ML20207T383
| ML20207T383 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 03/19/1987 |
| From: | Capstick R VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Rooney V NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20207T384 | List: |
| References | |
| FVY-87-32, NUDOCS 8703230506 | |
| Download: ML20207T383 (6) | |
Text
r VERMONT YANKEE NUCLEAR POWER CORPORATION RD 5. Box 169. Ferry Road, Brattleboro. VT 05301
,,,,,,o ENGINEERING OFFICE p
1671 WORCESTER ROAD FRAMINGHAM, MASSACHUSETTS 01701 TELEPHONE 617-872-8100 March 19, 1987 FVY 87-32 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attention:
Office of Nuclear Reactor Regulation Mr. Vernon L. Rooney BWR Project Directorate No. 2 Division of BWR Licensing
References:
(a) License No. DPR-28 (Docket No. 50-271)
(b) Letter, VYNPC to USNRC, FVY 86-34, " Proposed Technical Specification Change for Spent and New Fuel Storage,"
dated April 25, 1986 (c) Letter, USNRC to VYNPC, NVY 86-217, dated October 22, 1986 (d) Letter, VYNPC to USNRC, FVY 86-107, dated November 24, 1986 (e) Letter, USNRC to VYNPC, NVY 87-17, dated February 4, 1987
Subject:
Response to Request for Additional Information - Proposed Change No. 133, Spent Fuel Pool Expansion
Dear Sir:
By letter, dated October 22, 1986 (Reference (c)), you requested that Vermont Yankee verify compliance with Standard Review Plan (SRP), Section 9.1.3, particularly with the specified uncertainty factor. At that time, the heat loads for the Spent Fuel Pool (SFP) were re-evaluated using the SRP and Branch Technical Position (BTP) 9-2, and the Vermont Yankee licensed full power of 1593 HWL for the following conditions:
1.
The SFP contains the maximum number of fuel assemblies (fuel pool full), including the 136 assemblies from one 18-month cycle discharge, 10 days after shutdown.
2.
The SFP contains the maximum number of fuel assemblies (fuel pool full), including a full core discharge (368 assemblies) with annual cycle reloads, 10 days after shutdown.
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l United States Nuclear Regulatory Commission March 119, 1987 Attention:
Mr. Vernon L. Rooney Page 2 i
The calculated heat load for condition 1 above was 8.43 MBTU/hr and for Condition 2, 17.0 MBTU/hr. As stated in our response to Question 11 of Reference (c), these increased heat loads result in Residual Heat Removal (RHR) augmented fuel pool cooling being necessary for a longer period of time
.in order - to obtain additional decay and subsequent heat reduction sufficient -
to allow the Fuel Pool Cooling System to maintain pool temperature below our Technical: Specification limit of 1500F.
Subsequently, additional information was requested concerning the thermal-hydraulic heat loads calculated by Vermont Yankee When compared to the assumptions made by the NRC. These assumptions caused the NRC generated heat loads to be greater than those calculated by Vermont Yankee. The NRC calculations conservatively assume design full-power (1665 MWt) operation for i
the full 18-month or 12-month cycle, i.e., no refueling outage. Vermont Yankee has calculated heat loads based on the design full-power and continuous operation assumptions of the NRC.
The resulting heat load calculated for i
Condition 1 is 9.1 MBTU/hr and for Condition 2, 18.3 MBTU/hr. This calculation assumed ten days' decay time after shutdown and 39,420 and 35,040 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of operating time (i.e., 100% capacity factor) for the 18-month and 12-month cycles, respectively. The same heat loads corresponding to 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay time after shutdown are 10.34 MBTU/hr and 21.63 MBTU/hr, respectively.
i Finally, Vermont Yankee has calculated a third set of heat loads, based on the design full-power (1665 MWt) assumption but using a conservative but more realistic Cumulative Capacity Factor (CCF).
In this analysis, the reactor operating times were evaluated assuming a CCF of 84% (95% for the 12-month cycle and 93% for the 18-month cycle, excluding the refueling i
outage), well in excess of Vermont Yankee's historical performance. The highest cumulative unit capacity factor ever achieved by Vermont Yankee was 72.45%.
The 84% CCF corresponds to 29,529 hours0.00612 days <br />0.147 hours <br />8.746693e-4 weeks <br />2.012845e-4 months <br /> of operation for the 18-month cycle and 29,366 hours0.00424 days <br />0.102 hours <br />6.051587e-4 weeks <br />1.39263e-4 months <br /> of' operation time for the 12-month cycle. The heat loads calculated using 1665 MWL and the 84% CCF for Vermont Yankee was 8.8 MBTU/hr for Condition 1 and 17.8 MBTU/hr for Condition 2.
For relatively short decay times or shutdown periods, e.g. less than 20 days, the SFP Cooling System is augmented by the RHR System.
RHR augmented fuel pool cooling is j
sufficient to maintain bulk pool temperatures less than 1500F for decay times up to and including 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. Therefore, the cooling system capability for freshly discharged fuel meets the SRP Section 9.1.3 requirement at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.
To demonstrate that local boiling would not occur anywhere in the pool, under the most adverse heat load condition (full core discharge into an otherwise full pool), a shutdown period of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> was assumed. This was l
reflected in the RETRAN analysis previously submitted (Reference (b)).
I Based on the results of the third set of heat load calculations described
'above (i.e., 1665 MWt, 84% CCF), the conclusions reported in References (b) and (d), with regard to the capability of the Spent Fuel pool Cooling System have not changed and are valid.
4 L-
n United States Nuclear Regulatory Commission March 19, 1987 Attention:
Mr. Vernon L. Rooney Page 3 Also enclosed is the information you requested concerning the Vermont Yankee fuel pool RETRAN model. Provided are forward and reverse loss coefficients and the corresponding flow area for each junction in the model.
Note that the fuel assemoly inlet, rack inlet geometry, and the lower three spacer grids are lumped and modeled in the lower junctions. The upper four spacer grids and fuel assembly exit geometry were lumped in the outlet junctions. The frictional pressure drop calculation considered smooth surfaces from the standard RETRAN correlations; however, the results demonstrate that frictional losses are negligible.
The leveling pads at the bottom of the racks were modeled explicitly. The leveling pads were modeled in Junctions 44, 51, 59, 64, and 71 of the attached figure and are reflected in the loss coefficient data for these junctions.
Additionally, we have also enclosed two copies of each of the following Vermont Yankee fuel storage pool piping drawings for referral during your review:
G-191218, Revision 7 G-191219 Revision 5 G-191277, Revision 5 Information requented pertaining to the Spent Fuel Pool Cooling System capability and system design temperature limit will be provided under separate cover in the near futuee. We trust that the enclosed is responsive to your request; however, should you have any questions or require further information regarding this matter, please contact this office.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION l
N.
J R. W. Caps ick Licensing Engineer RWC/dlb Enclosures i
n 4
- 4 Table 1 Junction Summary Flow Area Forms Losses Forward Reverse Junction No.
g 2) 1 0.0707 0.50 1.00 2-0.0707 1.630 1.439 3 -31 0.2872 0.26
'0.26
- Includes 44,51,59 0.1489 283.6140 321.0727-Itvaling pads' 64.& 71
. R1maining 42-70 0.1489 23.7719 24.6481 82-111 0.1489 12.792 12.292 i
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