IR 05000213/1990083

Insp Repts 50-213/90-83,50-245/90-82,50-336/90-82 & 50-423/90-83 on 900924-28.No Violations Noted.Major Areas Inspected:Licensee Timeliness for Reporting/Operability Determinations

ML20217A214
Person / Time
Site:	Millstone, Haddam Neck  File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	11/05/1990
From:	Baunack W, Paolino R, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20217A208	List: IR 05000213/1990083 ML20217A202
References
50-213-90-83, 50-245-90-82, 50-336-90-82, 50-423-90-83, NUDOCS 9011200225
Download: ML20217A214 (18)

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I V. S. NUCLEAR REGULATORY COMMISSION REGION I -

' Report Nos. _50-213/90-83, 50-245/90-82 50-336/90-82, 50-423/90-83

.Dockec No , 50-245 50-336, 50-423 License N DPR-61, DPR-21 l'

DPR-65, NPF-49

< Licensee: Northeast Nuclear Energy _ Company P.O. Box 270 Hartford, CT 06141-0270

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Facility-Names: Connecticut Yankee Haddam Neck Nuclear Power Station and

' Millstone Nuclear Power Station Units 1, 2 and 3 Inspection At: Northeast Nuclear Energy Company, Corporate Office, Berlin, C ;

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. Inspection Conducted: September 24-28, 1990 Inspectors: /LA> ~r // - 6 - 9 O R. P olino, Senior Reactor Engineer, Plant date '

'S tems Section, EB/0RS-ll (Mr h-4~L // - f ~ 9 O W. H/ ~ Baunack, ySentbr Reactor Engineer date Op rati nal Programs Section, OB/ ORS .

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'W. J.'Raymo @ ,- Sen16r Resident inspector Il- V O '

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'M lls one Nuclear Power Station, DRP

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A. 'Vegel, Re ctor Engineer, Reactor Projects ll- 7-90

.date Secti n 4' , O P Approved!by: / #

C. J/ Anderson, Chief, Plant Systems-Section, date Engineering Branch, DRS~

-Inspection Summary: Inspection on September 24-28_,1990 (Inspection

+ Report Nos. 50-213/90-83, 50-245/90-82, 50-336/90-82, a 50-423/90-83) ,

Areas-Inspected: Special, announced inspection ;o review the licensee's timeliness for reporting / operability determinatfon Results: Based upon the review conducted, the team determined that the

' licensee.has an adequate program to ensure that-issues are properly evaluated

.for operability'and reportabilit No violations'were. identifie l 9011200225 902109 PDR ADOCK 05000213:

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i3 , E1[0 i P e r s o n s C o n t a c t e d - . . . . . - . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . - . . . . . . . . . 1

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L1'.1 Northea st' Nucl ea r- Ene rgy~ Company -... . . . . . . . . . . . . . .-. .-. .'. . . . . . . I

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.J3.0( Inspection Details?............................................. 2

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2 3 .1 s P roc ed u re s . . c. . . -. . . . . . .: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . : - 2

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' 3 . 2 c T r a i n i n g . . . .;- . . .._ -.,...3....._.................................,4

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3.4 Review of, Reporting Evaluation Forms .......;.....,.......... 5 q:

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DTimeliness Deficiencies ........,,-......... ,... 6L

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~3; Technical,Adequ'acy of-Operability /Reportabilit ,i c, , , .

1 De t e rm i n a t i o n s . . . . .; . . . . :. . . . .' . . . . . . . . . . . . . . -. . .. . . . . 7 _

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%,MMig .LAttachmentul:" .;. ~ . . _ > 1 DocumentsiReviewed

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i DETAILS

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1.0 Persons Contacted

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1.1. Northeast Nuclear Energy Company

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• E.-J. Mroczka, Senior Vice President, Nuclear Engineering and Operations  !

• C._F. Sears,.Vice President, Nuclear and Environmentel Engineering l

• Wi 0. Romberg, Vice President, Nuclear Operations  !

• E. A. DeBarba, Vice President, Generation Engineering and Construction j

• E. Scace, Director, Millstone Station  !

• L. Johnson, Director, Generation Engineering

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C. H. Clement, Unit' Director, Millstone 3  ;

• R. L. McGuinness,. Supervisor, Nuclear Licensing '

• G. P. VanNoordennen, Supervisor, Nuclear Licensing

• A, Blasioli, Supervisor, Nuclear Licensin M. Marino,. Unit Coordinating' Engineer, Nuclear Operations-

• T. B. Siiko, Licensing Engineer *

M.' V. Bonaca, System Manager, Reactor Engineering  !

J.~A. Blaisdell, Nuclear Engineering Consultant  !

V. Papadopoli, Senior Engineer, Quality Services Millsto'ne M. S. Wadkins, Engineer, Generation Electrical Engineering

.B. A. Tuthill, Supervisor, Generation Electrical Engineering

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=B. J. Smith, Technician, Generation Electrical ~ Engineering i;

-W.; J. Hayes, Jr., Senior Technician, Generation Electrical Engineering T. J. Mawson, Supervisor, Generation Mechanical Engineering

. .'L. Coleman,-Engineer, Generation Mechanical Engineering

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J. Becker,' Instrument and Controls Manager, Millstone 2  !

• D._D. McCory, Manager, Procurement Quality _ Services

j A. R. Roby,. System Manager, Generation Electrical Engineering 1,2 U.S.~ Nuclear Regulatory Commission- I

• 0. Have.'kamp, Chief, Reactor Projects Section_4A,' Division of s Reactor Projects 1

• 0enotes personnel present at the exit meeting on SAptember 28, 1990 j 2.0 Introduction l i

2.1 Scope The scope of this inspection was to evaluate the timeliness of the' i licensee's-reportability and operability program and its implementatio 'The basis for determination included: reviews of. training programs, m applicable audits, procedures and verification of licensee conformance to the established operability /reportability determination procedur '

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2.2 Background Safety and safety-support components, systems and. structures are '

designed to meet the requirements of the regulations, satisfy the

' licensing or design basis and conform to applicable codes and standard These requirements are explicitly stated for each licensed facilit For degraded-or non-conforming conditions of these systems and 3 structures; the licensee is guided by the Technical Specifications- '

and the provisions of 10 CFR Part 50, Appendix B, Criteria XVI, to promptly identify and correct . conditions . adverse to safety or qualit Reporting is required principally by Sections 50.72,50.73,-50.9(b)

and Part 21, Title ~10 of the Code of Federal.. Regulations (10 CFR) and the Technical Specifications. Each of these regulations requires documentation as well as a Section 50.71 report for an update of the final safety analysis report (FSAR). Collectively, these requirements constitute a process that forms the basis for the licensees to continue operations _or place the plant in a safe. condition, and to take prompt corrective action to return the plant to the licensing basi For the process to be complete, changes to the facility pursuant to 10 CFR 50.59 may be used as a partial fulfillment of the corrective action as required by Appendix .

Previous 1NRC inspections at the Mil _1 stone Nuclear Power Station have noted instances where-the timeliness for~reportability and-operability

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determinations could be improved. (Millstone-1 FWCI Operability-IR

, 50-245/89-25, Millstone 2. Service Water Check Valve-IR 50-336/89-24 and Millstone 3. Fast Bus' Transfer - IR 50-423/89-23). The issues of timeliness and Northeast Utilities (NU) reporting policies were _

C3 cussed with.the NRC on March 13, 199 The licensee responded to-the-NRC. concerns, by letter dated April.9, 1990, recognizing.that prompt identification, investigation and resolution of. safety questions is a matter.of high priority.__ In the April 9, 1990 response, the licensee acknowledg'ed that in the examples sited above, the reporting determinations (conducted under procedure NE0~2.25) could have been 1

. initiated at an earlier time and that ' procedural improvements.and

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heightened attention -in' this area-would help avoid future delays in

. reporting determination In the. NRC mid-SALP inspection-(Report No. 50-245/90-80; 50-336/90-80i 50-423/90-81) dated July 5, 1990,.the NRC staff noted.that further

. review was needed~to completely evaluate the effectiveness of licensee initiative The purpose of this inspection was to perform this review of licensee initiatives in order to complete the NRC's evaluation'of-their reporting progra .0 Inspection Details

. Procedures A' review was cmducted to determine if procedures have been -i established to assure that incidents or events which occur at the NU

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facilities are adequately reviewed for reportability to the NR ..

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Procedures which are intended to assure that incidents or events are reviewed for reportability are ACP-QA-10.01, Plant Incide;it Report .mi (For Millstone), ACP 1.2-16.1 Plant Information Report (for Connecticut Yankee) and NE0 2.25, Operability and Reportabilitv Determination (10.CFR-50.72, 10 CFR 50.73, and 10 CFR 50 9).

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Procedure ACP 1.2-16.1, Plant Information' Report (PIR) and ACP-QA-10.01, .

Plant Incident Report (PIR) both require that any situation or problem ,

which is' processed'under these procedures is evaluated for reportabilit l Both of these procedures require that.the shift superintendent as well as.the unit director perform reportability determinations. These two procedures are generally implemented at the= site and have a ver low threshold for events which are required to have a PIR prepare Procedure ACP 1.2-16.1 is more specific in identifying when a report-ability evaluation is to be performed for issues where reportability !

is uncertain. However, both procedures have provisions for the initiation.of reportability evaluation Issues for which reportability is uncertain in PIR's or identified during design. basis reviews, staf f work, or. other means are evaluated'

.by Nuclear Engineering and Operations. Procedure NE0 2.25. This-procedure provides instructions for the timely evaluation of-potentially reportable. items. Procedure _NEO 2-25 is implemented-

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-through the use of a Reportability Evaluation Form (REF).  ;

During"the review of NE0 2.25, it was noted that for several steps in l the body of the procedure where completion times were specified as .i

"shalls," the implementing form had the same_ steps identified as

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"should." Also, he body of the' procedure: requires that if an evalua-

. tion -cannot be cenpleted within 20 days a preliminary reportability '

recommendation'st all. be provided. This procedural requirement is not reflected on.the implementing form. The preliminary reportability

. recommendation n)t being specified on the form, may be the reason it has on occasion oeen provided late or notiat all. The licensee; indicated that these aspects of the' procedure-will be' reviewed for

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-incorporation into the next. revision of the procedur '

' Procedure NE0 2.25 was initially issued as:a method of evaluating-

.potentially reportable items in June 1987. To improve its effective-

. ness, it was revised in October 1988. .In JuneJ1990, the procedure was again revised-to reference NEC 2.29 for JCO's .(Justification for

' Continued Operation) and to clarify the time constraints for completing evaluations. Major changes incorporated in this. latest . revision

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include: Operability Determination Instructions; Clarification of responsibilities of individuals to promptly initiate an evaluation when warranted; clarifications of the need for reduced transit time between parties, places time constraints upon the NUp0C Director for 1 final determinations of the operability /reportability issue and strengthens requirements for timely evaluation i

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i-l The team concluded that procedures have been provided to assure that potentially reportable conditions identified both at the sites or by other means_.are evaluated and reported as required. The.reportability determination procedure has been revised several tires to' improve its effectiveness particularly in the area of timely completion of evaluations. Further improvements to the procedure are being considered to revise the implementing forms to reflect procedural requirement i 3.2 Training The-training provided to NU personnel describing their role in the reporting process was reviewed to verify that training has been provided describing company policies and more specifically the irplementation of Nuclear Engineering and Operations Procedure NE0 2.25, Operability and Reportability Determinations (10 CFR 50.72, 10 CFR 50.73, and 10 CFR 50.9).

The licensee has provided a Nuclear Engineering and Operations  ;

Procedure NE0 2.26, Departmental Training, which establishes the T

training requirements for. Nuclear Engineering and Operations Group 1 personnel. This procedure requires that supervisors of personnel who i perform quality activities are responsible for ensuring that these ;

personnel are properly qualified. Among the training specified is-procedural training. The training records for two groups.were

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-selected by the NRC inspector at random to-determine if these groups ;

Lhad received training in procedure NEO 2.25. The groups selected . j were the' Safety Analysis Group in the Engineering Department. and the Generation Electrical Engineering Group in the Generation Engineering j

and Design Department. The training records required-by procedure NEO 2.26, were provided and showed that both groups had received ^ !

training in' procedure NE0 2.25, i

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.. Additional instructions are provided to licensee personnel-in the !

Technical Staff and-Management Continuing Training Program This !

was verified to include training'in reportability requirements. The l lesson plan associated with reportability requirements training was; a reviewed and found to describe: why the training was provided, what- !

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the reportability requirements are, the NU reportability philosophy, !

the process for determining reportability and operability, and the policies and procedures which have-been established to implement reporting. The lesson plan clearly identified the companies policy -I

' for timeliness, conservatism and consistency in reporting. The procedures associated with determining reportability, particularly U

NE0 2.25 were emphasized in the training. "Further formal training is .

anticipated to be conducted when a new reporting guidance document is '

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In addition to formal. training dealing with reportability requirements, !

several memo's-have been issued by management to stress the importance of timeliness in the completion of reportability determination Discussions with various personnel that received this training confirmed-the increased emphasis which has been placed on reportability/cperability determinations, l

i The team concluded adequate reportability training is being provided and also that management has stressed the importance of timeliness Lin the completion of reportability determination .3 Audits _ -

In November of 1989, licensee management requested from Quality

. Services.that a special audit on NE0 2.25, Identification and Imple- l

. mentation of NRC Reporting Requirements, be . performed to determine i timeliness of evaluations performed compared with the timeliness ,

specified in the procedure. The request for this audit reflected- 1 managements increased interest in timely reporting of issues to:the . '

.NR The' audit reviewed eleven Reportability Evaluated Forms (REF's) ,

generated.during early 1989. Four of the eleven REF's audited were '

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'found not to be completed within the time constraints of the procedur "

Overall the audit findings indicated that.some time limits had been

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Lexceeded and that the. procedure needed to be clarified to prevent i this from-happening in the future, . This Inspection Team's review of i REF's generated during essentially the same time frame agreed with  ;

the audit findings. The audits findings could have been more specific 1

. in identifying: procedural weaknesses, such as the, Shall vs. Should ;j differences which existed between the body of the procedure u d the '

REF form which-implements the procedure. -Likewise the number of-  !

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Linstances in which procedural non-adherence was identified, particularly .i in the preparation of required preliminary recommendations, was not  !

clearly identified in the audit report. .The results of the NRC team's !

review was discussed in detail ~with the lead auditor who indicated  ;

that although the. specific procedural weaknesses and examples of  !

procedural non adherence identified by the NRC team were not clearly- !

identified in the audit report, they were discussed in detail during the exit;for the audi ,

The team felt that the audit was a good initiative on the part of 1 management, that it was thorough and comprehensive, and that existing . !

problems were noted. .However, the audit would have been more beneficia tif the findings which were stated to have been presented at the: audit exit were identified in the audit report'to management. Nevertheless,  ;

E procedural changes were made as a result of the.- audit and significant i improvements were noted in the REF's completed more recentl <

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s 3.4 Review of-Reporting Evaluation Forms b The team reviewed 65 reporting evaluations, as lined in Attachment A, including' supporting documentation, PIR's, LER's and licensee

correspondence. Specifically, reportability evaluation forms were reviewed for completeness and timeliness, verifying that the guidelines provided in the Operability and Reportability Determinations Procedure, NE0 2.25 were being followe Emphasis was placed on the timeliness of managerial reviews and the use of preliminary reportability/ - ,

k operability recommendations when the time limits could not be met.

P 'The; inspectors also evaluated the reportability/coerability determina- .

tions made to ensure that they conformed with cucrent NRC guidanc [

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" In. reviewing the reportability evaluations, the Inspectors compared !

the tirieliness ar.J quality of the evaluations initiated in 1989 with those initiated in 1990 to determine if increased management attention :

and an improved _ procedure have had an impact on the evaluation progra The inspectors noted that the evaluations performed in 1990, especially under revision two of NE0 2.25, were more thorough and usually met 4 y the time requirements with a few exceptions, in comparison to the evaluations performed in 1989'- Withstanding the evaluations discussed-

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beiow,-the inspectors noted that implementation of Revision 2:of NE0 2.25 and the-increased management awareness in regards to timely reporting, has had a positive impact on the timeliness and quality of

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the evaluation . Reportability Evaluations with Significant

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Timeliness Deficiencies

~REF 89-37 Millstone Unit 1 CU-2A Failure Due to Full Instrument Air Pressure This evaluation was initiated on February 7,1989, and reviewed by the~ originators supervisor on February 13, 1989. Six months J

later, on August 21, 1989, the Generation Facilities Licensing .,"

Manager received this evaluation for-revie Eventually, in October, 1989, this issue was evaluated as not reportable and the file was closed. The inspectors were concerned with'the six month gap (February-August,1989). During this period, the-evaluation was apparently'not being reviewed. .No action was:

being taken to' track down the evaluation to expedite the revie The licensee investigated this breakdown in the review and-- .

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tracking process and determined.that the reportability evaluation

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was lost in the in-basket of an employee who had left the compan .

The licensee recognized that the reportability evaluation program !

-was. deficient in tracking this issue and that further emphasis on - tracking reportability evaluations' was needed. To prevent

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such a problem from reoccurring, the licensee'has conducted training,to educate personnel on their responsibilities in i processing reportability evaluations to ensure that the evalua-tions are properly turned-over to different entities involved in

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evaluation's initiated in 1Y90, fewer timeliness deficiencies

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l reportability evaluation p'.ocess-has had a positive impact on (

the timeliness of evaluation l REF-90-43 Haddam Neck - Reactor Coolant System Hot and Cold Leg Isolation Valve Operability:

This issue was first documented and dispositioned for further repo'rtability. evaluation in a Plant Incident Report (PIR), dated June 7, 1990. The reportability evaluation was not initiated until JulyJ 10,1990, more than a month later. The team questioned why a reportability evaluation was not initiated when the deter-mination was made in the _ PIR that further . evaluation was necessar The licensee noted this deficiency, surmising-that further i guidance should be provided in the PIR to ensure that entry into .

. procedure NEO 2.25 is initiated promptly after its been identified '

.that further evaluation is necessary.

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3.412 Technical Adequacy of Operability /Reportability Determinations REF-90-45 Millstone Unit 2 Reactor Building Closed Cooling Water Heat Erchanger Anchor Bolting Failure

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Reportability evaluation form (REF) 90-45'was -initiated on July 18, 1990,.to initiate further review of a deficiency previously identified with the anchor bolts-on the reactor building closed cooling water (RBCCW) heat exchanger The

. anchor' bolt deficiency was identified by the plant staff'during inspections completed during the 1989 refueling o~utage and was dispositioned .in plant-incident report- (PIR) 89-17, dated _ _ _ !

February 22,'1990. Inspections.during the'1989 outage' identified corrosion on anchor boltsifor the "C" RBCCW heat exchangers, and ,

repairs were completed for all three RBCCW heat exchangers during

.the 1989 outag The REF addresses two issues that apparently were not completely evaluated at the time'the-PIR was dispositioned: -(1) effects of seismic loads; and, (ii) other examples of supports with corrosion similar-to that on RBCCW heat _exchangers. These issues.are discussed in detail in Attachment 2 of-this repor Reportability Determination

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The inspector reviewed the bases for the licensee's determination-

that the RBCCW anchorage problem was not repe 'able. The reportability determination was made on February 24, 1989, based l on the observation that the "C" heat exchanger had remained l functional in spite of the anchorage proble l

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' The conclusion was supported by the calculation results completed by March 1, 1989, and summarized in the March 29, 1989, mem ..d Although not documented at that time, the engineering personnel interviewed stated an operability assessment, including seismic considerations, was used for a reportability determination in 1989. The operability assessment was documented in the ",

September 11, 1990 memorandum in response.to REF 90-4 i The operability assessment concluded that the RBCCW heat exchangers )

r and-associated piping would have remained operable during a design 1 h

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basis earthquake (DBE), although design allowable. load restrictions i x would not necessarily be satisfie This conclusion was based ,j on engineering ' judgement and included .the following considerations:

piping systems are flexible under seismic loadings; the degraded bolts still' had some load carrying capability; and 'a preliminary !

analysis with degraded anchorage showed heat exchanger stability with_ margins ten to twenty percent below the design allowable limit '

The-licensee concluded that a. detailed coupled-model analysis would demonstrate.that allowable loads'would be exceeded, but *

- the system pressure boundary would remain intact and operabl ,

'The licensee-decided not to expend the resources to perform the more complex analysis since the modifications had been completed to correct: the deficiencie ,

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These sections _ of the-rules require the reporting of events that !

result in'the plant being seriously degraded or in an unanalyzed t co.ndition that significantly compromises plant safety,'or 3 conditions ~where'a single cause results in multiple trains to be "

inoperable. The reporting guidance of.NUREG 1022 allows the use-of engineering judgement to determine whether_an unanalyzed .

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condition exists. The' licensee concluded, based on analyses and engineering judgement,_that the RBCCW' system in the as-found ,

condition was operable for normal operational loads'and would a remain functional for design basis conditions such that plant safety was -not significantly compromised. Thus,-the condition was not' reportable. - The inspector did not identify any inadequacies in the licensee's conclusions.-

4.0. Exit Meeting The inspectors met with licensee corporate personnel at the conclusion of the inspection on September 28,.1990 the corporate office. The inspectors summarized the scope of the inspection and the inspection findings at that' tim During this inspection, no written material was provided to the licensee, ,

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ATTACHMENT 1

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_'D0CUMENTS REVIEWED .

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-REF# PLANT SUBJECT [

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b?' ~ 88-121 MP1 Possible Inadequate NPSH for LPCI Pumps h dueLto a Low Value (5 psig)_ of Containment '

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Spray Interlock 4

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89 )9i2

.MP3" - Undetectable Failures of Rosemount Transmitters'

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C l89-30 C PIR89-99: Fire'DamperDesign/ Inst $11ationProblems- 3

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.. 'i l89-33; MP Nonconforming Stud Material

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89-36 MP1 CRD System Leakage During LOCA j h ';-89 L37 MP CV-2A Solenoid Valve Failure Due To' Full Instr.

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,189-38 ,MP3i SBO Containment Responses-

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-MP2-  : SB0_ Containment-Responses; 'j m,.

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u-89-40A MPSi C&D Model ARRJBattery Char'gers

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Motor Cooler sCheck Valve Problems'

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(89-44? , LCY 480V. Load-Center-125V-DCcControl-. Power': Problems

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V59-I45 MP , NUSCO QSD' Audit:.A60496Lof:PDCR 2-30-8 ,

4 , 789-461  : CY? Cicle<15 Failed Fuel

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h CY' Non-Detectable Failure of Thermal-0verload-~

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DOCUMENTS ~ REVIEWED N REF#J -PLANT- SUBJE_CT 90-44; :MP1 Seismic Mounting of Diesel Control Panel L'90-45- MP2' RBCCW: Heat Exchanger Anchor Bolting Failure E90-d6 MP3 H2 Monitors and Recombiner Building Ventilation

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DOCUMENTS REVIEWED

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REF# PLANT SUBJECT 90-66' MP1 ECCS Pump NPSH Concerns-Supression Pool Temperature h 90-67 MP1 Loss of Feedwater Flow Analysis 90-68 MP1- CRD Pump Seal Cross Connection

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ATTACHMENT 2-ANCHOR BOLT DEFICIENCY (REF-90-45)

PIR 89-17 Deficiency and Resolution '

The deficiency was originally identified during the 1989 refueling outage when workers noted a loose anchor bolt on a base plate for the "C" RBCCW heat >

-exchanger.-_The deficiency was documented in PIR 89-1 Each RBCCW heat-

. exchanger has.four anchor bolts, two of the four anchor bolts on the "C" heat exchanger were found to be corroded to the point of failure, The two bolt ' failed at.the joint between the base plate and concrete pad with the nut separated'from_.the lower half of the anchor, Actions were taken during the--

outage under -plant design change evaluation-(PDCE) MP2-89-023 to repair the anchor bolts on all three heat exchangers, Nonconformance report 289-056 was issued to document the "C" heat exchanger anchor problem and to address the-question of corrosion of anchor bolts generically,

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Event Cause Analysis Licensee'inve:tigation under PIR 89-17 determined that the anchor bolt corrosion was caus~ed by salt water that infiltrated the ~ joint between the base ;

plate and the-concrete pad. _The anchor-bolts were found to be severely- '

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corroded-(necked down) from the original 1 inch nominal diameter to about 1/16 inch diameter, or less. The corrosion occurred over about 1 inch or so of the length of the anchor and only at the interface between the base plate and the-concrete (grout) anchor pad,

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LThe corrosion occurred slowly over the fifteen years that the heat exchangers

~had been.-in service. . ~ The base plates were wetted 'by, either leaks in service water piping or during periodic maintenance on the' heat exchangers, such as when the end bells are opened to clean or repair tubes, The corrosion-occurs due-to'the' base plate design configuration which ham ~pers quick'or complete drying of the anchor once the base plate was submerged'or sprayed with sea water,

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-PDCE MP2-89-023 The action taken during the.1989 outage to resolvethe RBCCW anchor bolt

. deficiency was to repair the anchors for all three heat-exchangers under PDC MP2-89-02 ,'

The spare anchors on'the RBCCW HX base plates, located just outboard of the '!

' failed ones, were also subjected to..the same wetting' conditions as the in' '

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service anchors, Licensee inspection of_the spaie anchor bolts'for each heat .

exchanger showed no corrosion. ~ Modifications to v heat exchanger base !

plates were performed and the base plates were re-anchored using the spare i anchor bolts. Also, the base plates were welded to-the underlying embedment plate 'c  ;

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Attachment'2 2

-An epoxy sealant was installed around the base plates (in the joint between .

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the base plate and the concrete pad) to prevent infiltration of standing or ,-!

sprayed sea water and thereby precluding further corrosion. A program to-inspect the base plates periodically was initie.ted to ensure the coating

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i remains intact.-

Engineering Evaluation i The licensee provided the following memoranda to document. engineering evaluations of the failed RBCCW anchors; Calculation 86-007-1074GP, dated March 1, 1989; engineering memorandum'PSE-SA-89-72, dated March 29, 1989; and engineering-memorandum PSE-SA-90-177,. dated September 11, 1990. The inspector interviewed members of the NUSCO-engineering group responsible for the calculations  ;

regarding RBCCW anchor integrity and the _ operability assessments completed for !

PIR 89-17 and REF 90-4 I

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Calculation 86-007-1074GP provided a structural evaluation of the RBCCW heat exchangers in the-as-found conditions. The heat exchangers function as an analytical; anchor for the RBCCW and service water piping. .The evaluation j determined whether the-loss of. anchorage affected the ability of the heat i exchangers .to function as an anchor- for the piping (i.e. , rigidity), or if the - !

loss of_ anchorage would invalidate the stability assumptions against overturning used in the original desig ,

The "C" RBCCW heat exchanger had two of four anchor bolts completely corrodeci through.' Site engineers stated that when the anchors for the "A" and "B" heat exchangers were replaced under PDCR MP2-89-023, one other bolt on each was found to be corroded through also, The calculation assumed a total loss of  :

anchorage on-all three heat exchangers to conservatively bound the field condition j

1-A simple hand calculation was used employing conservative _ methods to apply

RBCCW and service water piping nozzle reactions statically to the heat i

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exchanger nozzles. The heat exchangers are ~QA,; seismic category I equipment j

j-and'are designed for thermal, deadweight, pressure and seismic SSE loads. The j calculation did not consider seismic. loadings for the rigidity' and stability  ;

T evaluations, since the: plant had not experienced an operational basis i earthquake even '

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The calculation showed that, even with a complete . loss of anchorage, the heat exchangers~will function as designed for pressure, deadweight-'and thermal hydraulic load- The RBCCW and service water-piping;Lloads' remained restrained

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by the heat.exchangers, by. the frictional' shear of thelheat exchanger base -

plates and by the rotational resistance provided by the large _ moment of

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inertia provided by-the-heat exchangers. .The factors of-safety against

. overturning and sliding were 3.63 and 3.57, respectively. ~Thus, the loss'of-anchorage did not affect the ability o.f the heat exchangers to function as an

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anchor-for the piping systems. The inspector's review of the calculation-identified no inadequacies in the methods, input assumptions and result ;

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program to Evaluate Seismic Category I Supports '

During interviews with site engineering personnel, the inspector noted the licensee _had previously identified the potential for anchor bolt corrosion and the need t3 address the concern generically, particularly in light of the experience.with the RBCCW heat exchangers. The corrosion mechanism and the location of bolt wastage allowed significant loss of n.aterial with attendant

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, loss of margin to the' bolt design strength, with few obvious external

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indications of corrosion or degradation. Indirect evidence of underlying 1

. corrosion included cracked grout or rust weapage on or around the support base

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Surface visual inspections and bolt torquing techniques w$e deemed inadequate for detecting:significant hidden corrosion. Thus, a program to systematically !

d1sassemble.and inspect the areas under the support base plates was considered _ [

n_ecessary to assure corrosion was identified and corrected. Site engineering- !

began to_ develop an_ inspection program following the 1989 outage and inspections were in? progress during the 1990 refueling outage'to address the issu The-licensee presented for NRC review a copy of draft procedure EN.21216,  ;

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Inspection /of Supports-Exposed'to Water Intrusion. The procedure and i Inspections were being .'mplemented during the present. outage to identify and :

correct support-anchor _ deficiencies of the type noted on the RBCCW heat  !

exchangers. JThe licensee plans to inspect hangers in phases over-the course-

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of:several outage Thelinspector.' reviewed the draft procedure:and noted it containedcinstructions-for the inspection, protection and. replacement _of seismic hangers and anchor -$

bolts exposed to' sea water. ' The' procedure contained a list of supports that .;

would be' evaluated.: including hangers on the RBCCW heat exchanger drain and ;

relief ~ valve lines, and_those;in the' service witer pipe-tunnel, the intake i structure, and the diesel generator. areas. The list was developed based 1on a' <

review by plantiengineering--personnel of areas most susceptible to the type of ;

corrosion found on the RBCCW heat exchangers. The' licensee' stated.that the :

evaluation would be. done_ by physically disassethbling the supports -and .lif ting

the'b'as_e plates to allow direct visualLinspectionz of the underlying anchor 'The' procedure a~1so' contained instructions to use ultrasonic (UT) inspection

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techniques as_necessary to evaluate.the condition--of the anchor bolts, and to

develop criteriaLfor further use of the UT inspections for futurezsupport: .

N ' evaluation i

= Work was 1n progress during the week of September 28 to; inspect an initial l

,y  ? list ofz7-supports, including 405561, 491023, 427714, 60027, 49108-2,'527045-A' ,

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y land 527063. The initial supports 5 were chosen to obtain an estimate of the'

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extent of anchor bolt-degradation. The selection criteria used included-hangers.on seismic Category 1-components _known to be subject to periodic l wetting; hangers in'the service water pipe tunnel; hangers with rust on or . ,

around theibase plates; and, one hanger in the. susceptible area'(for wetting)

which showed no external signs of degradatio .

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D E Attachment'2- 4-n 1

'Th'e hangers (inspected aslof: October 1 showed :' varying degrees of '.orrosion,

-which were documented using;nonconformance. reports.and will be dispositioned ._

by;NUSC0' engineerin Preliminary findings were that' base piate operability

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E would not be compromis'ed and design. margins would be maintained, in spite of

'O :some obvious' corrosion.and degradatio This_ itenilis open pending further ' review of: documentation available at the-

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site'regarding the 1989 RBCCW bolt deficiencies; documentation-of the seismic

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considerations.-used-in the engineering evaluation for the-RBCCW'operabilityr I?

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. assessment.in?l989,:and .the actions taken since the 1989 outage and:i ,

, progress during' the present-' outage to. address the potential' support ._

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' degradations. This item will 'be.followed as'part of-the routine resident

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inspections: documented _in report 50-336/90-22, 1

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