ML18018A490

Forwards Response to Core Performance Branch Draft SER Open Item 27 Re Seismic/Loca Effects on Fuel Assemblies

ML18018A490
Person / Time
Site:	Harris
Issue date:	04/08/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0609, RTR-NUREG-609 LAP-83-96, NUDOCS 8304120613
Download: ML18018A490 (10)

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ACCESSION NBRe8304120613 DOC 83/04/08 NOTA ZED: NO DOCKET Harris Nuclear power Planti Unit ii Carolina FACIL:50.-400 Shearon 05000400 50~401 Shearon Harris Nuclear Power Plant~ Uni,t 2< Carolina 05000401 AUTH NAME

~ AUTHOR AFFILIATION MCDUFFIE<M.A, 'Carolina Power L light Co.

RECIP',NAME RECIPIENT AFFILIATION DENTONiH ~ RE Office of Nuclear Reactor Regulationi Director

SUBJECT:

Forwards response to Core Performance Branch d raf t SER Open Item 27 re sei,smic/LOCA effects on fuel assemblies.

D I'STR I BUT I ON CODE: B001S -COPIES RECEI VED 'TR ENCL SIZE ~

TITLE: Licensing 'Submittal: PSAR/FSAR Amdts L Related Correspondence NOTES:

RECIPIENT iCOPIES RECIPIENT COPIES IO CODE/NAME LTTR ENCL ID CODE/NAME LT'TR ENCL NRR/DL/ADL 0 NRR LB3 BC 1 0 NRR LB3 LA 0 KADAMBIg P 01 1 1 INTERNAL~ ELD/HOS 1 1 0 IE FILE 1 1 IE/DEPER/EPB 36 3 3 IE/OEPER/IRB 35 1 NRR/OE/AEAB -

0 NRR/DE/CEB 11 1 1 NRR/DE/EQB 13 2 2 NRR/DE/GB 28 2 2 NRA/DE/HGEB 30 1 1 NRR/OE/MEB 18 1 1 NRR/DE/MTEB 17 1 1 NRR/DE/QAB 21 1 NRH/OE/SAB 24 1 1 NRR/OHFS/HFEB40 NRA/DHFS/LQB 32 1 1 NRR/DL/SSPB 0 NRR/DSI/AEB 26 1 1 NRR/DS I/ASB 1 1 NRR/DSI/CP8 10 1 NRR/DSI/CSB 09 i NRR/DSI/ICSB 16 1 NRR/OS I/METB 12 1 NRR/OS I/PSB 19 1 1 NRR/DS I/RAB 22 1 1 NRR/DSI/RSB 2S 1 G F L 04 1 1 RGN2 3 3 RM/DDAMI/MI8 1 0 EXTERNALS ACRS 41 6 6 BNL(AMDTS ONLY) 1 1 DMB/DSS (AMDTS) 1 1 FEMA REP OI V 39 1 1 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 05 1 1 NTIS 1 1

TOTAL NUMBER OF COPIES REQUIRED

LTTR 51 ENCL

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Carolina Power & Light Company SERIAL: LAP-83-96 APR 08 1883 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS ~ 1 AND 2 DOCKET NOS ~ 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT OPEN ITEM RESPONSES CORE PERFORMANCE BRANCH

Dear Mr. Denton:

Carolina Power & Light Company (CP&L) hereby transmits one original and forty copies of one Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report Open Item response. This response is for the Core Performance Branch and is CP&L Open Item No. 27.

We will be providing responses to other Open Items in the Draft Safety Evaluation Report shortly.

Yours very truly, M. A. McDuffie Senior Vice President Engineering & Construction JDK/pgp (6504JDK) cct Mr. N. Prasad Kadambi (NRC) Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP) Dr. Phyllis Lotchin Mr. J. P. O'Reilly (NRC-RII) Ms. Patricia T. Newman Mr. Travis Payne (KUDZU) Mr. John D. Runkle Mr. Daniel F. Read (CHANGE/ELP) Dr. Richard D. Wilson Chapel Hill Public Library Mr. G. 0. Bright (ASLB)

Wake County Public Library Dr. J. H. Carpenter (ASLB)

Mr. J. L. Felley (ASLB)

/'304i20hi3 PDR 830408 ADOCK 05000400 PDR 411 Fayetteville Street o P. O. Box 1551 ~ Raleigh, N. C. 27602

OPEN ITEM 27 Show that the fuel assembly mechanical response to Seismic and LOCA forces meets the requirements of NUREG-0609 Appendix E (Section 4.2.3.3).

S"'Sl~iIC/LOCA EFriCTS Ots SHEAROll HARRIS FUEL ASS-hBLIES The fuel assembly responses resulting from the lat ral safe .shutdown earthquake, SSE and the most limiting main coolant pipe break accident, LOCA are analyzed using time history numerical integration techniques. Since the reactor vessel motions resulting from the transient loadings are asymmetric with respect to the geometrical center of the reaci.or core, the full fuel assembly core finite element model described in Refs. 1 and 2 is used to determine the fuel assembly deflections and grid impact forces.

The reactor core finite element model consisting of the maximum number of fuel assemblies across the core diameter was used to analyze the fuel assembly responses.

The Shearon Harris Units I and 2 plants have fifteen fuel assemblies arranged in a planar array with gaps to simulate the geometric clearance between the uel assemblies as w@l as .he clearance between the peripheral fuel assemblies and baffle plate.

The fuel assembly finite element model is constructed by preserving the essential dynamic properties such as the fuel assembly vibration frequencies, mode shapes, and mass distribution. The time history motion for the upper and lower core plates and the barrel at the upper core plate elevation are simultaneously applied to the simulated core model. The detailed discussions of the analytical procedure, the fuel assembly and core modeling, and the methodology are documented in Ref. l. The time history inputs representing the safe shutdown earthquake motion and the coolant pipe rupture transient were obtained from the time history analysis of the reactor vessel and internals finite element model.

Grid Analysis The . aximum grid impact forces for both the seismic and asymm tric LOCA accidents occur at the peripheral fuel assembly locations adjacent to the baffle wall.

The maximum grid impacl forces obtained =rom the safe shutdown earthq ake and the nozzle inlet break analyses were approximately 38 and 30 percent of the allowable grid strength, respectively.

ln order to comply with the requirements in the USNRC 4.2 standard review plan, the maximum gr-',d impac. responses obtained from the two transient analyses shall be co>>bined using the square-root-of-sum-of-squares (SRSS) method. The combined grid impact forces, are determined at al 1 the gri d el evati ons. The resulting maximum combined impact force for the Snearon Harris fuel assemblies was approximately 48 percent of the allowable grid strength which is determined experimentally based on the 95% confidence level on the true mean as taken from the distribution of measurements.

The effect of steam flashing on grid impact load was analyzed. The results indicated that the effect of steam flashing on fuel grid impact load is negligible.

Therefore, the 30 percent increase in the grid impact load specified in the Appendix A of SRP Section 4.2 to account for steam flashing effects was not included in the grid load calculation, Ref. 3.

Idion-Grid ComDonent Analyses The stresses induced in the various fuel assembly non-grid components are assessed based on ihe most limiting seismic and LOCA accident conditions. The fuel. assembly axial force resulting from the LOCA accident are the primary sources of stresses in the thimble guide tube and the fuel assembly nozzles.

The induced stresses in the fuel'ods result from the fuel assembly relative rdeflection during the seismic and LOCA accidents and are generally very small.

The combined seismic and LOCA induced stresses of the various fuel. assembly components is presented. in Table 1 and expressed as a percentage of the allowable limit. Consequen.ly, the fuel assembly components are structurally acceptable under the postulated accident design conditions for the Shearon Harris Units.

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TABLE 1 FUEL ASSEMBLY COMPONENT STRESSES (Percent of Allowable)

Uniform Stresses Combined Stresses Component (Direct/Membrane) (Membrane + Bending)

Thimble 78.5 64.2 Fuel Rod" 23.7. 20.4

.oo Nozzle Plate 6.0 Bottom Nozzle Plate 47.5 Bottom Nozzle Leg 7.7 8.2

• include primary operating stresses

- A negligible value

1. Beaumont, H.D., et.al, "Verification Testing and Analyses of the lixli Optimi'zed Fuel Assembly", WCAP 9401-P-A (Proprietary) and liCAP 9402-A (Non-proprietary), August 1982.

2. Gesinski, L.T. and Chiang, O., "Safety Analysis of the 17xl7 Fuel Assembly for Combined Seismic and Loss-of-Coolant Acci'den.," MCAP 8236 (Proprietary) and MCAP 8288 (Non-proprietary), December 1973.

3. Rubenstein, L.S. to Rahe, E.P., Letter dated 8/5/82.

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