05000423/LER-1997-016, :on 970206,piping Supports for Portions of RHR Sys in Condition Inconsistent W/Design Installation Details.Caused by Improper Installation of Spacer During Initial Const Activities.Restored Piping Supports

:on 970206,piping Supports for Portions of RHR Sys in Condition Inconsistent W/Design Installation Details.Caused by Improper Installation of Spacer During Initial Const Activities.Restored Piping Supports

ML20136F454
Person / Time
Site:	Millstone
Issue date:	03/07/1997
From:	Peschel J NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20136F440	List: 05000423/LER-1997-016, :on 970206,piping Supports for Portions of RHR Sys in Condition Inconsistent W/Design Installation Details.Caused by Improper Installation of Spacer During Initial Const Activities.Restored Piping Supports
References
LER-97-016, LER-97-16, NUDOCS 9703140111
Download: ML20136F454 (4)

LER-1997-016, on 970206,piping Supports for Portions of RHR Sys in Condition Inconsistent W/Design Installation Details.Caused by Improper Installation of Spacer During Initial Const Activities.Restored Piping Supports

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
4231997016R00 - NRC Website

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NRC FORM 366 U.s. NUCLEAR REGULATORY COMMisslON APPROVED BY OM8 NO. 3160 0104 (4-9 51 EXPIRES 04/30/98 N[o co?LEETom a70uts skarfo tNN Ato 50 was M"?'?o^"Ern?ti "^'33% 'an'a".t#a??o ^.W!R LICENSEE EVENT REPORT (LER) 193tnao '"'

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FACluTY NAME (1)

DOCKET NUMBER (21 PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 4 TITLE (4)

Piping Supports for Portions of the Residual Heat Removal system in A Condition inconsistent With The Design Installation Details EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVislON MONTH DAY YEAR FAciUTY NAME DOCKET NUMBER 02 06 97 97 016 00 03 07 97 oPERATINo 5

THis REPORT is SUBMITTED PURSUANT To THE REQUIREMENTS oF 10 CFR 5: (Check one or more) (11)

MoOE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2Hviii)

POWER 000 20.2203(aH1) 20.2203(a)(3)(i)

X 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(aH3Hii) 50.73(a)(2)(iii) 73.71 20.2203(aH2Hii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 20.2203(aH2Hiii) 50.36(c)(1) 50.73(a)(2)(v) specifY in Abstract below 20.22o3(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FoR THIS LER (12)

NAME TELEPHONE NUMBER (include Area Codel 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 TO NPRDS TO NPROS SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YEs X No submission (If yes, complete EXPECTED sUBMisslON DATE).

ABSTRACT (Limit to 1400 spaces,i.e., approximately 15 single-spacedtypewntten tines) (16)

On February 6,1997, at 17:00 with the plant in Mode 5 at 0-percent power, it was determined that piping supports for portions of the Residual Heat Removal system (RHS) were in a condition inconsistent with the design installation details.

Specifically, while implementing piping support modifications to address conditions previously identified in LER 96-007-00, workers found the piping system to be outside ofits designed configuration. An engineering review determined on February 7,1997 that the supports would not function as designed. As a result, an immediate notification was made pursuant to 10CFR50.72(b)(1)(ii)(B) of a condition during operation that resulted in the nuclear power plant being in a condition that is outside of the design basis of the plant.

This condition is significant in that had the plant experienced a design basis accident in containment the potential existed that the RHS might have been unable to fulfillits required safety function. However, there were no adverse safety consequences from this condition since the RHS system has never been required to operate in a post accident environment nor has it experienced a seismic event.

Immediate corrective action was taken to restore the piping supports to their designed configuration. In addition, a representative sample of other safety related pipe supports will be inspected to determine if conditions exist which would cause the piping support to not perform as designed.

9703140111 970307 PDR ADOCK 05000423 S

PDR.

.U.s. NUCLEAR REoOLAToRY Commission (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL Revision Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 4 97 016 00 J

TEXT (If more space is required, use additionalcopies of NRC Form 366A) (17) j i

1.

Description of Event

On February 6,1997, at 17:00 with the plant in Mode 5 at 0-percent power, it was determined that the as-built condition of the piping supports for piping within the Residual Heat Removal system (RHS) were in a condition inconsistent with the details of the design and installation for the support. These conditions were discovered during the implementation l

of piping support modifications to address conditions previously identified in LER 96-007-00.

As a result of a design modification to upgrade the operating temperature limits for the RHS, two lines required modifications to the supports. Specifically, a two way vertical restraint (3RHS-1-PSR 008) required rework to

]

accommodate the increased thermal loading. While modifying the support, workers found three piping supports to be outside of their designed configuration. One pipe clamp (3-RHS-1-PSST 006) attached to a rigid strut was found to be loose, such that it was not carrying the designated verticalloading. The second pipe restraint (3RHS-1-PSR008) was found to have a 1/8 inch (") gap between the pipe and its support at cold condition. The third pipe clamp (3-RHS PSST 009) was found to have rotated on the piping such that it was outside of its design tolerance.

j An engineering review determined on February 7,1997 that the supports would not function as designed. This determination resulted in entry into Technical Specification 3.4.10 since the section of line affected could not be isolated from the Reactor Coolant System (RCS). Additionally, an immediate notification was made pursuant to 10CFR50.72(b)(1)(ii)(B) of a condition during operation that resulted in the nuclear power plant being in a condition that was outside of the design basis of the plant.

This event is reportable pursuant to 10CFR 50.73(a)(2)(ii)(B) as a condition during operation that resulted in the nuclear power plant being in a condition that is outside of the design basis of the plant.

II,

Cause of Event

The cause for the loose pipe clamp (3-RHS-1-PSST 006) was determined to be the improper installation of a spacer during initial construction activities. Installation of the spacer negated the clamping action which provides the load transfer mechanism. The second and third pipe support discrepancies can be attributed to the loose clamp on 3-RHS-1-PSST 006. It is suspected that sufficient thermal displacement had occurred to result in pipe support gaps and rotation of the piping clamp to an extent which was outside of design basis of the piping system.

An in Service Inspection (ISI) report issued in 1991, and subsequent engineering review, for the loose piping support condition of 3-RHS 1-PSST 006 was dispositioned at that time as acceptable and failed to identify the improper installation of the spacer.

The subsequent investigation was unable to determine the initial date of occurrence nor did it identify any rework of these supports that might have resulted in the problems noted. Therefore, it could be presumed that this condition existed as the result of assembly during initial construction..

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l-i NRC FORM 36*A 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 Revision Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 4 97 016 00 TEXT (11more space is required, use additionalcopies of NRC Form 366A) (11) l l

111. Analysis of Event i

j Thsre were no adverse safety consequences from this condition since the RHS system has never been required to operate in a post-Loss of Coolant Accident (LOCA) or post-High Energy Line Break (HELB) environment nor has it experienced a seismic event. However, there is no practical way to verify the actualloading being carried by the supports in their "as-found* configuration.

This condition is significant in that had the plant experienced a design basis accident in containment such as a LOCA or a HELB, the potential existed that the RHS may not have been able to fulfillits required safety function.

l

IV. Corrective Action

immediate corrective action was taken upon the discovery of these deficiencies to restore the piping supports to their designed configuration The RCS was maintained at less than 112 degrees Fahrenheit ( F)in accordance with the requirements of TS 3.4.10. while the repairs were completed.

The failure of the pipe support will be treated as required within the ISI program. Specifically, a sample, in accordance

)

I with the American Society of Mechanical Engineers (ASME) code, Section IWF, Paragraph 2430, of other safety related l

pipe supports will be inspected prior to May 15,1997 to determine if conditions exist which would cause similiar piping supports to not perform as designed.

V,

Additional Information

This event was discovered as a result of modifications being performed in response to:

LER 96-007-00 "Condnment Recirculation Sorav. Quench Sprav. and Safety injection Systems Outside Desian Basis Due to Desian Errors" On April 3,1996, at 13:55, with the plant in Mode 5 at 0-percent power, it was determined that the plant had operated in a condition that was outside the design basis due to a deficiency in specific design conditions for a system needed to remove residual heat and mitigate the consequences of an accident. It was determined that the Containment Recirculation System (RSS) spray piping and supports were not adequately designed for thermal loads resulting from accident temperatures.

Accident temperatures could result in stresses above the design allowable stresses for plant

" Faulted

• conditions.

l It was initially determined that the higher RSS temperatures could result from a postulated loss of Service Water (SWP) to one or more RSS heat exchangers. It was subsequently determined that:

)

(a) unacceptable stresses in RSS piping, Quench Spray System (OSS) piping, and portions of piping which comprise the Safety injection (SI) flowpath, and the associated supports for those systems, I

could also result from the design basis accident temperatures inside containment, and (b) the original i

i-

2 o

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

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 of 4 97 016 00 TEXT (11more space is required, use additionalcopies of NRC Form 366A) (11}

design basis analyses for the RSS and OSS systems utilized support anchor movements which were nonconservative.

At the time of discovery the plant was completing a shutdown for unrelated reasons. Plant systems responded normally to the shutdown. No Engineered Safety Features Actuations were required or t

were initiated as part of the shutdown.

As corrective actions, design reviews of the RSS, OSS, SI, and other systems are being performed, design improvements will be made, and the systems will be restored to appropriate design basis requirements prior to declaring the systems operable for other than modes 5 and 6. As action to prevent recurrence, those systems which would be exposed to a Post-LOCA or Post HELB environment and which are required to mitigate the consequences of a design basis accident are being reviewed in order to determine whether or not they are susceptible to the same types of design deficiencies i

Similar Events

l None l

Manufacturer Data Ells System Code Residual Heat Removal / Low pressure Safety injection System:

BP Ells Component Code Support:

SPT