No. 3, a full core di: charge is considered an emergency event with respect to the design basis enclyses. The current Millstone Unit No. 3 analyses limit the allowed number of full core discharges to (6) for the (40) year life of the plant. Millstone Unit No. 3 has discharged the full core after both Cycle 1 and 2 operations and is currently planning on performing a full core discharge after Cycle 3. These full core discharges have been performed to provide greater flexibility during the refueling outages. If NNEC0 decides to continue this practice as a routine event beyond Cycle 3, then a reanalysis will be performed to classify this as a normal routine evolutim " f or to exceeding the current (6) full core discharge limitation.

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Attachment 2 B13641/Page 3 Table 1 Millstone Unit No. 3 End of Life Spent Fuel Pool Temperatures Results from Analysis (5.05% Enrichment)

Steady State Operation Heat load Maximum Cooling Water Plant Condition (BTV/hr.) Eool Temo Temoerature Normal Refueling 21.77 x 105 129.08'F 95'F (1/2 Core Offload)

Normal Operating 18.59 x 106 124.16'F 95'F End of Cycle Full 34.79 x 106 149.46'F 95'F Core Offload (more than 1 yr. of operation)

" Emergency" Refueling 35.05 x 106 149.86'F 95'F (1/2 Core + Full Core)

Less than 1 yr. of operation.

Loss of Fuel Pool Cooling. Peak Temperatures Following a Single Failure Normal Refueling 21.77 x 105 140'F II) 95'F l

Normal Operating' 18.59 x 106 140'F(2) 95'F End of Cycle Full 34.79 x 106 154.5'F(3) 95'F Core Off-load l

l Loss of Power Concurrent with a Design Basis Accident (4) l

! Normal Operating 18.59 x 106 145'F 95'F l

l (1) Loss of pool cooling for approx.1.9 hrs, with pool at initial temper-I ature of approximately 129'F.

i-(2) Loss of pool cooling for 4' hours with pool at initial temperature of approxi;nately 125'F.

(3) Loss r,f pool cooling for 1/2 hour with pool at initial temperature of apprcximately 150*F.

, (4) LOP with CDA, or loss of power following safety injection and recircu-lation modo. See FSAR Table 8.3-1. Pool cooling lost for four hours, i

Attachment 2 B13641/Page 4 l i

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8. Summary of Analysis All analyses were performed assuming only one of two trains of spent fuel pool cooling is in operation. l 1

For all normal refuelings, as defined in SRP 9.1.3, the spent fuel pool temperature is maintained below the SRP limit of 140'F. In the event of a single active failure, the alternative train of cooling can be initiated in sufficient time to keep the pool temperature below '

140'F.

For an emergency full core offload, the spent fuel pool bulk pool temperature is maintained below 150'F which is the maximum long term temperature limit per FSAR Section 9.1.3, Table 9.12. This limit is based on the limitations of the concrete structure at high tempera-tures. The predicted temperature is also below the no boiling criteria required by SRP 9.1.3. Single failures are not required to be assumed during this event per Standard Review Plan 9.1.3 During an end of cycle full core offload, the spent fuel pool bulk temperature is also maintained below 150*F. This meets the FSAR limits described above. The SRP does not have any temperature requirements for this event. Table 1 contains a postulated single active failure occurring during the end of cycle full core offload.

During this event, the spent fuel pool bulk temperature will rise to l

approximately 155'F until the alternate train of cooling is initiated.

With the other train of cooling on line, the pool temperature will decrease to less than 150*F.

! During a design basis-accident (containment depressurization actuation-l signal), coincident with a loss of power, cooling water to the spent

fuel pool coolers is lost for four hours. During this time, the pool I bulk water temperature rises to approximately 145'F and then decreases l when cooling is restored. This is within the maximum analyzed spent i _ fuel pool bulk water temperature for the system.

l The design temperature of the spent fuel pool cooling system compo-nents and piping is 200'F. None of the analyzed scenarios exceed this design temperature.

For purposes of the concrete and spent fuel pool liner design only, the spent fuel pool cooling is assumed to be lost during both a normal and emergency full core offload. The pool temperature is assumed to increase until it reaches 200'F. At 200*F, pool cooling is reiniti-ated and the cooldown time to 150'F was determined. This time-temperature transient is used in the qualification of the concrete and spent fuel pool liner. While there are no credible failure modes which will produce this transient, it conservatively envelopes all i

Attachment 2 813641/Page 5 credible failures. The impact of these transients is addressed in part D, below.

C. SFP Cooling System and Piping NNEC0 has reviewed the impact of this proposed license amendment on the SFP Cooling System and associated piping and has concluded that this change does not involve a significant hazards consideration.

The spent fuel pool cooling system design temperatures and the system piping stress temperature limits will not be exceeded under any operating scenarios. As such, this change does not increase the probability of occurrence or consequences of any accident previously evaluated.

The operation of equipment in the sper.c fuel pool cooling system has not been modified in any way by this change, and as such, it does not create the possibility for a new or different kind of accident than previously evaluated.

The maximum spent fuel pool fluid temperature predicted under any scenario is still significantly below the spent fuel pool cooling design temperature of 200'F. As such, it does not reduce the margin of safety.

D. SFP Structure Evaluation The spent fuel pool is located in the fuel building at Hillstone Unit No. 3. The building, approximately 112 feet by 92.5 feet, is support-ed on compacted fill and/or rock. The fuel building has a ground floor at grade elevation and a basement 13 feet below grade. The spent fuel pool portion of the building is reinforced concrete con-struction from approximately 24 feet below to 28 feet above grade.

The spent fuel areas _ are protected from tornado missiles by a rein-forced concrete superstructure. The spent fuel pool is "L" shaped, with the bottom of the pool approximately 13. feet below grade. The i floor is 8-foot thick reinforced concrete. The walls are 6-foot thick l reinforced concrete. . The walls and floor of the pool are lined with 1/4-inch thick stainless steel plate.

I The spent fuel pool was de:igned for the loads and load combinations specified in the " Millstone Nuclear Power Station-Unit 3 Final Safety Analysis Report," Section 3.8.4. The loading analysis was performed-l in two parts. The first part was a mechanical load analysis. The I mechanical loads included dead loads, live loads, hydrostatic loads, L hydrodynamic loads, and seismic loads. The second part was a thermal l

load analysis. The thermal loads included normal operating and acci-dent loads. The thermal and mechanical load analyses were combined i

Att chment 2 B13641/Page 6 and the design analysis was performed. The original analysis assumed the storage of 3.85 w/o U-235 fuel in the pool.

A calculation entitled "MP3 SFP Analysis for Temperature Profiles of 5.05 Enrichment Fuel" was prepared to address the effect of the change of fuel enrichment from 3.85 weight percent U-235 to a maximum fuel enrichment level of 5.05 weight percent U 235 on the spent fuel storage pool. The existing mechanical load analysis was determined not to be affected due to the fact that the total weight of the fuel assembly is the same regardless of fuel enrichment levels. The change from 3.85 weight percent U-235 to a maximum fuel enrichment level of 5.05 weight percent U-235 does result in revised normal operating and accident thermal loads. The above referenced calculation addressed both changes in steady state and transient temperature curves resulting from the change of fuel enrichment. The results of the calculation are summarized as follows:

o Both the maximum normal refueling and full core offload steady state temperature effects on the concrete structural elements for 5.05 weight percent U-235 fuel are less severe than that originally analyzed for effects of 3.85 weight percent U 235 fuel temperature transients, considering either normal heat load with loss of cooling or normal heat load, maximum To, with loss of cooling.

o The temperature transient cases associated with the 5.05 weight percent U-235 fuel are time dependent functions which happen at a quicker rate than the originally analyzed for 3.85 weight percent U-235 fuel transients. Because of this, the new transients impose less of a thermal gradient on the concrete structural elements and result in less concrete stresses and strains.

The original concrete analyses for 3.85 weight percent U-235 fuel bound the effects inat the 5.05 weight percent U-235 fuel would impose on the structure.

The above calculation also determined that the existing liner design is still valid for 5.05 weight percent fuel because of the following:

o The mechanical load analysis, as stated previously, remains the same, o The thermal load analysis for the 5.05 weight percent.U-235 fuel is bounded by the 3.85 weight percent U-235 fuel analysis. This is due to the fact that the liner analysis is governed by the maximum temperature that is applied. The maximum temperature that the liner is subjected to is the same for either fuel enrichment.

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In conclusion, the storage of 5.05 weight percent U-23R fuel will not adversely affect the spent fuel pool concrete and liner system and, i based on the above, does not involve a significant hazards consideration since it does not:

(1) Significantly increase the 3robability of cccurrence or consequences of an accident prev'ously evaluated; or, l

(2) create the possibility for a new or different kind of accident from any evaluated; or (3) Significantly reduce the margin of safety E. Conclusion NNECO has reviewed the proposed change in accordance with 10CFR50.92 and has concluded that it does not involve a Significant Hazards  ;

Consideration because this change will not:

1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The proposed change quali-fies the SFP Cooling System, piping ard pool structure out to end of pool life utilizing the original design basis acceptance criteria. Since SFP temperatures remain as before, there is no adverse impact . on the results of any previously analyzed acci-dent.

2. Create the possibility of a new or different kind of accident from any previously analyzed accident. Since there are no changes in the way the plant is operated or in the operation of ,

the equipment credited in the design basis accidents, the poten- t tial for an unanalyzed accident is not ' created. Also, no new failure modes are introduced.

3. Involve a significant reduction in the margin of safety. The proposed change qualifies the SFP cooling system, piping and pool structure through the end of the pool life. Since there is no impact on the consequences of any accident previously analyzed, there is no reduction in the margin of safety.

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