ML20136F665

Informs of Continuation of Reviews Re Allowable Value & Trip Setpoint.Tech Spec Changes Submitted in .After Acceptable Margin Established Analyses Will Be Performed. Tech Spec Change Request Will Be Submitted by 860401

ML20136F665
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/31/1985
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H Office of Nuclear Reactor Regulation
References
P-85503, TAC-47416, NUDOCS 8601070461
Download: ML20136F665 (2)

Text

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Company of Colorado 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 December 31, 1985 Fort St. Vrain Unit No. 1 P-85503 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. H. N. Berkow, Project Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Technical Specification LC0 4.4.1; Circulator Trip-Programmed Feedwater Flow-Low

REFERENCE:

1) PSC Letter dated 06/21/85, Lee to Johnson (P-85214),

2) PSC Letter dated 09/30/85, Brey to Hunter (P-85337).

Dear Mr. Berkow:

By letter dated June 21, 1985 (Ref. 1), Public Service Company of Colorado (PSC) submitted proposed changes to sections of the Plant Protective System (PPS) Instrumentation Fort St. Vrain (FSV)

Technical Specifications. The basis for the Technical Specification change request was the reformatting to the Westinghouse Standard Technical Specifications with the application of the Instrument Society of America Standard (ISA) S67.04-1982, "Setpoints for Nuclear Safety-Related Instrumentation used in Nuclear Power Plants." This Standard serves as a basis for establishing instrumentation setpoints to account for instrument inaccuracies and drift.

PSC also informed the NRC that a circulator trip parameter, Programmed Feadwater Flow-Low, had not been included in the original FSV Technical Specifications and was under evaluation for inclusion into the propos3d Technical Specifications change request.

The purpose of this circulator trip parameter is to provide equipment protection for the steam generators. Based on the code allowable limits, the bimetallic weld in the Superheater I/Superheater II crossover section and the Superheater I outlet tubes WI AD - D. CRUTCHFIELD (LTR D D cs a

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are the most restrictive components. Tripping of the helium circulator (s) is required to prevent prolonged operation of the steam generators ~ at excessively high temperatures. The steam generator tube temperature is related to the ratio of feedwater flow and helium flow (circulator speed). The allowable limit for steam generator tube temperature, which is derived from the codes to meet the specified design life, is a function of time at temperature.

Application of the long term design lifetime at temperature limit for a Technical Specification parameter, to protect against short term or transient conditions, imposes too restrictive of an operating margin. After applying the ISA Standard to account for instrumentation errors and drift, the trip setpoint is too close to the normal operating range. A relatively small magnitude change in feedwater flow or circulator speed would result in a helium circulator trip and impose unneccessary transients on the plant (potential for runbacks, a loop shutdown or a scram).

To provide assurance for continued plant operation without unnecessary transients resulting from minor perturbations in circulator speed as programmed with feedwater flow PSC is continuing to analyze the ASME Codes for short term or faulted condition limits, as addressed in FSAR Section 4.2.4. Also, PSC is investigating the potential for utilizing other monitored parameters to provide the required protection. Establishing acceptable short term allowable limits will provide the required margin for plant operation yet provide protection against control circuit failures or other malfunctions creating excessive time at temperature conditions which could cause component degradation.

As soon as an acceptable margin can be established in terms of short. term allowable limits of time at temperature, we will have analyses performed (consistent with previously submitted methodology) for an Allowable Value and Trip Setpoint and finalize the proposed Technical Specifications submitted in Reference 1.

Following PORC/NFSC reviews and acceptance, we will submit the new Technical Specification to the NRC Staff. Our updated commitment from that stated in Reference 2 is to submit this Technical Specification change request by April 1, 1986.

Should you have any questions concerning this matter, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, h*

H. L. Brey, Manager Nuclear Licensing and Fuels Division HLB /JS:bb. l