B13730, Responds to 910110 Telcon Request for Addl Info Re Spent Fuel Storage,Per Amend 39 to License NPF-49 Restricting Limited Storage of Spent Fuel in Spent Fuel Pool

Responds to 910110 Telcon Request for Addl Info Re Spent Fuel Storage,Per Amend 39 to License NPF-49 Restricting Limited Storage of Spent Fuel in Spent Fuel Pool

ML20070D508
Person / Time
Site:	Millstone
Issue date:	02/22/1991
From:	Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B13730, TAC-72997, NUDOCS 9103010009
Download: ML20070D508 (8)

Text

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-...,.~.aw. P O DOX 270 HARTFORD, CONNECTICUT 061410270 k ' J C$ [,U.l, ' CC, (203) 665+5000 February 22, 1991 Docket No. 50-423 B13730 Re: Spent Fuel Storage U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

References:

1. D. H. Jaffe letter to E. J. Hroczka, " Issuance of Amend-ment (TACNo.72997),"datedAugust 29, 1989,

2. E. J. Hroczka letter to U.S. Nuclear Regulatory Commis-sion, " Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications," dated April 20, 1989.

3. E. J. Mroczka letter - to U.S. Nuclear Regulatory Commis-sion, " Millstone Nuclear Power Station, Unit No. 3, Proposed Revision to Technical Specifications, Cycle 4 Reload--Spent Fuel Storage," dated November 30, 1990.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Response to Request for Additional Information Spent Fuel Storaae 4

in a letter dated August 29, 1989 i ment No. 39 to Northeast Nuclear (Energy Company for (NNECO) Reference 1), the NRC the Millstone Nuclear Power Station, Unit No. 3. This amendment was issued in response to a license amendment request submitted by NNECO on April 20,1989 (Reference 2),

and included, among other things, a rostriction that-limited the storage of spent fuel in the-spent fuel pool-to that spent fuel generated through Cycle 3 operation. The reason for this restriction was that the thermal hydraulic -

evaluation performed on the spent fuel pool- (SFP) cooling systeu piping and pool structure, addressing the impact of the increased fuel enrichment uti-lized for Cycle 3, only considered the thermal stresses associated with the increase in heat load resulting from the storage of spent fuel through Cycle 3. Further analysis was required to qualify the SFP cooling system beyond Cycle 3. This analysis was completed and a license amendment request to remove the Cycle 3 restriction was submitted to the NRC Staff-in a letter i dated November 30, 1990 (Reference 3).

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U.S.' Nuclear Regulatory Commission i B13730/Page 2 February 22, 1991 Subsequently, in a telephone conversation on January 10, 1991, the NRC Staff requested that NNECO provide additional information in response to verbal Staff questions associated with our November 30, 1990, license amendment request. The purpose of this letter is to provide the NRC Staff with this requested information. This % formation is provided in Attachment No. 1.

We trust you will find this information satisfactory, and we remain available to answer any questions you may have.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY m

E. J.E rocz a '

SentofVicePre[sident cc: T. T. Martin, Region 1 Administrator D. H. Jaffe NRC Project Manager, Millstone Unit Nos. 1 and 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 l

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Docket No. 50 421 fl1H32 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Response to Request for Additional Informatior.

31 February 1991 a.

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. U.Sl Nuclear Regulatory Commission B13730/ Attachment 1/Page 1 i February 22, 1991 Millstone Nuclear Power Station, Unit No. 3 j Response to Reauest for Additional Information Ouestion:

Provide an explanation of the factors resulting in the differences in the calculated maximum temperatures of an emergency full core off-load between the original design values, the 1989 cycle specific values, and the 1990 reanal-ysis values.

- Respona:

Table 1 provides the calculated s)ent fuel pool temperatures from the 1987 Final Safety Analysis Report (FSAR) (original design values), the 1989 cycle-specific analysis, and the 1990 reanalysis. The calculated maximum expected

, spent fuel pool temperatures after an emergency full core off-load were 149',

163', and 149.9'F for the 1987 FSAR, the 1989 cycle specific analysis, and the 1990 reanalysis, respectively. The major contributors in the increase in the expected temperatures between the 1987 FSAR and the 1989 cycle-specific analysis were the increase in the spent fuel pool cooling heat exchanger

, cooling water inlet temperatures, the n 1. decay heat curves provided -by Westinghouse for spent fuel, and the increase.in the new fuel enrichment. _ No sensitivity analyses have been performed to determine the relative impacts of a change to the decay heat loads or cooling water temperature on the calcu-lated spent fuel pool temperature. However, a review of the desi. . calcula-tions for the 1987 FSAR and 1989 cycle-specific analyses revealed ch6 follow-ing:

o A l' increase in cooling water temperature results in a l' . increase- in pool temperature, o An increase in decay heat load results in a proportional-increase in the temperature rise across the heat exchanger, o Minor changet in analytical method account for the remaining difference.

I As an example, the emergency core off-load case temperature rose 14'F between the 1987 FSAR and 1989 analyses. Of that increase,11*F is directly attribut-able to the cooling water temperature.- The decay heat load used was approxi-mately the same. The remaining difference is due to minor changes in the calculated exchanger effectiveness. For the normal operating condition the i

. analyzed temperature dropped by -7.4*F. In this case the cooling water temper-ature rise was compensated for by a decrease in the predicted decay heat-load.

l A determination of the relative impacts of these variables on the 1990 reanal-ysis is not possible as all the factors were analyzed by Holtec International '

via a computer program. The analytical methods were different, and no sensi-i tivity studies were performed. Best available data and realistic- but i

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U.S. Nuclear Regulatory Commission -l B13730/ Attachment 1/Page 2 February 22, 1991 conservative assumptions were used in the analysis. The goal was to achieve reanalysis temperatures that closely match design limits utilizing only realistic assumptions.

The 1990 reanalysis was performed by Northeast Utilities Service Company (NUSCO) and Holtec International. This reanalysis projected the end-of-life worst caso decay heat load based on the projected maximum capacity of the spent fuel pool and projected realistic operating conditions. The 1987 FSAR ,

j and 1989 cycle specific analyses used the LMTD method of heat exchanger

modeling. The heat exchanger effectiveness P was calculated from the original heat exchanger data. sheet information and used to calculate other, off design, conditions. As the flow rates and fouling factors used were the design values, this was a reasonable method since the heat transfer coefficient would not change significantly.

The 1990 reanalysis was performed using a QA validated heat exchanger thermal rating code. ST4, HTRI,-Rev. 1, dated 1974. This program was used to calcu-late a new heat exchanger effectiveness P and was used in the reanalysis.

. This program models the actual construction of the heat exchanger and ooes not

rely on the original data sheet. Holtec's QA program was reviewed and l approved by Northeast Utilities, and all analyses were performed under their program.

The actuel performance of the heat exchangers is better than was predicted by the origind esign method. This improved reanalysis method resulted in new projected emertency full core off loed temperatures less than what was pre-dicted for tho 1989 cycle-specific analysis and closer to the originally expected design temperatures.

The following summary illustrates the differences and major contributors to

, the decay heat load calculations for each of those cases. discussed above.

These were-the factors that determined the final calculated decay heat -load input into the spent fuel pool.

A. 1987 FSAR i

4 The original FSAR-decay heat load figures were-develoaed by Westinghouse

using a worst-case mix of spent fuels-from Millstone !) nit No.1 (boiling a

water reactor), Millstone Unit No. 2 (Combustion Engineering), and Millstone Unit No.-3 -(Westinghouse). The original design of the Millstone Unit Fo. 3 -spent fuel pool accounted for the. storage o.f; spent fuel from the ot'1er Millstone Units.. Westinghouse determined that a 3001 containing solel; Millstone Unit No. 3 fuel would have a lower decay 1 eat.

load and was therefore bounded by the worst-case mix calculation.

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i U.S. Nuclear Regulatory Commission 4 B13730/ Attachment 1/Page 3

! February 22, 1991

.j 1. Total Fuel Assemblies: 1096 Millstone Unit No I assemblies, j

! 630 Millstone Unit No. 2 assemblies, l j 512 Millstone Unit No. 3 assemblies for  ;

_ decay heat load calculations. 1

2. Refueling Off-load Size: 1/3 cone (64 assemblies) Millstone Unit I j: No. 3 fuel.

3. Time Since Shutdown

Normal refuel: 132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> (5.5 days)i-- ,

i emergency core-off-load: 10 days for full core, 70 days for previous discharge.

1 year (nominal).

4. Operating Cy*,le lengthi I

) 5. Burnup: 3 years (1100 effective full power days).

l 6. Enrichment: 3.8 w/o U-235.

The calculated resulting heat loads (Q) were as follows:

i

1. Normal 1/3-core off-load 1

i 0 (132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> after shutdown) 2.204 x_107 Blu/ hour F - Q-(25daysaftershutdown,normaloperating)=1.48x107 Btu / hour.

j 2. Emergency full-core off-load i

Q (10 days after shutdown) 4.035 x-107 Btu / hour B. 1989 Cycle Soecific Analysig-i The decay heat loads were based on the actual fuel inventory in the spent i fuel pool during -this time frame. A normal 1/3 core' off-load - was . not

analyzed because it would be bounded by the planned full core off-load -

this cycle.

i 1. Total Fuel Assemblies: 353 maximum.

j 2. Refueling Off-load Size: Full core (193 assemblies).

l 3. - Time Since Shutdown: 10 days.

l 4. Operating Cycle.' Length: Actual._.

5. Burnup

Approximate- values for actual fuel inven-t tory.

1-

- 6. Enrichment: Varied from 2.9 to: 4.5 w/o LU 235 based on

, actual inventory. ,

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4 U.S.' Nuclear Regulatory Commission B13730/ Attachment 1/Page 4 February 22, 1991 The calculated resulting heat loads (Q) were as follows:

1. Normal 1/3-core off load Q (132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br /> after shutdown) = not analyzed Q (25 days after shutdown, normal operating) = 0.9423 x 107 Btu / hour

2. End of cycle full-core off-load (Cycle 3 refuel) 0 (10 days after shutdown) 3.145 x 107 Blu/ hour )

3. Cycle 3 cmergency full core off-load Q (10 days after shutdown) 3.895 x 107 Blu/ hour C. 1990 Reanalysis to End-of-Peol Life This reanalysis was performed by NVSCO and Holtec International. It ;

projected the end-of-pool life decay heat load based on the projected l maximum capacity of the spent fuel pool and projected realistic operating conditions. Additionally, a more accurate analytical method was utilized in determining spent fuel pool cooling heat exchanger performance.

1. Total Fuel Assemblies: 2169 assemblies (all Hillstone Unit No. 3 fuel),

2. Refueling Off load Size: 1/2 core (96 assemb % C.

3. Time Since Shutdown: Normal refueling (:

11 days after shutdc .

full core)--start Fuel transfer rate is 3 assemblies per hour. Emergency full-core off-load -start 11 days after shutdown for full core. Fuel transfer rate is 3 assemblies per hour. Decay of 36 days for previous discharge.

4. Operating Cycle Length: 18 months for decay time between cycles.

This is conservatively short.

5. Burnup: 4.5 years (24-month operating cycle).

6. Enrichment: 5.05 w/o U 235 (maximum),

The calculated resulting heat loads (Q) were as follows:

1. Normal 1/2 core off load Q (316 hours0.00366 days <br />0.0878 hours <br />5.224868e-4 weeks <br />1.20238e-4 months <br /> after shutdown) 2.177 x 107 Btu / hour Q_(25 days after shutdown, normal operating) --l=859 x 107 Btu / hour

U.S. Nuclear Regulatory Commission B13730/ Attachment 1/Page 5 February 22, 1991

2. End of cycle full core off-load Q (345 hours0.00399 days <br />0.0958 hours <br />5.704365e-4 weeks <br />1.312725e-4 months <br /> after shutdown) 3.479 x 107 Blu/ hour

3. Emergency full core off-load Q (343 hours0.00397 days <br />0.0953 hours <br />5.671296e-4 weeks <br />1.305115e-4 months <br /> after shutdown) 3.505 x 107 Btu / hour Table 1 Tempertlure comparisons 1987 FSAR 1989 Cycle Soecific 1990 Reanalysis Normal Refueling 125'F Not Analyzed; full- 129.08'F Core Off load Planned Normal Operating Il8.9'F* 111.5'F 124.16'F End of-Cycle full- Not Described 150'F 149.46'F tore Off-load Emergency Full- 149'F 163'F 149.86'F Core Off-load Normal Operation 140'F 140'F 140'F Design L.imit Maximum Long Term 150'F 150*F 150'F Temperature (Design)

Maximum Short Term 200'F 200'F 200'F Temperature (Design)

Cooling Water Temper- 89.4'F Normal 95'F (Design) 95'F(Design) ature (CCP System) Maximum 83.9'F Emergency Maximum

• Temperature not described in FSAR, but found in design calculations. Provided here for comparison purposes only.

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