ML18153C902

LER 91-014-01:on 910724,determined That Scaling for Main Steam Flow Transmitters Incorrect,Resulting in Setpoints in Excess of Max Allowed Value of 110%.Caused by Use of Incorrect Methodology.Scaling Program underway.W/920219 Ltr

ML18153C902
Person / Time
Site:	Surry
Issue date:	02/19/1992
From:	Kansler M VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
91-501A, LER-91-014, LER-91-14, NUDOCS 9202210408
Download: ML18153C902 (6)

Text

ION SYSTEM*

REGULA~ INFORMATION*DISTRIBUTIO~YSTEM (RIDS)

ACCESSION NBR:9202210408 DOC.DATE: 92/02/19 NOTARIZED: *No DOCKET #

FACIL:50-281 Surry Power Station, Unit 2, Virginia Elect~ic & Powe 05000281

,. AUTH.NAME AUTHOR AFFILIATION KANSLER,M.R.- Virginia Power (Virginia Electric & Power Co.)

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-014-01:on 910724,determined that scaling for main ste~m flow transmitters incorrect,resulting -in setpoints in excess of max allowe~ value of 110%.Caused by use of incorrect methodolog~.Scaling program underway.W/920219 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR  ! ENCL . / SIZE:

TITLE: 50. 7 3/50. 9 Licensee Event Report ( LER), IncidentRpt, et_c__.,.,""----

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NOTES:lcy NMSS/IMSB/PM. 05000281 RECIPIENT COPIES *RECIPIENT. COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2_PD 1 1 BUCKLEY,B 1 1 INTERNAL: ACNW 2 2 ACRS 2 2

. AEOD/DOA 1 1 .AEOD/DSP/TPAB 1 - 1 AEOD/ROAB/DSP 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFBlO 1 1 NRR/DLPQ/LPEBlO 1 1 NRR/DOEA/OEAB 1 1 NRR/DRE;P/PRPBll 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1

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EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PQR 1 1 NSIC MURPHY,G.A 1 1 J NSIC POORE,W. 1 1 NUDOCS .FULL TXT 1 1 NOTES: 1 1 I

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I NOTE TO ALL "RIDS" RECIPIENTS:

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L PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATE YOUR-NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL 'I'EXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL 33

e e Virginia Electric and Power Company Surry Power Station

• P.O.Box315 Surry, Virginia 23883 Februatj'll, 199~

U.S. Nuclear Regulatory Commission Serial No.: 91-501A Document Control Desk Docket No.: 50-280 Washington, D. C. 20555 50-281 License No.: DPR-32 DPR-37

• Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following updated Licensee Event Report for Units 1 and 2.

REPORT NUMBER 91-014-01 This repo1-t has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed. by the Corporate Management Safety Review Committee.

Verytntly~

~ s l e1 r Stat~Manager Enclosure

- cc: Regional Administrator SuitE!2900 .

101 lv.larietta Street, NW Atlanta, Georgia 30323 9202210408 920219 PDR ADOCK 05000281 S PDR

e NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92

.MATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP-530), U.S. NUCLEAR

.. REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13150-0104), OFFICE I

OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACI.LITY NAME (1.1 DOCKET NUMBER (2) PAGET3J Surry Power Station, Unit 1 01s10101012 18 10 1 loF 014 TITLE 14) Main "Steam l:'l.OW .:::;etpoints in t:xcess or Technical. :::ipecirications Due to Flow Transmitter Scaling Inaccuracies Resulting From Incorrect Scaling Methodology EVENT DATE (!5) LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR /f) SEQUENTIAL NUMBER m REVlSION NUMBER MONTH DAY YEAR Surry, Unit 2 FACILITY NAMES DOCKET NUMBERISI 0 15 Io I o I o I 218 11 0 17 214 9 1 91 1 01 1 14

- 0 11 012 1 11 9 l2 0 1 s 1o IO I O I I I THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §:* (Check one or more of the following} (111 OPERATING MODE (9) N 20.402(b) 20.4051c) 50.73(o) (2)(iv) 73.71(b)

- 1-

. POWER LEVEL (10) Ii I o, 0

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LICENSEE CONTACT FOR THIS LER 112)

NAME TELEPHONE NUMBER M. R. *Kansler, Station Manager AREA CODE 810 I 4 315 I 71- 13 I l 18 14 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS* REPORT 113) 1:,,:, *>:*:*::,:, ,',',',' .;,,:c:,:c:::::::

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I I I I I I I .......... I I I I I I I SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAY YEAR EXPECTED n ~

SUBMISSION DATE 1151 YES (If yes, comp/,-is EXPECTED SUBMISSION DATE} NO I I I ABSTRACT (Limi'. ro 1400 spaces. i.e .. approximately fifreBn single-sf)8ce typewritten lines) (16)

~

On July 24, 1991, with Unit 1 and Unit 2 operating at 100% power, it was determined after analyzing recent special test data that the scaling for the Unit 2 Main Steam (MS) flow transmitters was incorrect. These flow transmitters provide input to certain Reactor Protection System (RPS) and Engineered Safety Features (ESF) actuation signals. It was determined that the scaling resulted in Unit 2 setpoints in excess of the maximum allowed value of 110% (at full load) of full steam flow specified in Technical Specification (TS) 3.7.D. On July 22, 1991, voltage biases were placed on the uncompensated Unit 2 steam flow signals to reduce the Hi Steam Flow ESF function actuation to a value below 110%. This event was caused by the. use of incorrect scaling methodology, employed in 1977, to rescale the MS flow transmitters to correct an observed difference in the MS and feedwater flow indications. Since the same incorrect scaling information was used in 1977 for Unit 1 ; similar voltage biases were introduced on Unit 1 on July 24, 1991. Following .an evaluation of special test data, the Unit 2 steam flow and feedwater *flow transmitters were respanned and the steam flow circuitry rescaled. Independent testing is being performed to enable these measures to be implemented for Unit 1. The event is being reported, pursuant to 10 CFR 50. 73(a) (2)(i)(B).

NRC Form 366 (6-89)

L___

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) . APPROVED 0MB. NO, 3150-0104 EXPIRES: 4/30/92 LICENSEE EV. REPORT (LERI ATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSlbN, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY.NAME (1) DOCKET NUMBER (2) PAGE (3).

LEA NUMBER (6)

Surry Power Station, Unit 1 TEXT (If more SfJIICtJ is rec,*uired, use additional NRC Form 3!i'6A 's) (17) 1.10 DESCRIPTION OF THE EVENT On July 24, 1991, with Unit 1 and Uni.t 2 operating at 100% power, it was determined after analyzing recent special test data that the scaling for the Unit 2 Main Steam (MS) flow* transmitters {EliS-SB,FIT} was incorrect. These flow transmitters provide input to certain* Reactor Protection System (RPS) {EIIS-JC} and Engineered Safety Features (ESF) actuation {EIIS-JE} signals. It was determined that the installed scaling factors resulted in setpoints in excess of the maximum allowed value of 1 f0% (at full load) of. full steam flow specified in Technical Specification (TS) 3.7.D (Table 3.7-4, Item 5).

The special test began on July 4, 1991 and was governed by procedure 2-ST-302. The objective of the test was to collect data* at various power levels to verify the manufacturer's calibration curves *for feedwater flow venturis {EIIS-SJ} 2-FW-FE-2476, 2-FW-FE-2486, and 2-FW-FE-2496, which were installed during the Unit 2 Cycle 1O refu~ling outage. Data collected in this test

. was used to perform final scaling calculations for the feedwater and steam flow transmitters. The data collected during the special test was discussed by telephone wi.th NRC staff members on ~uly 12, 1991. *

• Following data reduction and analysis, it was determined that the Unit 2 high steam flow setpoint for ESF actuation was non-conservative due to incorrect scaling. It was also determined that the steam flow setpoint for the steam flow/feedwater flow mismatch reactor trip was non-conservative due to incorrect density compensation and incorrect scaling. The nonconservatism in the steam flow/feedwater flow mismatch reactor trip was insufficient to exceed the value for reactor trip protective instrumentation settings in Technical Specification 2.3.A.3.c.

The event is being reported, pursuant to 10 CFR 50.73(a)(2)(i)(B), since this*

condition is prohibited by TS 3.7.D (Table 3.7-4, Item 5).

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS MS flow signaJs are used as an input to two of the plant's protective functions.

High steam line flow coincident with either low Reactor Coolant System (RCS)

{EIIS-AB} *average temperature (Tave) or low steam line pressure results in a signal to the ESF protection logic, initiating Safety Injection (SI) and MS line isolation signals, Steam fiow is also an input to the RPS logic, initiating a

.reactor trip, if the system senses a steam flow/feedwater flow mismatch condition coincident with low steam generator {EIIS-SG} water level in any steam generator.

NRC For,y, 366A (6-89)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) -APPROVED OMS.NO. 3150-0104 .

- * - EXPIRES: 4/30/92 .

LICENSEE EV. REPORT (LERI -ATEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD

• COMMENTS REGARDING.BURD.EN- ESTIMATE TO THE RECORDS TEXT CONTINUATION °AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (~) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

Surry Power Station, Unit 1 0 15 IO IO IO , 2 ,s IO 9 I 1 - o, 114 - 0 I 1 0 I3 OF O14 TEXT (ff more :,pa~ is mq,,ired, use tlddiriontll NRC Form 36tiA '*I (17)

The limiting steam line break analyses which take credit for the high steam flow channels are not sensitive to the steam flow setpoint. In the accident scenario, steam flow rapidly increases to a value well in excess of the setpoint. The Surry evaluation for small steam line breaks shows that break sizes in excess of 0.19 sq. -ft: per loop would exceed* the steam fiow setpoint. The limiting Surry break size is 1.4 sq. ft., well in excess of this value. Therefore, the limiting steam flow used in the accident analysis will exceed the setpoint by several hundred percent."

The recent review of scaling calculations indicated that the maximum error resu Its in a worst case steam flow of 115% vice 110% allowed by TS for generating the ESF signal. The Surry evaluation also. shows that, for smaller break sizes where the steam flow setpoint cannot be relied upon, adequate protection is provided by diverse channels (i.e.; high containment pressure, low RCS pressurizer {EIIS-PZR} pressure). --

The steam flow/feedwater flow mismatch reactor trip is a diverse, anticipatory trip which backs up the low-low steam generator level reactor protection -

function with one exception. The exception is a partial loss of feedwater due to failure of the steam generator level channel feeding the level. control system

{EIIS-LIC}. In this scenario, the two unaffected steam generators are adequate to remove the core heat until the reactor trip occurs on. low-low steam generator level in the affected steam generator, high RCS pressurizer level, or excessive difference between .RCS cold leg and hot leg temperatures (delta-T). In conjunction with the reactor trip, auxiliary feedwater is initiated on a low-low steam generator level signal prior to the onset of significant RCS heat-up. In -

addition, the nonconservatism in the steam flow/feedwater flow mismatch reactor trip was not sufficient to exceed the value for reactor trip protective instrumentation settings in -Technical Specification 2.3.A.3*.c.

Thus, the observed steam flow measurement uncertainties did not invalidate the acciderit analyses and the event has presented no adverse consequences to public -

health and safety.

3.0 CAUSE OF THE EVENT In 1977 the main steam flow transmitter spans were changed on both Unit 1 and Unit 2 using an inc9rrect scaling methodology. The respanning was undertaken to correct an observed difference in the values for steam flow and feedwater flow wherein the steam flow was reading higher than the feedwater flow. Based on

• evaluation of the special test 2-ST-302 data, it is now known that the difference in the steam flow and feed flow measurements observed in 1977 was due to an e*rror in the steam flow signal multiplier/divider scaling which caused an error in the steam flow indication ..

NRC Form 366A (6-89)

NRC F.ORM 366A' . U.S. NUCLEAR REGULATORY COMMISSION (6,89) . APPROVED 0MB NO. 3150-0104

- . . EXPIRES: 4/30/92

• LICENSEE EV * . REPORT (LERI - A T E D BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION *. AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-01'04). OFFICE OF MANAGEMENT,AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)

Surry Powsr Station, Unit 1 o 1s .Io Io Io 12 1s Io 9 I 1 _ o 1114 _ o 1 1 o I 4 oF o 14 TEXT /ff more spa .. is nxiuired, u"" additional NRC Form 366A's/ (17)

4. O IMMEDIATE CORRECTIVE ACTION($)

On July 22, 1991 for Unit 2 and July 24, 1991 for Unit 1, voltage biases were placed on the uncompensated steam flow signals. Based on the data obtained from 2-ST-302, this action reduced the Hi Steam Flow ESF actuation on Unit 2 below 110%. Since the same incorrect scaling information was utilized on Unit 1 in 1977, the voltage biases were also placed on the Unit 1 signals to effectively reduce the 1977 scaling error, thereby reducing the Hi Steam Flow ESF actuation value. '

5. O ADDITIONAL* CORRECTIVE .ACTION(S)

Following an engineering evaluation of the special test data obtained from 2-ST-302, the Unit 2 steam flow and feedwater flow transmitters were respanned and the steam flow circuitry and associated Hi Steam Flow ESF actuation functions were rescaled in Septem_ber 1991. These actions corrected the steam flow/feedwater flow mismatch to acceptable values.

• The Unit 2 scaling effort was reviewed for applicability to Unit* 1. From this review, it was determined that Unit 1 testing would be required. Therefore,

. independent testing is being* performed for Unit 1 in conjunction with the 1992 refueling outage. The test results will be evaluated and, as appropriate, the Unit 1 . steam flow and feedwater flow transmitters will be respanned and the steam flow circuitry and associated Hi Steam Flow ESF actuation functions rescaled.

6. 0 ACTIONS TO PREVENT RECURRENCE A scaling program is being undertaken to verify the scaling for ESF and RPS instrument loops. The program includes the development of a standard scaling methodology and scaling procedure. The new scaling process will then be utilized to perform the calculations for an instrument loop. Any enhancements identified in performing the first calculation will be incorporated into the methodology.

The verification of the scaling for the ESF and RPS instrument loops will then be coordinated with the development of upgraded calibration pro.cedures being

• performed under the Technical Procedures Upgrade Program.

7. Qi SIMILAR EVENTS None

8. 0 MANUFACTURER/MODEL NUM.BER N/A NRC Form 366A (6-89)