ML20126J260

Forwards Request for Addl Info Re thermal-hydraulics Section to Maintain Snupps FSAR Review Schedule.Responses Requested by 810527

ML20126J260
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	04/21/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Bryan J, Koester G KANSAS GAS & ELECTRIC CO., UNION ELECTRIC CO.
References
NUDOCS 8104240124
Download: ML20126J260 (5)

Text

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Mfa acg[0,,' . UNITED STATES i f

3[gy a o NUCLEAR REGULATORY COMMISSION

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\,*. **** / APR 211981 0M tj h j 3 Docket Nos.: STN 50-482

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N'RMy2 ons 3 Iggy and STN 50-483 9 A "

Mr. John K. Bryan Mr. Glenn L. Koester

• Vice President Vice President - Nuclear

. Union Electric Company Kansas Gas and- Electric Company 1901 Gratiot Street 201 North Market Street Post Office Box 149 Post Office Box 208 St. Louis, Missouri 63166 Wichita, Kansas 67201

Dear Gentlemen:

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Subject:

SNUPPS FSAR - Request for Additional Information As a result of our review of your application for operating licenses we find >

that we need additional information regarding the SNUPPS FSAR. The specific information required is as a result of the Thermal-Hydraulic Section r,f the Core Perfonnance Branch's review and is listed in the Enclosure.

To maintain our licensing review schedule for the SNUPPS FSAR, we wili need responses to the enclosed request by May 27, 1981. If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses so that we may review our schedule for any necessary changes.

Please contact Mr. Dromerick, SNUPPS Licensing Project Manager, if you desire any discussion or clarification of the enclosed report.

Sincerely,

'bP{.Sn w Robert L. Tedesco, Assistant Director for Licensing Division of Licensing

Enclosure:

As stated cc: See next page Tills DOCUMENT CONTAINS POOR QUAL.lTY PAGES e+.-,wa gy-g -mi

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Mr. J. K. Bryan-Mr. Glenn L. Koester Vice President '- Nuclear Vice President - Nuclear '

Union Electric Company

=P. O. Box 4149 Kansas Gas and ' Electric Company St. ' Louis , . Missouri 63166 201 North Market Street P. O. Box 208

~ Wichita, Kansas 67201 cc: . Gerald Charnoff, Esq.

Shaw, Pittman, Potts, Dr. Vern Starks-

Trowbridge & Madden - Route 1, Box 863 1800 M Street, N. W. Ketchikan, Alaska 99901 Washington, D. C. 20036' Mr. William Hansen Kansas City Power & Light Company V. S. Nuclear Regulatory Commission '

ATTN: Mr. D. T. McPhee Resident Inspectors Office Vice President - Production . RR #1 1330 Baltimore Avenue Steedman, Missouri 65077 ,

Kansas ..C.ity, Missouri 64101 Ms. Treva Hearn, Assistant General Counsel Mr.-Nicholas A. Petrick Missouri Public Service Commission Executive Director, SNUPPS - P. O. Box 360 1 5 Choke Cherry Road Jefferson City, Missouri 65102 Rockville, Maryland 20850 Jay Silberg, Esquire Mr. J. E. Birk Shaw, Pittman, Potts & Trowbridge Assistant to the General Counsel 1800 M Street, N. W.

Union Electric Company Washington, D. C. 20036 St. Louis, Missouri '63166- ~

Mr. D. F. Schnell Kansans for Sensible Energy Manager - Nuclear Engineering P. O. Box 3192 Union Electric Company Wichita, Kansas '67201 P. O. Box 149 Francis Blaufuse '

Westphalia, Kansas 66093 Ms. Mary Ellen Salava Route 1,. Box 56 Mr. Tom Vandel Burlington, Kansas 66839 Resident Inspector / Wolf Creek NPS c/o USNRC P. O. Box 1407 Mr. L. F. Drb1 Emporia, Kansas 66801 Missouri - Kansas Section .

American Nuclear Society

• 15114 Navaho Mr. Michael C. Keener Olathe, Kansas 66062 Wolf. Creek Project Director State Corporation Commission Ms. Wanda Christy State of- Kansas 515 N. 1st Street Fourth Floor, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612 Floyd Mathews, Esq.

Birch, Horton, Bittner & Monroe 1140 Connecticut Avenue. N. W.

Washington, D. C. 20036 l

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4 Reouest for Add'itional Information_' i SNUPPS - FSAR' 49f.0 THEMAL-HYD: AULICS SECTION, CORE PERFORMANCE BRANCH 492. 2 . - The effects'of fuel rod ' bowing must be included in the thermal- '

hydraulic design. The predicted extent of rod bow (gap closure) versus exposure and the effect of rod bowing on DNBR must be accressed. Use of the staff report " Revised Interim Safety Evalua-tien Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Lignt Water Reactors,' ' February- 16,1977, repre-sents an acceptably conservative treatment of rod bowing.

492.3' Operating experience on-two pressurized water reactors (not of the Westinghouse design) indicate that significant reduction in core flow rate can-occur over a relatively short period of time as a result of crud deposition on the fuel rods. In establishino the -

Technical Specifications for Callaway and Wolf Creek we wi1T require provisions to assure that the minimum design flow rates are not exceeded. Therefore, provide a description of the flow me sure ments capability for Callaway and Wolf Creek as well as a Sescription of .

the procedures to measure flow and the actions to be taken in the i event of an indication of lower than design flow.

492. 4 The NRC approval of the THINC-IV codh, for use in the thermal-hydraulic design, indicates that the pressure gradient at the core exit must be modeled. Provide a revised THINC-IV calculation  ;

at the steady state reactor design conditions including the modeling of the core exit radial pressure gradient. Provide the following specific information from that calculation:

l. minimum DNB ratio (value and location)

2. hot channel flow vs. axial position

3. hot channel enthalpy vs. axial position

4. hot channel quality vs. axial position 3

5. hot channel void fraction vs. axial position ..  !

6. the assumed core exit pressure gradient. l 100ROR8W1 .

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1 492-2 1

-l 4 92. 5 Insufficient.information has been provided to justify;the design pm'er level- of 2389 Mwt (7nt of full pnwer). during three-loop --

operation Temperature differences in .the active cold legs of-

fe'..' degrees could exist. during three-loop operation. Therefore a radial power tilt and an increase-in enthalpy rise factor could result. As a result, we request that a complete detailed description of the following items be provided: -

1. The method of determining the temperature distribution among ,

the cold legs and the associated radial power tilt; *

2. The method of accounting for differences (if any) in the l

=three-loop thereal-hydraulic design;

3. The instrumentation available and monitoring procedures during three-loop operation;-  ;

4 Tne UBR Technicei Specification and how it will be imple-  ;

mented for three-loop operation;

5. The reactor protective system setpoints related to DNBR protection and how they are generated;

6. The effects of anticipated operational occurrences on the cold leg temperature distributions and how this effect is included in the design. .

1 492. 6 Please state your intent regarding tne use ot the Westinghouse optimizeo fuel assembly in your plant. If the use of tnis aesign is being consiaered, provide a discussion of the status and senedule for any revisea suomittals.

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492. 7: Pitase stete your intent. regarding the use.'.of the Westinghouse

~"improveo inermal- t>esign Proceaure" oescrioeo in Wt,AP-8067, dated July.1975. . lf ycu intend to u:e thece metheds respenses tb ene .

followin;; questions will be required:

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(a) Provide a block diagram depicting sensor, process equipment, computer,-and readout devices for each parameter channel.- ,

used in the uncertainty analysis. Within each element of the block diagram,' identify the accuracy, drift s range, span,-operating limits and setpoints. Identify the overall accuracy.of_ each channel transmitter to final output and specify the minimum acceptable accuracy for use with

• the new precedure. Also identify the overall accuracy of the

. output value and maximum accuracy. requirements for each input chcnnel of this finci cutput device.

(b) . Discuss the method'(s) for incorporating environmental ef_fects (e.g. , noise, EMI) on instrument channels into the uncertsinty; analysis.

(c) Provide data to verify that the plant instruments will perform with a high degree of confidence, within their design accuracies. This infomation may be obtained from operating history of identical instruments installed in other plants. Tnis request pertains to the instruments  ;

affecting the uncertainties in the design procedure (as identified in question 1 above), the overtemperat' re aT trip, the high flow trip, the low pressure trip and the pump voltage trip. -

(d) Provide the ranges of applicability of sensitivity factors..

(e). Demonstrate that the linearity assumption of equation 3-8 in WCAP-8567 is valid when the WRB-1 correlation is used.

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