05000423/LER-1996-019, :on 960627,reactor Coolant Sys Power Operated Relief Valve Block Valves Inoperable Due to Potential Structural Design Deficiency.Full Scale Testing Has Been Performed by Kei & Results Provided in Rept

:on 960627,reactor Coolant Sys Power Operated Relief Valve Block Valves Inoperable Due to Potential Structural Design Deficiency.Full Scale Testing Has Been Performed by Kei & Results Provided in Rept

ML20134B986
Person / Time
Site:	Millstone
Issue date:	01/24/1997
From:	Peschel J NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20134B964	List: 05000423/LER-1996-019, :on 960627,reactor Coolant Sys Power Operated Relief Valve Block Valves Inoperable Due to Potential Structural Design Deficiency.Full Scale Testing Has Been Performed by Kei & Results Provided in Rept 05000423/LER-1996-019-01, Forwards LER 96-019-01 Documenting Event That Occurred at Millstone Nuclear Power Station Unit 3 on 960627.Util Commitments Made within Ltr,Listed
References
LER-96-019, LER-96-19, NUDOCS 9701310186
Download: ML20134B986 (4)

LER-1996-019, on 960627,reactor Coolant Sys Power Operated Relief Valve Block Valves Inoperable Due to Potential Structural Design Deficiency.Full Scale Testing Has Been Performed by Kei & Results Provided in Rept

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
4231996019R00 - NRC Website

text

-

I NRC FORM 366 U.s. NUCLEAR REGULATORY COMMISslON APPROVED BY OM8 NO. 3150-0104 g.g53 EXPIRES 04/30/98 IN oWA'T O E 7"IoPR O E'SF C

fS E O TfD L

' ^a"'?o*L"u3"! "^PoV!e "c%Jea" ornea ^$ta A

LICENSEE EVENT REPORT (LER) l5mu? 'Oc'LTTe,"u'"ony"'E8"fais"s# '"M,"#ot"e

^

l 25% ^&caWi#1=E"M'M73T"oi"d'He""

(See reverse for required number of digits / characters for each block) l FAc JTY NAME (1)

DOCKET NUMBER (2)

PAGE13)

)

l Millstone Nuclear Power Station Unit 3 05000423 1 of 2 TITLE 44)

Reactor Coolant System Power Operated Relief Valve Block Valves Inoperable due to Potential Structural Design Deficiency EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FAclLITIES INv0LvED (8)

MONTH DAY YEAR YEAR SEQUENTIAL Revision MONTH DAY YEAR FAclDTY NAME DOCKET NUMBER l

06 27 96 96 019 01 01 24 97 OPERATING THIS REPORT is SUBMITTED PURSUANT TO THE REQUIREMENTS oF 10 CFR i: (Check one or more) (11)

MODE (9) 5 20.2201(b) 20.2203(a)(2)(v)

So.73(a)(2)(i)

So.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.22o3(a)(3)(i)

X so.73(a)(2itii) 50.73(a)(2)(x)

LEVEL (10) 000 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER l

20.22o3(a)(2)(iii)

So.36(c)(1)

So.73(a)(2)(v)

Specify in Abstract below or in NRC Form 366A 20.2203(a)(2)(ivl So.36(c)(2)

So.73(a)(2)(vii)

LICENSEE CONTACT FoR THis LER (12)

NAME TELEPHONE NUMBER linclude Atma Codel l

J.M. Peschel, MP3 Nuclear Licensing Manager (860)437-5840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE r

10 NPROS TO NPROS l

l SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR submission l

YEs f No (if yes, complete EXPECTED sUBMisslON LATO.

ABSTRACT (Limit to 1400 apaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On June 27,1996, at 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> with the plant shutdown in Mode 5, the Reactor Coolant System (RCS) Power Operated R:,li:f Valves (PORV) Block Valves (3RCS*MV8000A/B) were determined to be unable to perform their intended safety l

functions to close and reopen under design basis accident conditions. Tests performed at Kalsi Engineering Inc. (KEI) l provided evidence showing the valves would require greater thrust to close than had been previously calculated, and damage to the valve during attempted closure under design basis conditions could prevent reopening. Since this was potentially a condition outside the design basis of the plant, an immediate notification was made at 1445 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.498225e-4 months <br /> on June 27, 1996, pursuant to 10CFR50.72(b)(1)(ii)(B) and on July 26,1996, a Licensee Event Report was submitted under the provisions of 10CFR50.73(a)(2)(ii)(B). Failure of the valves to perform their required opening or closing function during d: sign basis events could result in difficulty controlling Reactor Coolant System (RCS) pressure and inventory, thereby not:ntially increasing the severity of an accident. The cause of this event appeared to be a structural design deficiency.

Further full scale testing has been performed by KEl and the results provided in a report. Based upon these new results, an inspection will be performed on the subject valves and, as necessary, a plant modification w!il be imp!cmented, prior to entry into MODE 4, that will ensure that the valves can perform their intended safety function and are restored to an operable condition.

~ Y655id'1ii6 970124 P

9 PDR ADOCK 05000423 S

PDR g.

gFoRM 366A U.s. NUCLEAR REGULATORY Commission UCENSEE EVENT REPORT (LER)

(

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3) l YEAR SEQUENTIAL REvlSION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 4 96 019 01 l

\\

l TEXT Uf more space is required, use additional copies of NRC Form 366A) (17}

1.

Description of Event

l l

On June 27,1996, at 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> with the plant shutdown in Mode 5 the Reactor Coolant System (RCS) Power l

Operated Relief Valves (PORV) Block Valves (3RCS*MV8000A/B) were determined to be susceptible to damage that would prevent them from performing their intended safety functions to close and reopen under design basis accident conditions. Tests performed at Kalsi Engineering Inc. (KEI) provided evidence showing the valves would require gr3ater thrust to close than had been previously calculated, and damage to the valve during attempted closure under l

d: sign basis conditions could prevent reopening. Since this was a condition outside the design basis of the plant, an immediate notification was made at 1445 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.498225e-4 months <br /> on June 27,1996, pursuant to 10CFR50.72(b)(1)(ii)(B) and l

subsequently, on July 26,1996, LER 96-019-00 was submitted pursuant to 10CFR50.73.(a)(2)(ii)(B). This supplemental LER is also being submitted under 10CFR50.73.(a)(2)(ii)(B) to detail the additional evaluation and rtsultant corrective actions determined necessary, l

11. Cause of Event

Th3 cause of this event has been determined to be a structural design deficiency. The valve body and the valve wedge on each PORV Block Valve are fabricated of type 316 stainless steel. With valve body guide rail-to - valve l

wedge guide slot surface clearances less than roughly 60 thousandths of an inch, binding can occur between the guide rail and wedge surfaces due to a build-up of gouged material. Therefore, if the PORV Block Valve clearances are less i

than 60 thousandths, this binding condition could exist under design loading resulting in abnormally high closing thrust i

l requirements which, in tum, could prevent the valves from performing their intended safety function. Since the valves have not been inspected to determine their as-built clearances, they will continue to be considered inoperable until found otherwise.

l Ill. Analysis of Event The safety function of PORV Block Valves (3RCS*MV8000A/B) is to be able to open or close in order to control RCS pressure in the event of the failure of the PORV.

In Generic Letter (GL) 89-10 (June 28,1989), " Safety-Related Motor-Operated Valve Testing and Surveillance," the US Nuclear Regulatory Commission (NRC) staff requested holders of operating licenses and construction permits to l

provide verification of the capability of safety-related motor-operated valves (MOVs) and certain other MOVs in safety-rzlited systems by reviewing MOV design bases, verifying MOV switch settings (initially and periodically), testing MOVs under design basis conditions where practicable, improving evaluations of MOV failures and necessary i

corrective action, and trending MOV problems.

l As part of the plant's response to GL 89-10, calculations were performed by Kalsi Engineering Inc. to determine the thrust required to operate valves under design basis conditions that could not be tested in-situ. The pressurizer PORV Biock Valves (3RCS*MV8000A/B) were among those evaluated. The initial eva!uation for these valves indicated the possibility of unpredictable behavior in mid-stroke which could adversely effect the amount of thrust required to operate ths valves. Mock-up tests were initiated in early June,1996 to obtain empirical data upon which to base future calculations and comparisons to the conservative bounding values previously used. These tests were performed at K!!si Engineering Inc. (KEI) in Sugarland, Texas using a qualified separate effects rig which was part of the EPRI MOV Program. The purpose of the test was to determine stem thrust at mid-stroke where the unpredictable behavior was expected to occur. During setup and equipment checkout of the KEl simulation rig prior to the test, loads were applied to the disc in various positions to verify the load profile that would be used during the test. This evolution resulted in an l

• U.s. NUCLEAR REoULATORY Commission (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAoE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 4 96 019 01 TEXT (11more space is required, use additional copies of NRC Form 366A) (17) un:xpected guide rail failure which had not been previously observed during EPRI full scale testing. The lower 2-1/2" of both guide rails deflected and sheared from the SS-316 specimen simulating the valve body. The test apparatus functioned properly, but the failure precluded continuation with the intended test.

Th3 failure exhibited during the test setup would have caused the valve to require more thrust for closure under design basis conditions than had been previously calculated. Although impossible to accurately predict at the time, it was assumed that the thrust required would be more than the available actuator output. Based on this information, the vilves were assumed to be incapable of performing their intended safety function.

Failure of the valves to perform their required opening or closing function during design basis events could result in difficulty in controlling Reactor Coolant System (RCS) pressure and inventory, thereby potentially increasing the severity of an accident.

Subsequently, Kalsi Engineering, Inc. (KEI) conducted additional full scale testing on actual components. This testing, documented in KEl Report 1959C, Rev 0, dated September 26,1996, determined that a structural design deficiency existed. The original testing, which was performed with a partial mock-up (i.e., only the body guide rails and the disc guide slots) demonstrated a failure of the guide rails under load. The subsequent full scale testing, with the addition of tha body seat ring and disc seating surface, was performed at design basis disc loading and temperature. Catastrophic failure of the guide rails was not expected, as their displacement would be limited by the disc contact with the seat, but some damage was anticipated based upon the results of the earlier test.

In the tests performed, both guide slots and guide rails showed significant damage. This damage was in the form of a

" rough machining" action between components that removed material (metal chips) and displaced material (galling),

which did not preclude valve operation when guide clearances were greater than 60 thousandths. Galling, when the guide clearance was close to the minimum value of the manufacturer's recommendations, was sufficient to re, Lire much higher thrust values which exceeded the capability of the currently installed actuators (i.e., the valves would not hnve closed under design conditions if the as-built guide clearances were at, or near, the minimum specified by the manufacturer. Because the PORV Block Valve as-built clearances have not yet been determined by inspection; they continue to be considered inoperable until found otherwise..

Examination of the components from the testing that produced a failure of the high operating thrust requirement indicated that the much greater thrust required was due to the buildup of material gouged from the component surfaces becoming trapped in the guide rail-to-guide slot interface, effectively reducing the clearance to zero, resulting in binding. This effect has been previously identified by the EPRI Performance Prediction Program for carbon steel on carbon steel guide surfaces with a guide clearance of less than.0625" (EPRI TR-103244, Nov.,1994) and can be applied to stainless steel because of the similar properties.

IV. Corrective Action

The valves were declared inoperable in accordance with the Technical Specifications. With the plant shutdown in mode 5, the Reactor Coolant System (RCS) Power Operated Relief Valves (PORV) Block Valves (3RCS*MV8000A/B) ars required to be maintained open as a portion of the Cold Over-Pressure Protection System. Based on this l

l requirement, Operations personnel have ensured the valves were open and maintained open following notification and j

d: termination that the valves were inoperable.

Further testing has been performed by Kalsi Engineering to determine the full extent of the problem and a report has t

be:n provided to the plant. Since the results of the additional testing identify the above-described galling prob'em, an inspection of the subject valves will be performed and, as necessary, plant modification will be implemented, pior to entry into MODE 4, to correct the problem and restore operability.

~.

_. _. _.. _ _ _ ~ _... -. _ _ _.. _ _ _ _

• U.S. NUCLEAR REGULATORY CoMMISSloN (4-95)

LICENSEE EVENT REPORT (LER)

{

TEXT CONTINUATION

]

FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REViStoN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 of 4 96 019 01 TEXT Uf more space is required, use additional copies of NRC Form 366A) (17) 1 1

V.

Additional Information

None

Similar Events

No similar events have been reported.

Manufacturer Data PORV Block Valves,3RCS*MV8000A/B are 3 inch 1550# class stainless steel (SA351 CF8M) gate valves, style N-6226-EMO-SP manufactured by Crane-Aloyco Co. The valve is operated by a SMB-00 electric motor actuator manufactured by Limitorque Corp.

Ells System Codes Reactor Coolant System - AB Ells Eauipment Codes Block Valve - SHV