IR 05000440/1987009

Insp Rept 50-440/87-09 on 870429-30.Violations Noted: Design Review Performed by Architect/Engineer Did Not Verify Adequacy of Design

ML20215B527
Person / Time
Site:	Perry
Issue date:	06/08/1987
From:	Danielson D, Liu W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20215B472	List: IR 05000440/1987009 ML20215B471 ML20215B501
References
50-440-87-09, 50-440-87-9, NUDOCS 8706170356
Download: ML20215B527 (6)

Text

'

F

'

..

.'U.S. NUCLEAR REGULATORY COMMISSION

REGION III

.

't

-

-ReportLNo.50-440/87009(DRS)

~ '

. Docket No. 50-440- License No. NPF-58:

' Licensee:.. Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name: . Perry Nuclear Power Plant, Unit 1

,

Inspection At: Per_ry Site... Perry, Ohio.and Gilbert /Comonwealth, Reading,

-Pennsylvania Inspection Conducted: : April 29-30, 1987. at Gilbert /Comonwealth Company May 1, 1987 at Perry Site, Perry, Ohio Inspector: C iu # 87 Date

' Approved By: - H. Danielson Chief /#/87 Materials and Processes Date Section-

-Inspection Summary Inspection on April 29-30 and May 1,1987 (Report No. 50-440/87009(DRS))

Areas Inspected: Special announced inspection of actions pertaining to the licensee event report (LER 87003) associated with the failure'of the inboard Lcontainment ' isolation steam supply valve (1E51-F063) for the reactor ' core

.' isolation cooling (RCIC) system and LER 87027' associated with the loss of

condenser vacuum resulting from a break between the three inch main turbine

- stop -valve drain line header and the 24 inch main steam drain ' manifol H j(92700).

L 'Results: One violation was identified - Inadequate design review, Paragraph ,

?

>

"

-8706170356 870600 iPDR ADOCK 05000440 J)t PDR

F ..

..

}

'

.

DETAILS Persons Contacted

, .The Cleveland Electric Illuminating Company (CEI)

.

. .

• **K.~Pech,-General Supervising Engineer, Mechanical Design

,

]

• • C. Angstadt, Senior Project Engineer- -j

, **R. Matthys, Lead Mechanical / Piping, Quality Assurance.(QA) i

~

, **G. Dunn, Compliance Engineer

• • R. Stadel, Pipe Support Engineer

• • T. Jameson, Lead Mechanical Engineer, BO Gilbert Commonwealth Company (G/C) >

• Leininger, Director:of Projects

• J. Ioannidi, Project Manager

• F. Yurich, QA Program Manager

.

• Alley, Project Engineer, Structures

• J. Kadingo, Lead Mechanical Engineer i

• Denotes'those attending the exit interview at G/C on April 30, 198 [

- Denotes those attending the exit interview at CEI on May 1, 198 . Licensee Event Report (LER 87003, Revision 1) Background Information On January 10, 1987, the inboard containment isolation steam supply-valve (IE51-F063) for the reactor core isolation cooling (RCIC) .

system failed to open from the normally closed position causing ~the RCIC system to be inoperable. On March 13, 1987, the'RCIC outboard containment isolation valve (1E51-F064) was closed after the inboard containment isolation valve again failed to open. The closure of the RCIC outboard containment isolation. valve caused the RCIC system-to beLinoperable in accordance with Technical Specification-requirement The RCIC system is designed to maintain sufficient reactor water inventory should the vessel lose feedwater supply during a reactor vessel isolation condition. Should the RCIC system become inoperable when it is. required to be in service during a loss of feedwater, the High Pressure Core Spray (HPCS) system provides protection against a single failure event by performing the redundant function of maintaining reactor water inventory and adequate core coolin The RCIC system is therefore not an emergency core cooling (ECCS) system, i

- 2 i m, _ .. - ----- - - -

,

.

.., Corrective Actions As'a result of the. aforementioned events,' the licensee took various y corrective actions:which include the replacement of tne failed motors, theLreplacement of the operator, testing of the motor operated valve (MOV) using a motor operated valve ' analysis and test system (MOVATS),

and disassembly and inspection of the operator and valve internal Extensive testing and troubleshooting could not identify the root cause of the'above failures. Consequently, the licensee determined i to initiate.an engineering design modification to change the motor j operator. for Valve.1E51-F063 from a normally closed position to a l normally open position. This change increased the. availability of

.the RCIC system as it eliminated the active requirement of this valve for RCIC system initiatio The Effects of piping Ruptures in Conjunction With the Design Modification (1) The lice'nsee submitted its final design deficiency report-pertaining to pipe. rupture analysis on April 22, 1987, pursuant

,to 10 CFR 21 requirements. The report revealed that during the design modification described above, the full effects of-postulated.high energy piping ruptures were not accounted fo This error could result ~ in the loss of a single loop of residual heat removal (RHR)-shutdown cooling and low pressure coolant injection (LPCI). However, the number of operable loops / systems.of emergency core cooling systems (ECCS) would still meet the minimum requirements as defined in the Perry FSAR, Section_6.3.1.1. (2) The corrective actions which have been completed at the time of this inspection are identified below:

(a) One pipe rupture restraint was installed in each RHR room to limit the movement of the RHR piping due to the postulated jet impingement loa (b) One jet impingement bumper was installed in each RHR room to limit the deflection of non-safety related flush piping connected the RHR syste (c) Four pipe supports installed in the RHR A room were modified to withstand the postulated jet impingement loads. Design Review for the Design Modification l

L The design modificatian was to change the RCIC system inboard containment isolation valve (1E51-F063) from a normally closed

_ _ - _ _ _ - - _ -

r a

.

.

position to a normally open position. Gilbert / Commonwealth Inc. was the Architect / Engineer for the Perry facility and was responsible for the design modificatio The initial plant configuration specified the valve (IE51-F063) to be a normally open AC operated valve. The design of the RHR/RCIC piping and the effort to analyze the pipe break loads as well as the required supports / restraints and jet shields proceeded on this basis. Furthermore, all the pipe supports / restraints were installed prior to December 1983 with the exception of whip restraints RHR No. 6 and No. The reason that Restraints No. 6 and No. 7 were not installed was because the licensee informed the Architect / Engineer that the position of Valve 1E51-F063 was to be changed from normally open to normally closed due to equipment qualification concerns in the RHR rooms. Results of the engineering analysis performed by Gilbert revealed that the new loads associated with the valve in a normally closed position were smaller than the original design loads of the normally open position. Consequently, a number of supports / restraints could have been either deleted / remove However, the licensee decided to leave the supports / restraints intact; if they were already installed. The design evaluation for the valve in a normally closed position was completed in December 198 During start-up test activities, problems were encountered with the operator and the valve failed to open from a normally closed positio Consequently, an evaluation was initiated to determine the work scope and design effort required to change Valve IE51-F063 back to a normally open position. This decision was made by the licensee in Jaauary 198 The licensee requested Gilbert perform an evaluation of the effects of the change on the room environment for EQ consideration and on the design of the pipe rupture supports / restraints and other structure Gilbert concluded that the existing evaluation as described above in this paragraph met the design basis requirement since design loads due to postulated pipe breaks were accommodated by the existing design. Gilbert completed the evaluation and issued the results in February.198 Later it was noted that the effects of the postulated pipe ruptures in the RHR rooms were based on a review of design input information only. The output documents such as drawings were not reviewe Based on a misinterpretation of the design input information, it was assumed that all required pipe supports / restraints such as whip restraints RHR No. 6 and No. 7 were installed in the plan The licensee's review concluded that the cause of this design oversight ,

was that in performing the review, the design input information was 1 incorrectly interprete Further, the review did not include the i

.

(;

'.-

.

verification of. design output documents such as applicable drawing As a. result,.the reviewer made the assumption that the. required pipe rupture supports / restraints had been installed in the plant when.in fact'they were not. The failure to perform an adequate design review during a design modification is a violation of li0 CFR 50, Appendix B, Criterion II (440-87009-01)

Subsequent to the identification of the above concern, the. licensee redesigned and installed the required pipe supports / restraints. In addition, design reviews were performed to determine generic i implications. These reviews concluded'that the design oversight was an isolated' case. Consequently, no reply to this violation is required and the NRC. inspector has no further questions regarding this matte . Licensee Event Report (LER 87027 - Revision 0) Background Information On April 13, 1987, a manual fast reactor shutdown was initiated because of a loss of condenser vacuum and a steam leak in the turbine buildin The loss of. condenser vacuum was due to a hole in a 24 inch main steam drain manifold. The manifold fractured at the toe of the weld at the junction of a 3 inch main turbine stop valve' drain line heade The steam. leak was from the drain line header which had dislodged from the manifold. The licensee's evaluation concluded that the cause of the break was believed to be high frequency vibration resulting from extensive. steam flashing and water particle impingement in the drain line heade The drain line piping is non-safety related and performs no safety function. The main condenser is not required to support the safe shutdown of the reactor. Therefore, the event is not consider safety significan Corrective Actions The licensee promptly took various corrective actions to prevent recurrence. The corrective actions included: the replacement of the high pressure drain manifold with a heavier wall manifold (0.688 inch versus 0.375 inch), the installation of additional pipe restraints on the manifold and drain lines, the replacement of the 3 inch main turbine stop valve drain line header with a 6 inch line downstream of a throttle valve to reduce steam flow velocity into the manifold, and rerouting of an existing 6 inch drain line on the. manifold to have more evenly distributed flow force ___

f,'

u ;. 3 f' ,

j c. .

.

Review of Documents and Technical Discussion - l The NRC. inspector reviewed the applicable documents associated with the-aforementioned modifications and held' technical discussions with licensee representatives in. conjunction'with the modifications. The inspector concluded that the licensee's corrective actions appeared to be adequate and effective. Further, the modification' work was-monitored by the licensee during plant startup on April 19-21, 198 '

No problems were foun .

Within the areas inspected, no violations or deviations were identifie . Exit Interview The inspector met with site representatives (denoted in Persons Contacted

. Paragraph).~at the conclusion of the inspection. The inspector summarized the scope.and findings of the.. inspection noted in this report. .The-inspector also discussed the likely informational content of the inspection' report with regard to documents.or processes reviewed by the inspector during.the inspection. The licensee did not identify any such documents / processes as proprietar <

!

.

l 1