ML19332F701

LER 89-008-01:on 890910,ATWS Mitigating Actuation Circuitry Automatically Bypassed at About 42% Reactor Power.Caused by Less than Adequate Tech Spec Change.Power Descension to 38% initiated.W/891211 Ltr

ML19332F701
Person / Time
Site:	Point Beach
Issue date:	12/11/1989
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-89-158 LER-89-008, LER-89-8, VPNPD-89-648, NUDOCS 8912180150
Download: ML19332F701 (5)

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WISCONSIN N v' 1Electnc POWER COMPANY S I W McNgm Po, Box 2046, Wwoukee. WI L3201 (414)221 2345.-  !

VPNPD-89-648- 10 CFR 50.73

. NRC-89-158' f December 11,-1989 M

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-Document Control Desk y Mail Station P1-137 p Washington, D. C. 20555  ;

c Gentlemen:

DOCKET'50-266-L LICENSEE EVENT REPORT 89-008-01' AMSAC NOT ENABLED AS REQUIRED BY-TECHNICAL SPECIFICATIONS

.2OINT BEACH NUCLEAR PLANT, UNIT 1 Enclosed-is Licensee Event Report 89-008-01 for Point Beach Nuclear Plant, Unit 1. .This report is provided in accordance with 10 CFR' 50.73 (a) (2) (1) (B) , "Any; operation or condition -'

prohibited-by the plant's Technical Specifications."

This report describes an event.in which the ATWS mitigating system: actuation circuitry (AMSAC) was automatically bypassed L at a power level greater than the 40% power level required by the TechnicalLSpecifications.

,If any;further information is required, please contact us.

'very truly yours,  !

p, fio g C. W. Fy Vice President Nuclear Power Enclosure Copies to NRC Regional Administrator, Region III NRC Resident Inspector 8912180150 hh66 PDR ADOCF PDC S

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,nei l l l On September 10, 1989, Point Beach Nuclear Plant Unit 1 was decreasing power from greater than 40% to less than 40%. At i somewhat less than 40% power, the ATWS mitigating actuation i

circuitry (AMSAC) is designed to be bypassed automatically based upon first-stage turbine pressure. Power, as indicated by first-stage pressure at this power level during power decreases, is less than reactor power. In fact, the AMSAC was automatically bypassed'at about 42% reactor power. Technical Specifications require the AMSAC system to be enabled at greater than 40% reactor power. Power was reduced to 38% and held until the bistable could be reset. The enable / disable bistable setpoint was reset to 30% to l ensure the operation of the system at less than 40% reactor power.

A revision to the Technical Specifications will be submitted to the NRC to clarify that turbine power of greater than 40% is the power level at which AMSAC is required.

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The Anticipated Transient Without Scram Mitigating System Actuation Circuitry (AMSAC) was installed per the requirements of 10 CFR 50.62

.under plant modification 85-213. Final acceptance testing of the modification was completed on May 8, 1989. The new circuit causes a turbine trip and starts auxiliary feedwater pumps based upon input from first stage turbine pressure and the position of main feed pump breaker or main feedwater regulating valve position. The circuitry is independent of existing reactor protection and emergency safeguards circuitry.

The P-20 bistable setpoint disables the AMSAC below 40% reactor power as derived from turbine first stage pressure. At 40% power and above, the AMSAC arms itself automatically. Status of the bistable is monitored continuously on main control board 1-C03.

The P-20 bistable alarm module is an " Action Pak" series (model)

AP1004-6016-R, manufactured by Action Instruments Inc. of San Diego, California (manufacturer identifier A051). The module features a 4-20 amp input, a normally deenergized double pull --

double throw (DPDT) relay, an adjustable setpoint and is powered by a 120 VAC.

EVENT DESCRIPTION:

On September 10, 1989, Unit 1 was operating at 49% of maximum power in an effort to conserve nuclear fuel. AT 0121 hours0.0014 days <br />0.0336 hours <br />2.000661e-4 weeks <br />4.60405e-5 months <br />, a

-further reduction in power was ordered to meet the low system load demands.- As power descended, the " Loss of Feedwater Turbine Trip Block," (trip function disabled), status light illuminated on main

-control board 1-C03. Reactor power (based on reactor thermal output) at the time of the alarm was 42%. Point Beach Technical Specification Table 15.3.5-3, Item 3.b, indicates this protective channel shall be enabled when reactor power is equal to or greater than 40% reactor power.

Unaware of the root cause of the problem, operators entered a three hour limiting condition for operation and immediately proceeded to reduce power to 38% to ensure Technical Specification compliance.

Instrument and Control personnel were summoned to troubleshoot a potential instrument fault. The AMSAC protective channel tested satisfactorily. The setpoint of the controlling bistable (P-20) for the AMSAC channel was reset to 30% reactor power based on turbine first stage impulse pressure. This change " armed" (enabled) the AMSAC protective channel at a more conservative setpoint.

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Ql l3 0F 4l Tun w, . - we % au.,on Investigations determined Technical Specification Table 15.3.5-3, Item 3.b, incorrectly based'the setpoint of the P-20 bistable at 40%

reactor power versus the intended "40% reactor power as derived-from turbine first stage impulse pressure." Normally, turbine first stage impulse pressure has a near linear program with respect to reactor power (based on reactor thermal output or RTO).

However, at this intermediate range of operation, RTO will deviate from turbine power from 2% to 8%. The reason for the deviation is based on thermal losses to the moisture separator reheaters and steam generator blowdown. These losses are not sensed by the turbine first stage impulse pressure.

The Technical Specification change to include the AMSAC channel was based upon reactor power due to ease of operator reference. The change request received proper reviews by Corporate Licensing, plant Regulatory Services, the Manager's Supervisory Staff and the i Offsite Review. committee. The Technical Specification which was impacted by the addition of AMSAC appeared on Table 15.3.5-3, l

Item 3.b. It is suspected that the words "as derived from turbine L first stage impulse pressure" were not added to the Technical Specification change because the- table format does not allow for detailed text additions or explanations. Discussions with plant staff and licensing personnel indicated there was a clear j understanding that the P-20 bistable setpoint was based on power as sensed by the turbine first stage pressure (turbine power). The

intent to base the setpoint on turbine power (rather than reactor L power) is documented in modification request 85-213 as well as Westinghouse Specification WCAP-10858P-A, Revision 1 -- AMSAC L Generic Design Package -- Logic 3 Functional Requirements.

L i CAUSES AND CORRECTIVE ACTIONS:

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The cause of the event was considered a less than adequate Technical Specification change. Short term corrective action included an immediate descension in power to 38% to ensure Technical Specification compliance. The setpoint of the P-20 bistable was reset to 30% reactor power as sensed by turbine first stage pressure. This ensures the AMSAC protective channel will

" arm" (enable) prior to exceeding 40% reactor thermal output. A l revision to the Technical Specifications will be submitted to the NRC to clarify that turbine power of greater than 40% is the power level at which AMSAC is required. This specification change request will be submitted to the NRC for approval prior to March 1, 1990.

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The health and safety of the general public, as well as plant employees, was not affected during this event. AMSAC is independent of design basis reactor protection and emergency safeguards instrumentation identified in the Final Safety Analysis Report. The protective functions of these systems were in no way degraded by the loss of the AMSAC channel because of the diverse design.

The automatic protective functions served by the reactor protection channels and the emergency safeguards channels were available at all times during the event. Also available was the capability of manual actuation of each of these channels. At no time were the reactor protection or emergency safeguards channels challenged during the event.

GENERIC IMPLICATIONS:

There are no industry generic concerns. This event was evaluated to be primarily a human performance issue. There have been two Licensee Event Reports at Point Beach which involved incorrect setpoints with the Technical Specifications (see Unit 1-LER's 82-09 and 82-25). In both cases, a Technical Specification change was made to correct the error.

REPORTABILITY:

This Licensee Event Report is provided pursuant to 10 CFR 50.73(a)(2)(i)(b), "Any operation prohibited by the plant's Technical Specifications."

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