ML20044B671

LER 93-002-00:on 930122,determined That Current SG LP Signal Block Reset Values Greater than Allowed Ts.Caused by Improper Design.Test Procedures Will Be Revised by 930917 to Specify Desired Value for Block function.W/930222 Ltr

ML20044B671
Person / Time
Site:	Fort Calhoun
Issue date:	02/22/1993
From:	Gates W, Chutima Taylor OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-002, LER-93-2, LIC-93-0049, LIC-93-49, NUDOCS 9303010164
Download: ML20044B671 (6)

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Omaha Public Power District P.O. Box 399 Hwy.75- North of Ft.Calhoun Fort Calhoun, NE 68023-0399 402/636-2000 February 22, 1993 LIC-93-0049 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station PI-137 Washington, DC 20555 i

Reference:

Docket No. 50-285  :

Gentlemen:

Subject:

Licensee Event Report 93-002 for the Fort Calhoun Station Please find attached Licensee Event Report 93-002 dated February 22, 1993.

This report is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B). If you should have any questions, please contact me.

Sincerely, M/, E &

W. G. Gates l Vice President l l

WGG/jrg j Attachment c: J. L. Milhoan, NRC Regional Administrator, Region IV S. D. Bloom, NRC Project Manager R. P. Mullikin, NRC Senior Resident Inspector INPO Records Center I

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9303010164 930222 45.5129 PDR S ADOCK 05000285 PDR

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Fort Calhoun Station Unit No. 1 05000285 1 OF 5 Tm.E n Inappropriate Steam Generator Low Pressure Signal Block Reset Values EVENT DATE (5) LER NUMBER (6) REPORT NUMBER (7) OTHER FACluTIES INVOLVED (B)

MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000 01 22 93 93 002 00 02 22 93

]0b"0 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REOUIREMENTS OF 10 CFR 8: (Check one or more) (11) 1 MODE (9) 20.402(b) 20.405(c) 50.73fa)(2)(iv) 73.71(b)

S ER 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

UNEL (10) 100 20 405(O(1)Ci) 50 36(e)(2) 50.73(a)(2)Mi) OTHER 20.405(a)(1)(iii) X 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify in Abstract 20.405(a)(1)(lv) 20.405(a)(1)(v) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(viii)(B) "(("

50.73(a)(2)(x)

UCENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBEst (mesvoe Area Code)

Calvin C. Taylor, Shift Technical Advisor (402) 533-6754 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM CouPONENT MANUFACTURER CAUSE O SYSTEM COMPONENT MANUFACTURER TO RS SUPPLEMENTAL REPORT EXPECTED (14) MCMH DAY YEAR EXPECTED YEB SUBMISSION (If yes. cornp6ete EXPECTED SUBM!SSION DATE) )( NO DATE (15)

ABSTRACT (Umit to 1400 spaces, i.e., approximately 15 singte-spaced typewritten hnes) (16)

During a review of calibration procedures for the Steam Generator Pressure Loops, it was determined that the current Steam Generator Low pressure Signal (SGLS) block reset values, for all four channels of both steam generators, were greater than that allowed by Technical Specification (TS) 2.14, Table 2-1. This TS allows SGLS to be bypassed (manually) below 550 psia, however, the block reset values were found to range from 562 to 566.5 psia. This condition has existed since the 1988-89 (Cycle 13) Refueling Outage. SGLS was not in a bypassed condition at the time of discovery of this event.

The root cause of this event has been determined to be improper design of the pressure indicating controllers.

An operability evaluation has been completed and procedures have been revised to ensure manual reset of SGLS prior to exceeding 550 psia. Additional corrective actions include revising surveillance tests and submitting a TS amendment request.

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Fort Calhoun Station Unit No. 1 05000285 2OF5  !

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i BACKGROUND j

'A main steam line break would result in an abnormally high rate of steam flow from t either steam generator, accompanied by a marked decrease in steam pressure. To protect against the adverse reactivity consequences of a main steam line break, a reactor trip ,

is initiated by low Steam Generator (SG) pressure. Four pressure transmitters on each  !

SG provide input to the reactor protective system to initiate a reactor trip if either i SG's pressure drops below the trip setpoints on two-out-of-four channels. The trip  !

setpoints are required to be greater than or equal to 500 psia. .

These SG pressure transmitters also provide input to Pressure Indicating Controllers '

(PICS). The PICS serve as inputs for two-out-of-four coincidence matrices (independent I of the reactor protective system) that generate a Steam Generator Low pressure Signal (SGLS). The two SGLS matrix outputs serve as inputs to two Steam Generator Isolation l Signal (SGIS) logic circuits. The setting limit for steam line isolation is greater 7 than or equal to 500 psia. ,

SGIS automatically isolates the SGs to minimize the blowdown of the SG(s) in the event [

of a main steam line break outside containment and to limit containment pressurization "

during a break inside containment. Two channels of SGIS are derived from a logical '

combination of a Containment Pressure High Signal (CPHS) or a SGLS because either  !

condition is symptomatic of a main steam line break. The combining logic for SGIS is arranged such that either channel of SGLS or CPHS will initiate SGIS signals, isolating both SGs. Isolation is accomplished by automatic closure of the main steam isolation i and bypass valves, both feedwater isolation valves, the backup feedwater isolation ,

valves, and both feedwater bypass valves.

A key-operated bypass switch allows SGLS to be manually blocked when SG pressure is less than the block permissive setpoint. This blocks a SGIS actuation due to a SGLS but does -

not inhibit a CPHS from actuating SGIS. This block capability accommodates a controlled plant cooldown. This SGLS block is automatically removed when pressure rises above a reset value. The pressure difference between the block permissive setpoint and the block reset value (reset being higher than permissive) is the reset span. '

Fort Calhoun Station Technical Specification 2.14, Table 2-1, " Engineered Safety  !

Features System Initiation Instrument Setting Limits", establishes the Low Steam  ;

Generator Pressure setting limit as greater than or equal to 500 psia for steam line isolation. Note (2), on page 2-64a of the Technical Specifications, is associated with  ;

the steam line isolation setting limit. This note states: "May be bypassed below 550  ;

psia and is automatically reinstated above 550 psia." The intent of this note is that  !

SGLS is to be active above 550 psia. This is to be ensured by the automatic reset

• function.

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NUMBER NUMBER J Fort Calhoun Station Unit No. 1 05000285 3 OF 5  ;

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EVENT DESCRIPTION  !

i Modification MR-FC-85-136 replaced the existing Sigma PICS with Dixson PICS during the i 1988-89(Cycle 13)RefuelingOutage. The Dixson PICS were to have a ten psi reset span for the block permissive function, and an adjustable block permissive setpoint.. A 10 psi reset span would result in a block reset value of 550 psia or below if the block permissive setpoint is 540 psia or below; a larger reset span would require the block t permissive setpoint to be correspondingly lower than 540 psia in order for the block  !

reset value to be maintained at 550 psia or below.  ;

On January 22, 1993 at 0800, while in Mode 1 at 100% power, it was determined that SGLS  !

block reset values were greater than 550 psia on all four channels of both SGs, with values ranging from 562 to 566.5 psia. This condition was discovered as a result of a review of calibration procedures for the Steam Generator Pressure Loops. This review

'had been initiated to evaluate procedural guidance on setting the block permissive j setpoints.  :

I With the block reset values between 562 and 566.5 psia, the SGLS had been allowed to be  !

in a bypassed condition at pressures greater than 550 psia during recent SG l pressurizations fellowing plant shutdowns. This was determined to be in conflict with i TS 2.14, Table 2-1, Note (2), and therefore reportable pursuant to 10 CFR 50.73(a)(2)(i)(B). SGLS was not in a bypassed condition at the time of discovery l

. of this event. i i

SAFETY ASSESSMENT  !

This event was found to have~no safety significance. An operability evaluation found the consequences of a main steam line break-(with SGLS bypassed between 550 and 566 l psia) to be bounded by the hot full power and hot zero power cases. The event did not  ;

affect the ability of the reactor protective system to perform its design function.  !

Also, the short amount of time spent with the SG pressure in the range of 550-566 psia {

results in a low probability of occurrence of a main steam line break in this condition. .

-Finally, Emergency Operating Procedure E0P-05, " Uncontrolled Heat Extraction", instructs  !

operatorstoensuretheappropriatevalvesareisolated(closed)intheeventtheSGIS is not present. l f

Several licensed Combustion Engineering plants have SGLS block reset limits which are j 100 psi greater than the trip setpoint limit. Applying similar criteria'to the Fort i Calhoun Station' limits would result in a block reset limit of 600 psia. j I

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NUMDER NUMBER Fort Calhoun Station Unit No. 1 05000285 4 OF 5 l 93 --

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i TBCT (W more opmos is requ6 red, use addh6onal copies at NRC Form 30GA} 07)

-CONCLUSIONS The root cause of this event has been determined to be an inadequate design of the PICS.

OPPD provided specifications for the PICS requiring a 10 psi tolerance for the block

, permissive reset span and an adjustable block permissive setpoint. The manufacturer  :

supplied PICS with non-adjustable block permissive setpoints and reset spans of  :

approximately 26 psi.  !

A contributing factor was inadequate calibration procedures, in that there were no-provisions for verifying the block reset value. Since the modification used inadequate i calibration procedures, post-modification testing was also deemed inadequate.  !

The Pressurizer Pressure Low Signal (PPLS) circuitry includes similar provisions for i manual bypass and automatic reset. The PPLS PICS were checked and found to have ,

i appropriate block reset values.  ;

CORRECTIVE ACTIONS The following corrective actions have been or will be taken:  :

1. A caution tag has been placed by the PICS, instructing control room operators to manually reset the block permissive prior to exceeding 550 psia.

Procedures that address verification of the reset function have been revised ,

to ensure manual reset prior to 550 psia.

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2. The supplier has been contacted regarding the discrepancy with the PICS. A  !

NUPIC Joint Audit of the supplier was conducted in January 1992. The Audit 1 established that the company is implementing an effective program which meets l the requirements of 10 CFR 50, Appendix B. )

3. Actions have been taken since installation of MR-FC-85-136 to improve the i post-modification testing program (reference LER 92-006-01). These actions  :

included a procedure change, training and a review of the post-modification '

testing specified for a sample of modifications. It was concluded that there was not a generic problem with post-modification testing.

4. Surveillance Tests IC-ST-MS-0026 through 0033 (procedures for calibration of '

SG pressure loops) will be revised by September 17, 1993, to specify a desired I value and tolerance range for the block reset function. ,

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TEXT CONTINUATION AND RECORDS MANAGEMENT BFWWCH (MNBB m4). U.S. NUCLEAR ,

REGIA.ATORY ODMMiBBON WASHINGTON DC 202Mio01. AND TO  ;

THE PAPERWORK REDUCTON PRalEC.T 015:>0104). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON,0C 20$03.

FADUfY NAME(1) DOOMET NUMBER R LER NUMBER R PAGE R SEQUENT AL FIEVlfAION I NUMBER NUMBER Fort Calhoun Station Unit No. 1 05000285 93 -- 002 -- 00 mr e . % .a. o .oa.n.i a,F.- a NaC rm ini i

- 5. A Technical Specification amendment request, to allow higher values for the  :

SGLS block permissive / block reset function, will be submitted by  :

June 15, 1993.

-PREVIOUS SIMILAR EVENTS l LER 91-005 discusses a previous event involving a Technical Specification violation associated with SGLS. This LER, however, was not associated with the SGLS block permissive / block reset function.

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