ML20134P165

Forwards Revised Pages to NSPNAD-8412, Safety Evaluation for RCC Guide Thimble Plug Removal. Removal of Thimble Plugs Will Drop Differential Pressure Across Core & Reduce Liftoff Forces Acting on Fuel

ML20134P165
Person / Time
Site:	Prairie Island
Issue date:	08/30/1985
From:	Musolf D NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
TAC-57544, TAC-57545, NUDOCS 8509060161
Download: ML20134P165 (9)

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Northern States Power Company 414 Nu:ollet Mail Minneapchs. Minnesota 55401 August 30, 1985 Temne st2> 330-ssoo Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Additional Information Concerning RCC Guide Thimble Plust Removal This letter and the attached documents are being submitted in response to NRC Staff comments from Mr L Phillips and Mr II Balukjun concerning our April 22,1985 submittal, entitled " Safety Evaluation for RCC Guide Thimble Plug Removal."

Attachment A contains revised pages to our safety evaluation (NSPNAD-8112) submitted with the letter referenced above. Contained in the revised pages for Section 5.1 is a discussion of the liftoff forces for both Exxon and Westinghouse fuel assemblies. The removal of thimble plugs will make the core less resistive to flow. The resulting flow increase is accompanied by a decrease in pump differential pressure which is characteristic of centrifugal pumps. Since the resistance of the rest of the system, i.e. piping and steam generators, remain the same, the pressure drop across the rest of the system will increase with the increasing flow.

The differential pressure across the pump equals the sum of the pressure drop across the rest of the system plus the pressure drop across the core.

with the pump differential pressure decreasing and the pressure drop across the rest of the system increasing, it is clear that the pressure drop across the core will decrease. Therefore, the removal of the thimble plugs will drop the differential pressure across the core and reduce the liftoff forces acting on the fuel, decreasing the possibility of fuel liftoff.

Please contact us if you have questions concerning this issue.

bW.

David Musolf Manager - Nuclear Suppo Services DMM/TMP/tp Attachment A. Revised pages for NSPNAD-8412 c: Regional Administrator-III, NRC NRR Project Manager, NRC Resident Inspector, NRC G Charnoff k 0509060161 050830 I\

PDR ADOCK 05000202 p PDR I

ATTACHMENT NORTHERN STATES POWER COMPANY Revised Pages to NSPNAD-8412 Old Page New Page 6 6 16 16

-new page- 18A 22 22 25 25

-new page- 25A 29 29

LIST OF TABLES l

P. ass i

2.1 Bypass Flow Calculations 13 l 3.1 Prairie Island Thermal Hydraulic Reference Conditions 17 3.2 Slow Rod Withdrawal Transient and Thermal Margin Results 18 3.3 Safety Limit Curve Results i 18A 4.1 Summary of Plant Transient Analysis Results 23 4.2 Parameter Values Used in Full Power Transient Analysis 24 5.1 RCS Flow Rate Measurement and Instrument Uncertainties 25A

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3.6 Safety Limit Curves The safety limit curves for Prairie Island, given in Technical Specifications Section 2.1, along with the associated overpower and overtemperature AT setpoints were calculated using FaH=1.55 and FQ=2.51 (Reference 5). Tech Spec Figure 2.1-1 shows the currently accepted safety limit curves.These are calculated as the loci of points where MONBR=1.3. A spectrum of bounding points have been renalyzed using the accepted NSPNAD methods (NSPNAD-8102P Rev.2) assuming FAH=1.55, FQ=2.32 and Wgyp=6.0%. Table 3 shows these results. Since all points are greater than 1.3, the current safety curves (T.S. Figure 2.1-1) bound operation with the thimble plugs removed. These values bound PI 1 Cycles 9 and 10 and PI 2 Cycles 9 and 10.

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t W-._.__.-....__-. .

TABLE 3.3 Power (% Rated) Pressure (psia) Average Temp ('F) MONBR 92 2400 611 1.492 100 2400 605 1.470 120 2400 584 1.447 100 2250 598 1.455 120 2250 576 1.425 59 1700. 593 2.255 100 1700 573 1.449 120 1700 550 1.361 Page 18A of 34

4.3 LOCA-ECCS Analysis LOCA-ECCS analyses for both fuel types, i.e. Exxon TOPR00 and Westinghouse OFA, are currently being performed with 6.0% bypass flow.

These analyses will be in place for the Prairie -Island Unit 1 Cycle 11 L

startup. The thimble plugs will remain in the core until this time.

4.4 ATWS Analysis

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The current analysis (ref: Prairie Island Updated Safety Analysis Report) is based on similar, but not necessarily bounding, plant parameters.

The purpose of this analysis is to show that if the Prairie Island Units were modified with the Alternate Mitigating Systems Actuation Circuitry i

l (AMSAC), it would satisfactorily meet the proposed alternative 3 as l defined in Volume 3 of NUREG 0460, which requires " modifications to i

I reduce susceptibility to common mode electrical failures and to provide mitigation of most ATWS events." This issue is currently being reviewed by the NRC.

i Therefore no reanalysis or Technical Specification changes are required for removal cf the thimble plugs since there is no licensing requirement, at this time, to provide a plant specific bounding ATWS analysis.

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5.0 MECHANICAL AND STRUCTURAL ANALYSIS The fuel assembly mechanical design and the vessel internals structural design assume upper limit bounding flows. Removal of the thimble plugs decreases the active core flow and increases the total vessel flow due to the decreased bypass flow resistance.

1 5.1 Fuel Mechanical Design and Fuel Liftoff Forces Analysis Exxon and Westinghouse have calculated the liftoff flow for their l fuels to be 206,000 gpm or greater (Reference 11 and 12) for the Prairie Island plants.

The current best estimate RCS flow rates for Prairie Island Units 1 and 2 are 196,883 and 195,331 gpm respectively. These flows will increase 0.6% due to removal of tne thimble plugs (Section 2.2) and approximately 0.6% due to the lower resistance of the Westinghouse fuel (based or, a total Westinghouse core). Also a 2.0*.' measurement and instrument uncertainty must be applied (Table 5.1). The revised flow I rates are then 203,237 and 201,635 gpm for Prairie Island Units 1 and 2 respectively. Since these flow rates are bounded by the above analyses, the thimble plugs can safely be removed, 5.2 Reactor Vessel and Internals Mechanical and Vibration Analysis A review of the Westinghouse reactor vessei and internals mechanical and vibration design analysis shows that thimble plugs are not a consideration. In addition, original hot functional tests at Prairie Island were run at 140% rated flow (249,200 gpm) and showed no adverse effects. Therefore the design analysis will remain bounding after the thimble plugs are removed.

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TABLE 5.1 RCS FLOW RATE Measurement & Instrument Uncertainties 5

Instrument Leading edge flow meter LEFM absolute accuracy and repeatability .67%

Elbow Taps Absolute accuracy (from LEFM) .67%

Repeatability .15%

AP transmitter accuracy t .6%

i DVM Calibration Accuracy .6%

DVM Absolute Accuracy .6%

RMS Total Accuracy 1.94%*

2.0% will be used l

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9.0 References

1. " Prairie Island Nuclear Generating Plant Units 1 & 2, Updated Safety Analysis Report", Docket Numbers 50-282, 50-306.

2. " Reload Safety Evaluation Methods for Application to PI Units" NSPNAD-8102P, December 1982.

3. XN-NF-80-61, " Prairie Island Nuclear Plants TOPROD Safety Analysis Report", Revision 1, March 1981.

4. XN-75-32(P)(A), Supplements 1, 2, 3, 4, " Computational Procedure for Evaluating Fuel Rod Bowing" October 1983.

5. WCAP 8090, " Fuel Densification Prairie Island Nuclear Generating Plant Unit No. 1," March 1973.

6. " Prairie Island Units 1 Cycle 10 Final Reload Design Report (RSE)"

NSPNAD-8411P, October 1984.

7. " Prairie Island Unit 2 Cycle 9 Final Reload Design Report (RSE)"

NSPNAD-8404P Rev.2, May 1984.

8. XN-NF-83-38, " Prairie Island Units 1 and 2 Limiting Break LOCA/ECCS Analysis using EXEM/PWR" May 1983.

9. WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analysis.

10. " Flow of Fluids Through Valves, Fittings, and Pipe," Crane, Technical Paper No. 410.

11. Telecopy NLG:020:85, N. Garner (Exxon) to T.M. Parker (NSP) " Evaluation of Increased Coolant Flow at Prairie Island" August 26, 1985.

12.

Letter 85NS-G-023, R.T.Meyer (Westinghouse) to T.M. Parker (NSP)

" Release of Westinghouse Proprietary Information to NRC" August 22, 1985.

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