ML20057E894

Provides Response to Request for Addl Info Re License Amend Request 93-004,transmitted Via Util Re Reload Analysis

ML20057E894
Person / Time
Site:	Comanche Peak
Issue date:	09/24/1993
From:	William Cahill, Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-93334, NUDOCS 9310130341
Download: ML20057E894 (4)

Text

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. Log # TXX-93334 File # 10010 l 2 C 916(2.2)

TUELECTRIC  !

i wim.m J. c.hm. Jr. D D # M' E '

Grcup Vsee l' resident i

i U. S. Nuclear Regulatory Commission  !

Attn: Document Control Room '

Washington, DC 20555  ;

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) l DOCKET NOS. 50-445 AND 50-446 >

REQUEST FOR ADDITIONAL INFORMATION CONCERNING l LICENSE AMENDMENT REQUEST 93-004 - RELOAD ANALYSES l i

REF: 1. TU Electric letter logged TXX-93204, dated May 28, 1993, from W. J. Cahill Jr., to the NRC Gentlemen:

l By letter dated September 22, 1993, the NRC requested additional information  !

concerning TV Electric License Amendment Request 93-004 transmitted by '

Reference 1. The following information is provided in response to that i request. I i

NRC Question 1 i You have discussed the use of different co-resident fuel assembly )

designs in Reference 1 (page 6 of 21, Attachment 2). Please provide the l 1

reference for the method that has been used for the core reload with I mixed fuel for CPSES Unit 1, Cycle 4. Have all the provisions from'the reference been satisfied such as that required for the analysis for the  ;

effect of stress from seismic forces between the different fuel types '

(Siemens and Westinghouse) and the DNBR penalty factors required.for transition cores?

, 10 Electric Response:  !

A. The methods used for the calculation of the mixed core DNB penalty are the same as used for the DNB analyses described in TU Electric report "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for CPSES 4 Licensing Applications," RXE-89-002, June 1989, approved by the NRC via letter dated August 5, 1993. Using these methods, a full core model of the mixed core configuration was developed and used to assess the effects of the mixed core on the DNBR.

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-TXX-93334 i Page 2 of 4  !

B. The effects of t!.e mixed core on the large break LOCA analysis  !

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were evaluated in accordance with TU Electric report, "Large Break ;

Loss of Coolant Accident Analysis Methodology," RXE-90-007, ,

December,1990, approved by the NRC via letter dated April 26,- i 1993.

t C. Both mechanical and thermal-hydraulic compatibility between the existing Westinghouse (H) fuel assemblies and the co-resident .

Siemens Power Corporation (SPC) fuel assemblies are evaluated in ,

the Reload Safety Evaluation. A comparison between the design criteria used for the fuel assemblies and the components has been ,

performed. It has been confirmed that both SPC and H have '

performed evaluations which demonstrate that their respective fuel ,

assembly designs meet all applicable design criteria. In  !

addition, SPC has evaluated the interaction between the co-resident H fuel assemblies and their fuel assemblies and confirmed that all applicable design criteria are satisfied.

NRC Question 2 4 You have discussed meeting the minimum measured flow requirement in i 1

Technical Specification (TS) 3.2.5c in Reference 1 (page 6 of 21, -

Attachment 2). Will this reload incorporate low leakage core loading?

If so, this type of loading has resulted in reduced indicated RCS flow  ;

rates. Will this reduced indicated RCS flow be a problem for CPSES Unit -

1 Cycle 4? Please provide the total flow rates in gpm measured from i the calorimetric heat balance for the current cycles for Units 1 and 2.  !

, Also please provide the references that approved the 1.8% uncertainty e 1 for the flow measurement and the 0.5% for the effects of the lower ,

plenum flow anomaly. .

TU Electric Response:

The Cycle 4 core configuration is a " low leakage" core design, as were j the Cycle 2 and Cycle 3 core configurations. The reduced indicated RCS '

flow phenomenon seen in low leakage core designs is caused by hot leg  ;

4 temperature streaming. At CPSES, the N-16 based Transit Time Flow Meter ,

(TTFM) is used to perform the precision flow calorimetric measurement. i As such, the accuracy of the flow measure.r,ent is not affected by hot leg !

temperature streaming. Furthermore, the evaluations of the existing -

flow margins have been based on Cycle 3 operation, in which a low i leakage core configuration was used.  ;

i For CPSES-1, Cycle 3, the "as-measured" RCS flow rate was 413,127 gpm. t i

For CPSES-2, Cycle 1, the "as-measured" RCS flow rate was 418,993 gpm.  !

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TXX-93334 Page 3 of 4 The 1.8% uncertainty for the RCS flow measurement is incorporated into Technical Specification 3.2.5 and is based on uncertainty calculations originally performed for CPSES-1 by Westinghouse and for CPSES-2 by TV Electric. W methodology was licensed by TU Electric and used to confirm that the 1.8% allowance remains valid for CPSES-1.

References:

CPSES-1 Tech Specifications (through Amendment 14), and NUREG-0797, SSER 12, Page 4-1, 4-2.

The allowance for the lower plenum flow anomaly was obtained from WCAP-11528, "RCS Flow Anomaly Investigation Report," April 1988.

NRC Question 3 Please provide the reference for the approved method used for obtaining the overtemperature and overpower N-16 reactor trip setpoints for obtaining the total uncertainty as discussed in Reference 1 (page 12 of 21, Attachment 2).

TU Electric Response:

The W methodology (WCAP-12123) was used by W in the calculation of the Reactor Trip System (RTS) and the Engineered Safety Features Actuation System (ESFAS) setpoints for the CPSES-1 Technical Specifications. TU Electric applied this methodology in the calculation of the RTS and ESFAS setpoints for the CPSES-2 Technical Specifications. TV Electric used the same methodology for the revised overtemperature N-16 and overpower N-16 uncertainty calculations as was used for the CPSES-2 Technical Specifications which have been approved by the NRC.

References:

WCAP-12123 (reviewed by the NRC prior to Unit 1 OL), and NUREG-0797 SSER 22, Page 7-7.

NRC Question 4 Please explain the difference between how the power is calculated using the N-16 power indication and that from the calorimetric power l indication as discussed in Reference 1 (page 11 of 21, Attachment 2).

TU Electric Response:

l- The N-16 power indication, in the same manner as the Nuclear Instrumentation System (NIS) power indication is normalized to the core power determined through the daily power calorimetric measurement. The proposed change affects how uncertainties associated with the use of the

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TXX 93334 i Page 4 of 4 i i

panel front meter in the adjustment of the N-16 power indications are r considered. Currently, the t.ncertainty is treated in a deterministic f manner. The Technical Specifications require renorma~lization if the l absolute difference between the N-16 (or NIS,' power and the calorimetric  !

power is greater than 2% rated thermal power (RTP). The uncertainty i associated with the panel front meters ( 1.5% RTP) is subtracted from  !

the 2% RTP allowance, such that renormalizations are required if the -j difference is greater than 10.5% RTP. The proposed change would  ;

incorporate the panel front meter indication uncertainty into the i statistical combination of uncertainties; thus, the actual renormalization would only be required if the difference is greater than i 2% RTP. This same change was incorporated into the setpoint. .I calculations performed by CPSES-2 and approved by the NRC through  !

incorporation into the CPSES-2 Technical Specifications.

Should you have questions concerning this submittal, please contact Bob '

Dacko at (214) 812-8228. .;

L Sincerely, f i

William J. Cahill, Jr.

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By: N D. R. Woodlan Docket Licensing Manager BSD j c- Mr. J. L. Milhoan, Region IV .

Resident Inspectors, CPSES (2)

Mr. T. A. Bergman, NRR ,

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