05000324/LER-1996-001, :on 960202,planned Reactor Manual Scram Inserted to Support Start of Unit 2 B212R1 Refueling Outage. Caused by Accelerated Aging.New Assembilies Replaced & Increased Frequency of Monitoring Performance

:on 960202,planned Reactor Manual Scram Inserted to Support Start of Unit 2 B212R1 Refueling Outage. Caused by Accelerated Aging.New Assembilies Replaced & Increased Frequency of Monitoring Performance

ML20107M418
Person / Time
Site:	Brunswick
Issue date:	04/26/1996
From:	Levis W, Tabor S CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BSEP-96-0172, BSEP-96-172, LER-96-001, LER-96-1, NUDOCS 9604300386
Download: ML20107M418 (5)

LER-1996-001, on 960202,planned Reactor Manual Scram Inserted to Support Start of Unit 2 B212R1 Refueling Outage. Caused by Accelerated Aging.New Assembilies Replaced & Increased Frequency of Monitoring Performance

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3241996001R00 - NRC Website

text

1 CP&L Carolina Power & Light Company P.O. Box 10429 Southport, NC 28461-o429 XER 261996 SERIAL: BSEP 06-0172 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 DOCKET NO. 50-324/L! CENSE NO. DPR 62 SUPPLEMENTAL LICENSEE EVENT REPORT 2-96-01 Gentlemen:

l in accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power &

Light Company submits the enclosed Supplemental Licensee Event Report. The original report fulfilled the requirement for a written report within thirty (30) days of a reportable occurrence.

Please refer any questions regarding this submittal to Mr. George Honma at (910) 457-2741.

Sincerely, W. Levis, Director-Site Operations 300.101 Brunswick Nuciear Piant SFT/sft Enclosures 1.

Supplemental Licensee Event Report 2.

Summary of Commitments cc:

Mr. S. D. Ebneter, Regional Administrator, Region 11 Mr. D. C. Trimble, NRR Project Manager - Brunswick Units 1 and 2 Mr. C. A. Patterson, Brunswick NRC Senior Resident inspector i

The Honorable H. Wells, Chairman - North Carolina Utilities Commission F

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PDR ADOCK 05000324 S /l S

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g digits / characters for each block) 1 FACluTY NAME (1)

DOCKET NUMBER (2)

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Brunswick Steam Electric Plant, Unit 2 05000324 1OF3 TITLE (4) l Control Rod Average 5% insertion Time Exceeds Technical Specification Requirements l

EVENT DATH (5)

LER NUMBER (6)

REPORT DATi (7)

OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION MONTH DAV YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000 02 02 96 96 --

01 01 04 26 96 OPERATING 01 MODE (9) 20.2201(b) 20.2203(a)(2)(v) j 50.73(a)(2)(i)

So.73(aH2)(viii)

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POWER 2a2203(aun 2a2203(an3Ha sanaH2Ha 50naH2Hx) 25 LEVEL l10) 20.2203(aH2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(aH2Hii) 20.2203(aH4) 50.73(aH2Hiv) y oTHER

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LICENSEE CONTACT FoR THis LER (12)

NAME TELEPHONE NUMBER tinclude Area Codel i

Steve Tabor, Sr. Analyst - Regulatory Affairs (910)-457-2178 CoMPLETF ONE LINE FOR EACH COMPONENT FAILURE DEE CRIBED IN THis REPORT (13)

R OR AB E R OR AB E

CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONLVT MANUFACTURER X

AA V

G080 Y

SUPPLEMENTAL REPORT EXPECTED (14)

MONN DAY YEAR EXPECTED YES submission X NO DATE (15)

(If yes, complete EXPECTED SUBMISSION DATE).

A38 TRACT (Limit to 1400 spaces. i.e., approximately 15 single-spaced typewntten lines) (16)

On February 2,1996, with Unit 2 operating at approximately 25% reactor power, a planned reactor manual scram was instrted to support the start of the Unit 2 B212R1 refueling outage. Subsequent review of the control rod scram time data rzysaled that the core average 5% insertion time was 0.387 seconds which exceeds the limit of 0.358 seconds specified in Tcchnical Specification 3.1.3.3. This time was primarily skewed by the 5% insertion times of 10 control rods with insertion tirms ranging from 0.53 to 1.078 seconds. Additionally,10 two-by-two control rod arrays exceeded the Technical Speciscation 3.1.3.4 limit for average scram insertion time. The cause of the degraf.ed control rod performance is attributed to rece erated aging of the control rod Hydraulic Control Unit Scram Solenoid Pilot Valve (SSPV) Buna-N diaphragms. The Eccelerated aging resulted from a less than optimal formulation of the Buna N diaphragm material. Corrective actions include replacement of the degraded SSPV assemblies, increased frequency of monitoring SSPV performance, and sample population inspections to monitor possible premature SSPV degradation.

Review of the Unit 2 scram insertion time data indicates that there was no operational safety significance associated with ths measured core average 5% scram insertion time. The actions taken at the time of this event met the requirements of tha Technical Specification and the degraded SSPV condition would not have prevented the fulfillment of the Control Rod Drive system safety function. Additional GE analysis determined that: (1) the impact of slower scram speed at the core avsrage 5% insertion point does not present a substantial safety hazard and (2) increasing the core average 5% insertion time to 0.49 seconds results in a less than.01 increase in the Critical Power Ratio. Thus this event is not considered reportable in accordance with the requirements of 10 CFR 50.73. Voluntary report 2-96-01 was submitted on March 1, 1996, because of the current generic issues involving SSPV diaphragms. This report supplements voluntary LER 2-96-01.

Additionally, this event was reported to the industry in Operational Experience Report OE 7683 on February 13,1996.

NAC FDAM 366 44 954

l.

$6RC On February 2,1996, following insertion of a planned manual Unit 2 reactor scram, review of the control rod scram time data revealed that the core average 5% insertion time was 0.387 seconds which exceeds the limit of 0.358 seconds specified in Technical Specification 3.1.3.3. This time was primarily skewed by the 5%

insertion times of 10 control rods with insertion times ranging from 0.53 to 1.078 seconds. Additionally,10 two-by-two control rod arrays exceeded the Technical Specification 3.1.3.4 limit for average scram insertion time.

This event is not considered reportable in accordance with the requirements of 10 CFR 50.73 since the actions taken at the time of this event met the requirements of the Technical Specification and the degraded Scram Solenoid Pilot Valve (SSPV) condition would not have prevented the fulfillment of the Control Rod Drive system safety function. Voluntary report 2-96-01 was submitted on March 1,1996, because of the current gsneric issues involving SSPV diaphragms. This report supplements voluntary LER 2-96-01. Additionally, this event was reported to the industry in Operational Experience Report 7683 on February 13,1996.

CAUSE OF EVENT

A representative sample of those SSPV assemblies which exhibited normal and slow start of motion times wEre disassembled and inspected by both CP&L personnel and an independent laboratory. The investigations concluded that the degraded scram times resulted from accelerated aging of the Buna-N diaphragms. The accelerated aging was found to be caused by a less than optimal formulation of the BureN diaphragm material. Specifically, the lack of antioxidant protectant in the formulation did not allow the material to resist oxidation induced degradation over a significant period of time in operation. This determination is consistent with the information regarding premature Buna-N diaphragm degradation provided in General Electric RICSIL 069.

A review of scram time history over the cycle revealed that indications of the failure mechanism were limited to l

a small number of degraded SSPVs. Out of the 56 control rods tested during the cycle, five SSPV assemblies wsre replaced to correct slow response times. Since these indications were manifested randomly rather than as widely occurring degradation, data review during the cycle did not identify a trend toward degrading scram times as the cycle progressed.

NRC FORM 366414 99

NflC FORM'366A U.S. NUCLEAR REGULATORY COMMISSION wes LICENSEE EVENT REPORT (LER)

TfT CONTINU ATION i

FACILITY NAME (1)

H

~30CKET LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Brunswick Steam Electric Plant, Unit 2 05000324 3

OF 3

1 96 -- 01 01 TEXT (11more space is required, use additiona copies of NRC Form 366A) (11) l

CORRECTIVE ACTIONS

Tha Unit 2 SSPV assemblies were replaced with new assemblies during the B212R1 outage. The new ass mblies utilize Buna-N exhaust diaphragms. From a review of industry data, the most limiting documented cass of Buna-N diaphragm service life is 2.9 years (Operational Experience Report 7543). Thus, it is expected that the current SSPV configuration will provide acceptable scram time performance throughout the present j

cycts.

)

Both Unit 1 and Unit 2 currently have Buna-N diaphragms installed in the exhaust side of the SSPV assemblies.

As en interim measure, monitoring equipment has been temporarily installed on Unit 1 to support scram time L

data collection during weekly Reactor Protection System functional tests. This increased monitoring will allow cartirr detection of diaphragm degradation.

A simple population of the Unit 1 and Unit 2 Buna-N diaphragms will be inspected to determine.if degradation his occurred (i.e., stiffening or cracking) and the need for exhaust diaphragm replacement during respective rafu:1 outages.

SAFETY ASSESSMENT

Revi:w of the Unit 2 scram insertion time data indicates that there was no operational safety significance associated with the measured core average 5% scram insertion time. The actions taken at the time of this evsnt met the requirements of the Technical Specification and the degraded SSPV condition would not have prsysnted the fulfillment of the Control Rod Drive system (AA) safety function. Additional GE analysis d5tarmined that: (1) the impact of slower scram speed at the core average 5% insertion point does not present a substantial safety hazard and (2) increasing the core average 5% insertion time to 0.49 seconds results in a Isss than.01 increase in the Critical Power Ratio. The Unit 2 scram time average was much less than 0.49 s:conds, with the actual time measured at 0.387 seconds.

PREVIOUS SIMILAR EVENTS

A similar event involving slow scram times was reported in LER 1-96-02; however, the cause of that event is attributed to the use of Viton SSPV exhaust diaphragms.

Ells COMPONENT IDENTIFICATION System /Comoonent Ells Code Control Rod Drive AA l

1-MC F03 366A H 95)

- -..,. ~ _

Enclosure List of Regulatory Commitments Th3 following table identifies those actions committed to by Carolina Power & Light Company in this document.

Any other actions discussed in the submittal represent intended or planned actions by Carolina Power & Light Company. They are described to the NRC for the NRC's information and are not regulatory commitments.

Pl:ase notify the Manager-Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

4 Committed

Commitment

date or outage A sample population of the Unit 1 and Unit 2 Buna-N B111R1/8213R1 diaphragms will be inspected to determine if degradation has occurred (i.e., stiffening or cracking) and the need for exhaust diaphragm replacement during respective refuel outages.

1 i