ST-HL-AE-1894, Forwards Proposed Draft Rev 2 to Tech Spec 6.9.1.6 Re Methodology for Boron Dilution Analysis,Per Discussion & Presentation During Wk of 870112

Forwards Proposed Draft Rev 2 to Tech Spec 6.9.1.6 Re Methodology for Boron Dilution Analysis,Per Discussion & Presentation During Wk of 870112

ML20209G802
Person / Time
Site:	South Texas
Issue date:	01/27/1987
From:	Wisenburg M HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ST-HL-AE-1894, NUDOCS 8702050464
Download: ML20209G802 (5)

Text

o s-The Light Company n-,,, uu,o i>- i>.a in,x im um, . ,_ ,,mi m3msmn January 27, 1987 ST-HL-AE-1894 File No.: G9.06 10CFR50.36 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Unit 1 Docket No. STN 50-498 Drnft Revision 2 Technieni Snecifications (Tech Snees)

Please refer to our January 13, 1987 letter which transmitted proposed Tech Spec changes based upon the results of the Boron Dilution Safety Analysis Methodology. During the week of January 12, 1987, a presentation of the proposed changes was made to Messrs. B. Perch and M. Dunnefeld of your Staff.

As a result of this discussion and presentation, it was identified that a reference to the methodology for the Boron Dilution Analysis would be included in Specification 6.9.1.6.

As such, the following discussion and the attached marked up copy of the proposed Specification 6.9.1.6 are provided for your information and use:

The analysis methodology employed for the South Texas Project FSAR boron dilution safety analysis is the same as that employed in the Comanche Peak FSAR boron dilution safety analysis. This methodology was presented to the NRC during the licensing process for Comanche Pea'.c and was approved for Comanche Peak as documented in NUREG-0797 j

(Comanche Peak 's Safety Evaluation Report) Section 15.2.3.1.

Variations in the approach taken in the analysis between Comanche Peak and South Texas are delineated in the subsequent paragraphs.

For Comanche Peak, the analysis is based on the solution of a differential equation with inputs of volume, dilution flowrate,

! boron worth and minimum plant shutdown margin to obtain the time

, from alarm to the loss of plant shutdown cargin. The solution technique for South Texas employs the same dif ferential equation with inputs of volume, dilution flowrate, boron worth and minimum 1 l time allowed (the acceptance criterion) from alarm to the loss of plant shutdown margin. The solution, at South Texas, to the same differential equation used at Comanche Peak is the minimum plant shutdown margin such that the acceptance criterion is satisfied.

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O Ilouston Lighting & Power Comimny ST-HL-AE-1894 File No.: G9.06 Page 2 In order to satisfy the acceptance criterion in Modes 3, 4 and 5  !

Comanche Peak utilizes a safety grade microprocessor which continuously monitors the neutron count rate, comparing it to prior values of the count rate. An automatic mitigation sequence is actuated to prevent a loss of plant shutdown margin following an alarm from the microprocessor. Since more time is required to mitigate a boron dilution event without the presence of an automatic system such as that employed at Comanche Peak, the Technical Specifications minimum allowable shutdown margin requirement may increase to meet the safety analysis acceptance criterion. South Texas satisfies the acceptance criterion in Modes 3, 4 and 5 via a set of curves defining the minimum shutdown margin versus boron concentration. Maintaining the plant shutdown margin at a value greater than that defined by the curves satisfics the acceptance criterion for the minimum time available prior to a loss of plant shutdown margin.

The analysis methodology as applied to South Texas Project is susunarized in Appendix D of the Final Report of the Probabilistic Boron Dilution Analysis submitted in September 1986. The above referenced curves defining regional areas for acceptable shutdown margin will be provided in the STP Cycle Specific Core Data Report as delineated in Specification 6.9.1.6.

If you should have any questions on this matter, please contact Ms F. A. White at (512) 972-7985.

}

M. R. hu Deputy roject Ma ager FAW/1ja 4

Attachment

ST-HL-AE-1894 Fils No.: C9.06 Ilouston Lighting & Power Company Page 3 CCI Regional Administrator, Region IV M.B. Lee /J. E. Malaski Nuclear Regulatory Commission City of Austin 611 Ryan Plaza Drive, Suite 1000 P.O. Box 1088 Arlington, TX 76011 Austin, TX 78767 N. Prasad Kadambi, Project Manager A. vonRosenberg U.S. Nuclear Regulatory Commission City Public Service Board 7920 Norfolk Avenue P.O. Box 1771 Bethesda, MD 20814 San Antonio, TI 78296 Robert L. Perch , Project Manager Brian E. Berwick, Esquire U.S. Nuclear Regulatory Commission Assistant Attorney General for 7920 Norfolk Avenue the State of Texas Bethesda, MD 20814 P.O. Box 12548, Capitol Station Austin, TX 78711 Dan R. Carpenter Senior Resident Inspector / Operations Lanny A. Sinkin c/o U.S. Nuclear Regulatory Christic Institute Commission 1324 North Capitol Street

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P.O. Box 910 Washington, D.C. 20002 Bay City. TX 77414

Oreste R. Pirfo, Esquire Claude E. Johnson Hearing Attorney Senior Resident Inspector /STP Office of the Executive Legal Director c/o U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, DC 20555 i

P.O. Box 910 Bay City TI 77414 Citizens for Equitable Utilities, Inc.

c/o Ms. Peggy Euchorn M.D. Schwarz , Jr. , Esquire Route 1, Box 1684 Baker & Botts Brazoria, TX 77422 One Shell Plaza Houston, TX 77002 Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission J.R. Newman, Esquire 1717 H Street Newman & Holtzinger, P.C. Washington, DC 20555 1615 L Street, N.W.

Washington, DC '20036 j T.V. Shockley/R. L. Range Central Power & L!ght Company

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P. O. Box 2121 Corpus Christi, TX 78403 Revised 1/2/87 l

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. ATTACHMENT

. si.HL AE /SW .

l PAQE / _OF1 INSERT C STP CYCLE SPECIFIC CORE DATA REPORT 6.9.1.6 The F y limits for RATED THERMAL POWER (F ) and the minimum SHUTDOWN MARGIN requirements based on boron dilution accident analysis shall be established for at least each reload core and shall be maintained available in the Control Room. The limits shall be established and implemented on a time scale consistent with normal procedural changes.

The analytical methods used to generate the F limits and the SHUTDOWN MARGIN requirements based on boron dilution accident analysis shall be those previously approved by the NRC*. If changes to these methods are deemed necessary, they will be evaluated in accordance with 10CFR50.59 and submitted to the NRC for review and approval prior to their use if the changes are determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.

A report containing the F,y limits for all core planes containing Bank "D" control rods and all unrodded core planes and the plot of predicted (F .P Rel vs al Core Height with the limit envelope and the minimum SHUTDOWN MARGIN requirements based on boron dilution accident analysis shall be provided to the NRC Document Control desk with copies to the Regional Administrator and the Resident Inspector within 30 days of their implementation.

INSERT D

• WCAP 8385, " Power Distribution Control and Load Follow j Procedures", WCAP 9272.A. " Westinghouse Reload Safety Evaluation Methodology" and (

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• ATTACHMEN ST.HL AE- if PAGE 10F6L g2 g STRATIVE CONTROLS DEC 2 41S86 O RADIAL PEAKING FACTOR LIMIT REPORT Sec.:I u sser c- RTP 6.9.1.6 F limits for RATED THERMAL POWER (Fxy ) shall be provided the NRC Regional inistrator with a copy to Director of Nuclear ctor Regulation, Attention: ief, Reactor Systems Drar.ch, DPL-A . . Nuclear Regulatory Commission, Was on, D. C. 20555, for al re planes containing Bank "D" control rods and all un ed core planes the plot of predicted (F P Rel) vs Axial Core Height with th . envelope at least 60 days prior to each cycle initial criticality unl otherwise oved by the Commission by letter. In addition, in the that the limit sho change requiring a

, new substantial or an am submittal to the Radial Pea Factor Limit Report, it will be 1tted 60 days prior to the date the limi uld become effective unie herwise approved by the Commission by letter. An formation RTP needed upport F x will be by request from the NRC and need not be inclu his report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator _of the Regional Office of the NRC'within the time period specified for each report.

I 6.10 RECORD RETENTION O- 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal-maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety;

c. All REPORTABLE ENENTS;

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;

e. Records of changes made to the procedures required by

Specification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and

h. Records of annual physical inventory of all sealed source material of record.

SOUTH TEXAS - UNIT 1 6-20 t

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