IR 05000285/1997013

Insp Rept 50-285/97-13 on 970630-0718.Violations Noted. Major Areas Inspected:Operations,Maint & Engineering

ML20217E677
Person / Time
Site:	Fort Calhoun
Issue date:	10/02/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20217E658	List: IR 05000285/1997013 ML20217E654 ML20217E672
References
50-285-97-13, NUDOCS 9710070132
Download: ML20217E677 (34)

Text

_____

.

I e-ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50-285 License No.: DPR-40 Report No.: - 50 285/97 13 Licensee: Omaha Public Power District Facility: Fort Calhoun Station Location: Fort Calhoun Station FC 2 4 Ad I i

P.O. Box 399, Hwy. 75 - North of Fort Calhoun 1

-

Fort Calhoun, Nebraska L Dates: June 30 through July 18,1997 Inspectors: D. Pereira, Reactor inspector, Engineering Branch M. Runyan, Reactor Inspector, Engineering Branch W. Wagner, Reactor inspector, Engineering Branch -

Approved By: Thomas F. Stetka, Acting Chief, Engineering Branch Division of Reactor Safety

= ATTACHMENT: - Supplemental Information i

i 9710070132 971002 PDR 0 ADOCK 05000285 PDR

.

.

TABLE OF CONTENTS i

1. Operations . . . . . . . . . ........................................... 1 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Operator Work Arounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Operations Support of Condition Reports . . . . . . . . . . . . . . . . . . 2:

01.3 Safety Review Committee Activities . . . . . . . . . . . . . . . . . , . . . 4 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 6

!

07- Quality Assurance in Operations . . . . . . . . . , . . . . . . . . . . . . . . . . . . . 6

.. l l . M a i n t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

.

M2 Maintenance and Material Condition of Facilities ano Equipment ......-9 M4 : Maintenance Staff Knowledge and Performance . . . . . . , , . . . . . . . . . 9 M7 - Quality Assurance in Maintenance ............,,............ 10 ill. Engineering ..................................................-11-E2- Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . 11 E2.1 Engineering Support of Operating Experience . . . .. . . . . . . . . . . 11 E2.2 Engineering Support of Condition Reports . . . . . . . . . . . . . . . . 12 E2.3 Engineering Support of Engineering Assist Requests . . . . . , , . . 16 E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 17 E7,1 - Quality Assurance Audits and Self Assessments ........... - 17 E7.2 Interview with Design Engineers . . . . . . . . . . . . . . . . . . . . . . . 19 E7.3 Root-Cause Analysis .................-.............. 19-E8 Miscellaneous Engineering issues , . . . . . . . . . . . . . . . . . . . . . . . . . 20 E8.1 -(Closed) Inspection Followup item 285/9405-01 ........,.. 20-VI. M an agement Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

.X1 - Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 ATTACHMENT: Supplemental Information

- ii

_

.

.

EXECUTIVE SUMMARY Fort Calhoun Station NRC Inspection Report 50-285/97 13 The objective of this inspection was to evaluate the effectiveness of the Fort Calhoun Station controls in identifying, resolving, and prever. ting problems that degrade plant safety. This review was focused on the following areas:

• Safety rev,ew committee activities

• Root-cause analysis j

Corrective action

Self assessment

Operating experience feedback The inspection consisted of an extensive review of plant documents, employee interviews, and meetings with licensee personnel to discuss technical or administrative question Operations

Operator work arounds were receiving appropriate management attention, and resolution and closure of operator work arounds were scheduled in a timely manner consistent with licensee priorities (Section O2.1).

In general, the operations staff conducted the day to-day resolution of problems in l

an effective manner. In addition, the plant review committee, the condition review group, and the corrective action group were effective in analyzing, assessing, and resolving issues of plant safety, and reliability. These groups were also effective in determining safety significance, prioritization, and appropriate root-cause determination (Section 02.2).

Operations personnel were knowledgeable and supportive of the correction acticn process. In addition, operations' involvement in the corrective action process was satisfactory and effective (Section 04.1).

The operations self assessment was effective in identifying concerns and recommendations. Operator work arounds, operations control center, and inconsistent expectations of the corrective action program were the three major concerns determined during the self assessment (Section 07).

)

iii

_ _ _ _ _

.

. . . . . . .

.

. . .

.

.

Maintenance

The corrective action process was being satisf actorily implemented by the maintenance department (Section M2.1).

Maintenance personnel were supportive of the corrective action program and confident in its effectiveness (Section M4.1).

The quality assurance audit and the self assessment provided a critical assessment of maintenance activities and were effective in identifying areas for improvemed (Section M7.1),

i l

Enaineerina

• The operating experience information was found to be disseminated app'opriatel Evaluation reviews and corrective actions for operating experience were being controlled and the licensee identified examples of improvements of the program in a self assessment, completed in June 20,1997. These improvements were l

considered to be effective in presenting plant events information to the plant staff (Section E2.1).

Engineering performance in thc disposition of condition reports was good, though

, several problems were noted including: one instance where a precondition concern

!

was not addressed; three instances where generic implications were not adequately considered; and, one instance where the cause of an event was not determined. A concern related to the manner in which the licensee and the vendor handled a butterfly valve overtorquing event was also identified (Section E2.2).

Engineering exhibited good performance in the disposition of the engineering assist requests reviewed by the inspectors. One example was identified where a condition adverse to quality was not reported on a condition report (Section E2.3).

The licensee had adequately assessed its engineering processes over the past 2 years. Within an audit of configuration control, the licensee identified a condition adverse to quality, but f ailed to initiate a condition report in accordance with procedure. The licensee had taken effective corrective actions to reduce the number of overdue responses to condition reports (Section E7.1).

• Engineers were found to be supportive of the corrective action program and confident in its effectiveness (Section E7.2).

The licensee was found to be effectively implementing a good root cause analysis process (Section E7.3),

iv

-

-- - -

-

-

.. . . . . . . _ . . . _

.

. . _ _ _ _

. .. . . . . . . .. . . .. .. . . . . .. .. . . . ..

.

.

Report Details 1. Operations 01 Conduct of Operations 01.1 Operator Work Arounds Lnjoection Scope (40500)

The inspectors reviewed the licensee's controls for ensuring timely corrective action of operator work arounds. The inspectors reviewed Operations Department Policy and Directive OPD-4-17, " Control Room Deficiencies and Operator Work Arounds,"

Revision 4, which described the process for controlling work arounds. The inspectors also interviewed operators to determine if the outstanding operator work arounds were impacting plant operations.

, Observations and Findinas Directive OPD-417 defined operator work arounds at, a plant component, which is deficient or inoperable, and some alternate action is required by an operator to I

compensate for the condition of the component. Operator work arounds could complicate operator responses to emergencies or transient condition The inspectors determined that the licensee maintained a high priority on operator work arounds, which received management attention during the weekly and a quarterly review process. The inspectors found that the current work around list had 20 open items listed. The licensee identified that 12 of the 20 items on the work around list were safety related. The inspectors interviewed the supervisor of operations who was responsible for reviewing the work arounds list for completion dates to resolve the work arounds, and for assessing their aggregate impact on plant operations. The supervisor of operations indicated that he was not pleased with the large number of operdtor work arounds and was concentrating on removing as many as possible. The supervisor of operations indicated that the current work arounds would not impact plant operation, (i.e., nuclear safety or safe shutdown capability).

During the interviews with the operators, the inspectors found that operators expressed no operational impact concerns with the outstanding operator s ork arounds. However, the operators felt that resolution or repair of the operator work arounds was slow and that the work around numbers appeared to be climbin ..

.

The inspectors identified that tSe operator work around list was presented weekly at-the plant review conimittee meeting, and that any operator work arounds that were considered significant, and could be resolved on-line, were placed in the plan of the-day operations priorities section to ensure prompt action for resolution. The inspectors verified that each operator work around was tracked and scheduled for elimination by a maintenance or corrective action resolution ite The inspectors found that the licensee had scheduled 11 operator work arounds for completion by the end of August 1997. The inspectors noted that 7 of these 11 operator work arounds were safety related, in addition, the inspectors noted 1 that the licensee had scheduled all but one of the operator work arounds to be completed by the end of the 1998 refueling outag Conclusi.gng The inspectors concluded that operator work arounds were receiving appropriate management attention, and that resolution and closure of operator work arounds

- were scheduled in a timely manner consistent with licensee priorities. Although a number of work arounds existed, the inspectors coi.cluded that the licensee had applied sufficient management attention to maintain the total numbers steady or slightly declining over the foreseeable futur .2 Ooerations Suooort of Condition Reoorts I Insoection Scoce (40500)

The inspectors evaluated the operations staff's efforts to resolve identified problems by reviewing 37 condition reports that were assigned to operations for resolution, observing the activities of the corrective action group, the condition review group committee, and by reviewing the evaluations of the enndition reports: The inspectors held discussions with operational staff personnel to determine how day-to-day resolution of problems were handled. The inspectors also interviewed selected operations staff personnel to determine their knowledge of the corrective action process and procedures at Fort Calhoun Station, Observations and Findinos The operational staff interviewed by the inspectors were knowledgeable of the corrective action process. The operations staff indicated that the day-to-day resolution of problems were resolved through the condition reporting process with Standing Order SO R 2, " Condition Reporting and Corrective Action," Revision The inspectors noted that the condition reporting system estab!!ahed measures to ensure that conditions adverse to quality were promptly identifieo, reported, and -

corrected. In addition, the condition reporting system established measures to ensure that the root causes of significant conditions adverse to quality were determined and corrective action taken to prevent recurrenc .

_

__

.

.

Upon receipt of a condition report, the shift supervisor reviewed the condition report for immediate corrective action to ensure safety of the plant, personnel, and the general public. Also, the shift supervisor ensured an initial operability and reportability determination were completed, in addition, the shift supervisor ensured reports were made for conditions requiring 1,4, or 24-hour reports to the NRC or other outside agencie The inspectors attended several corrective action group and condition review group meetings that were held every day. In addition, on July 2,1997, the inspectors attended a special plant review committee meeting. The inspectors determined that these groups performed the following functions:

• The corrective action group meeting provided overall administration of the condition reporting system and advised the condition review group regarding ownership and significance level ast.ignments of the condition reports, in addition, this group assigned trending codus to the condition reports, and evaluated condition report data trends to identify problem areas and adverse trend *

The condition review group meeting reviewed condition reports, and ensured that initial operability and verbal reportability determinations were promptly made, in addition, this group assigned a condition report significance level and a condition report owner. Fina'ly, the condition review group recommended plant review committea review of Level 1 and 2 condition report *

The plant review committee reviewed and approved Level 1 and 2 significant condition report responses. Level 1 and 2 condition reports represented significant conditions adverse to quality, such as nuclear safety or plant reliability concerns, The inspectors observed good communications between the various work groups, a good questioning attitude, comprehensive _and constructive analyses of the safety significance and root causes, and appropriate management attendance in every meetin For example, in the special plant review committee meeting held on July 2,1997, at approximately 7:30 a.m., the discussion for the root causes of the fire pump failures to start were presented for Condition Report 199700683. The root cause of the event was stated to be sand in the river. The plant review committee decided to reconvene at 2 p.m. that afternoon for reconsideration of the root causes of the fire pump failures to start. At the afternoon plant review meeting, the root causes were expanded to include a second root cause, that the design of the intake 3 l

.

.

, . . . . . ,. .

. .. .. . .

,

.

'6

. ..

f

'

structure did not provide a means to prevent, control, or limit, the amour t of sand, which collected at the diesel driven fire pump suction. The inspectors noted that these root causes appeared to address the cause of the start failures of the fire j pump. The licensee had made plans to modify the diesel engine fire pump with a j sparger i.nd had revised operating procedures to alleviate the problem. The inspectors considered these actions to be appropriat The inspectors concluded that the day-to-day resolution of problems was conducted in an effective manner through the condition report process, in general, the inspectors concluded that the operations staff conducted the day-to-day resolution of problems in an effective manner. In addition, the inspectors concluded that the plant review committee, the condition review group, and the corrective action group were effective in analyzing, assessing, and resolving issues of plant safety, and reliability. These groups, and the plant review committee, were also effective in determining safety significance, prioritization, and appropriate root-cause determinatio .3 Safety Review Committee Activities Insoection Scoce (40500)

The inspectors evaluated the effectiveness of the plant review committee and the safety audit review committee by reviewing committee minutes and audits conducted from June 1,1996, to June 1,199 . Observations and Findinas The inspectors found that the plant review committee's normally scheduled meetings demonstrated a good questioning attitude by the review committee members, who were also aggressively seeking out areas needing plant improvement. This was evident to the inspectors in the meeting minutes of Plant Review Committee Meeting 96160, which discussed the root-cause analysis for Condition Report 199601476, regarding an event where the reactor coolant system cooldown limit was exceeded. The discussion questioned the adequacy of the corrective actions and the need to review the Technical Specification changes before submittal to the NRC. The inspectors found another good example of a questioning attitude in reviewino Plant Review Committee Meeting 97-029 minute which discussed the operability evaluation for Condition Report 19970020 Condition Report 199700207 addressed failed Valve CH-462, where the lower wedge broke into four or more pieces. Although Valve CH-462 was repaired, there remained a concern for the chemical and volume control system, the reactor coolant system,.and possibly the safety injection system due to the effects that the missing wedge pieces could have on safe shutdown systems and components. After a detailed discussion, a plant review committee action item was assigned for system

. engineering and operations to implement appropriate compensatory actions regarding the missing pieces. Condition Report 199700207 remained opened during this inspectio ,

s

_ _ __ ___________ ___-_______- - _ - - - - - - - - - .

.

F

.

L

The inspectors found that the plant review committee met at a greater frequency 1 than the once per calendar month as requiied by the Technical Specifications. For

' example, there were 60 plant review committee meetings convened between January and June of 1997. Fifty-four of these were special meetings called to (

==

order by the chairman or the alternate chairman to discuss pertinent plant issue For example, Plant Review Committee Meeting 97-029 was convened to review and make recommendations on the proposed scope of the radiation controlled area for the steam line break event, which occurred on Apnl 21,199 The inspectors found that the safety audit review committee also demonstrated a good questioning attitude regarding their review of plant problems. For example, in Safety Audit and Review Committee Meeting 197, the committee members questioned the following items of concern; a temporary modificatic.) involving the

-

caps on the main steam safety valves; an industry event concerning the challenge to main steam safety valves; and problems of murky refueling pool water; and, the inoperable nuclear instrument during the core reloa When reviewing the safety audit and review committee charter, the inspectors noted that the composition of the committee, as described in Section V.a. of the charter, was different from that described in Technical Specification 5.5.2.2.

h Technical Specification 5.5.2.2 states that the safety audit and review committee chairperson shall be a senior vice president, with other members being composed of

. - -

the vice president and division managers.Section V.a of the charter made no reference to the senior vice president as chairperson, but Section V,b said that the vice president shall appoint a member of the safety audit and review committee as chairperson. Upon further review, the inspectors found that the licensee had

'

submitted an " Application for Amendment of Operating License," in their letter (LIC-96-0183) to the NRC, dated November 20,1996. This application included a proposed change to Technical Specification 5.5.2.2 that would replace the " Senior Vice President" as the chairperson with a " Member as appointed by the Vice President." The inspectors found that the Application for Amendment of Operating License was under review and had not been approved by the NRC at the time of this inspection. The inspectors found, by reviewing the 1997 safety audit and r- review committee meeting minutes, that the licensee had been implementing the change to Technical Specification 5.5.2.2 since January 17,1997. The licensee expressed that they understood that administrative changes to the Technical Specifications did not require NRC approval prior to implementation. The inspectors explained that whenever a senior menagement position of responsibility is deleted r from the Technical Specifications that the change represents a reduction in previous commitments and requires NRC review and approval prior to implementation, in response to the inspectors finding, the licensee initiated Condition Report 199700786, which was discussed with the vice president, division managers, and licensing manager, to take immediate actions to restore the official safCy audit and review committee back into Technical Specification complianca. The inspectors did not identify any other instances where a Technical Specification amendment was implemented prior to NRC approval.

_

K

'

--i misem

e

"

,,

Technical Specification 5.5.2.2, states, in part, that, *(tlhe Safety Audit and Review Committee shall be composed of a chairperson: that is a senior vice president."

The licenceo's failure to maintain the senior vice president as the chairperson of the safety audit and review committee is considered to be a violation of Technical l Specification 5.5.2.2 (50 285/9713 01).

c. Conclusions The inspectors concluded that the plant review committee and Safety Audit Review Committee wtre aggressively seeking out areas needinf; plant improvemen Operator Knowledge and Performance insoection Scoce (405Mi The inspectors interviewed ten operations personnel to determine their knowledge and involvement of the corrective action process, Observations and Findinog The inspectors found that the majority of the operations personnel interviewed were very positive about the corrective action process. Although most of the operations personnel did not personally initiate a condition report, they ensured that the condition report was initiated through the shift technical advisor. They indicated that the computer based system for the condition report was at first difficult to

-

overcome due to poor typing and computer skills. However, the operators stated that with practice, the system had become easier to use, in addition, the operators stated that they had no difficulty in obtaining approval for the condition report, and that implementation and support of the corrective actions were good, Conclusiong -

The inspectors concluded that operations personnel were knowledgeable and ,

supportive of the corrective action process. The inspectors also concluded '

operations involvement in the corrective action process was satisf actory and appeared to be effectiv Quality Assurance in Operations Insoection Scope (40500)

The inspectors reviewed the Operations Department Self Assessment Report FC-0026 97, dated May 15,1997, for the assessment conducted during the period of March 17 26,1997. The inspectors reviewed the self assessment to determins if it was effective in identifying strengths and areas of concer i

\\

-- . - .

- _ - - - - . - - - - - . - - -

.

.

b. Observations and Findinns

The inspectors found that the licensee's self assessment was conducted using i experienced senior reactor operators from five utilities and six different nuclear power plants. The inspectors determined that the self assessment was I accomplished with experienced personnel who perform the same or similar ,

operations activities and pro;; esses at their resr.ective plant l This self assessment determined that, overall, Fort Calhoun's operations activitlew i and processes were effective. However, the licensee identified a tntal of 27 concerns and areas for improvement. The licensee personnel documented the 27 Items on 9 conditio's reports, and 11 commitment identification documents. The 3 most significant concerns were as follows:

Work Arounds. The licensee's self assessment team determined that there i appeared to be a continued acceptance of a large number of operator work around *- Ooerations Control Center. The licensee self assessment team identified a need for increased management attention to, and support for, the personnel and activities of the operations control center ' Some examples of the self-assessment team's findings were as follows:

• When additional operations personnel are needed, there was a !

tendency for the control room staff to use operations personnel assigned to the operations control center instead of calling in- ,

additional operations personnel. The licensee stated this leaves the operations control center with a reduced, or no, staff. The self assessment recommended that operations establish an improved policy for calling in operations personnel to assist the control room staff in lieu of routinely using operations control center personne * A licensed senior operator in the operations control center did not appear to know if a procedure existed with the composite responsibilities of the operations control center staff. The self assessment recommended that the operations control center staff be provided with additional training, or guidance as to the identity and content of Procedure OPD 41 * Inconsistent Excectations of the Correctite Action Prooram. The licensee's ;

self assessment team identified inconsistent management / supervision, which had led to inconsistent understanding and implementation of the corrective action program. Two examples are presented below:

I

i l

l

.

.

Lonn Time Operator Work Arourld. During a tour of the turbine building, a self assessment auditor observed a nonlicensed operator venting Gauge PL 1913B, " Intake Tunnel Pressure Indication," prior to

'

taking reading. Afte wards, the assessor discussed this venting of the gauge prior to taking a reading with the shift supervisor. The supervisor stated that this evolution could possibly be a work around.

! The assessor initiated Condition Report 199700302, which discusy 3d

!

the appearance of a high tolerance for operator work arounds. Thc licensee stated that the root cause of the condition report was the insensitivity to activities that could be considered operator work arounds and acceptance of longstanding work practices mandated by system design or inadequacies. The inspectors noted that the one corrective action item was to discuss this example with the operators during the Cycle 4 requalification training, which was scheduled to be completed in August 15,1997. The inspectors verified that this example was presented to the corrective action group, which remarked that the example was not adverse to plant operation *

Failure to Imnlement Timely Corrective Action. The self assessmont team identified Condition Report 199600757, which documented a I plant shutdown caused by the failure of a reactor coolant pum The root cause identified that this failure was " lack of sufficient supply of lubricating oil due to Check Valve RC 285 being installed incorrectly." The assessor noted that one corrective action (e.g.,

provide training to maintenance personnel regarding the significance of this event) had not been completed. The assessor noted that the completion date was March 31,1997. The assessor noted that Fort Calhoun Station had completed a refueling outage since identification of this corrective action. The assessor stated that this action should have been completed prior to that outage. The assessor initiated Condition Report 199700304, which described the failure to implement timely corrective action. The root cause of the condition report was the lack of scheduling training prior to the 1996 refueling outage due to other manpower demands, and not recognizing the need to complete the training prior to the 1996 refueling outage. The inspectors verified that training was provided to maintenance personnel and was completed on May 14,199 c. Conclusions The inspectors concluded that the operations self assessment was effective in identifying concerns and recommendations. Operator work arounds, operations control center, and inconsistent expectations of the corrective action program were the three major concerns determined during the self assessment. The licensee was still evaluating these items for corrective action )

l

_ _ _ . ._ _ _ _ -

.

.

lk.Malntmimitt M2 Maintenance and Material Condition of Facilities and Equipenent a. LnJpfatian Scone (405001 The inspectors reviewed 27 condition reports that were assigned to maintenance for disposition. The inspectors alto reviewed several maintenance work requests to determine whether corrective actions were being inappropriately handled with this mechanism in lieu of the condition reporting process, b. Observations and Findinas The inspectors found that condition reports clearly described the actual or potential conditions adverse to quality, the apparent cause for the condition was determined, and that the corrective actions taken to prevent recurrence were appropriate to the condition reported. The inspectors also found that the maintenance work requests reviewed were appropriate for the identified problem and that conditior, reports were initiated when require c. Conclusions The inspectors concluded that the corrective action process was being satisfactorily implemented by the maintenance departmen M4 Maintenance Staff Knowledge and Performance a. insoection Scone (40500)

The inspectors interviewed four maintenance personnel to determine their perceptions of the corrective action program, b. Observations and Findinas The inspectors found that maintenance personnel responded favorably to the corrective action process. They expressed thet they would not hesitate to initiate a condition report. They indicated that the computer based system was not user friendly for them, mainly due to lack of computer skills. Maintenance personnel indicated that they had good support from their supervisors to ensue that condition reports were initiated promptly. They expressed confidence in the corrective action process and that management would apply adequate resources to correct any plant problem Conclusions The inspectors concluded that maintenance personnel were supportive of the corrective action program and confident in its effectivenes _ _ _____ ___-_ _-__-_ _ _ _ _ _ _ _ _ _ _- - _ _ .

,

,

M7 Quality Assurance in Maintenance

' Inspection Scopo (4QLQQJ

.

The inspectors reviewed the " Fort Calhoun Nuclear Station 1996 Maintenance Self Assessment Report," dated February 28,1997. The inspectors also reviewed Quality Assurance Audit Report 18, " Maintenance Activities Development of the Maintenance Rule 10 CFR 50.65," dated June 14,1996. The review elso included the scope of the self assessment and the qualifications of the plant staff and j

outside plant representatives in the self assessment. In addition, the inspectors interviewed the self assessment team leader and two team representatives to gain

their insight on the effectiveness of their effort and the responsiveness of management and staff to their finding Q)servation and Findinas The inspectors found that the results of the maintenance self assessment were positive in that no significant safety problems or concerns were identified. The inspectors found that the self assessment of the maintenance department was conducted by the licensee's sponsored team which included members from outside utilities. The inspectors found that the scope of the self assessment had performance attributes that included plant material condition, wnrk control, preventive maintenance, organization and administration, conduct of maintenance, procedures, maintenance history, materials management, maintenance facilities and 4 equipment, and maintenance personnel knowledge and performance. The inspectors found the self assessment team consisted of qualified representatives assigned to evaluate the specific attributes of the maintenance department activities. Although the licensee's team did not identify any safety significant findings they did identify 29 areas for improvement. The inspectors found that the licensee had issued seven condition reports, which addressed all areas for improvement identified in the self assessment report. The inspectors found that the self assessment team had good cooperation and response from both maintenance management and staff in response to issues raise The inspectors found that Quality Assurance Audit 18 was conducted May 6 20, 1996, at the request of the special services department. Although Audit 18 did not identify any significant safety problems, Condition Report 199000718 identified three Maintenance Rule components that were not in the Maintenance Rule log as required by Standing Order G 96. The inspectors reviewed Condition Report 199600718 and found that the corrective actions taken to revise the Maintenance Rule log were completed on July 8,1996, Conclusions The inspectors concluded that the quality assurance audit and the self assessment provided a critical assessment of maintenance activities and were effective in identifying areas for improvemen .

. . _ . _ _ . .

.

.

Ill Enoineerina E2 Engineering Support of Facilities and Equipment E Enoineerina Support of Operatina Experienga Insoection Scope f46500) '

The inspectors reviewed the licensee's operational experience feedbhck program

- to determine its effectiveness in assessing, documenting, and informing appropriate plant personnel of significant plant events in an effort to prevent their occurrence at the Fort Calhoun Station. The inspectors reviewed the Quality Procedure NOD OP 21, " Operating Expeilence Review Program," Revision 6, which defined responsibilities for performing evaluations of industry operating experience events, implementing corrective actions, issulng periodic status reports, and conducting periodic program effectiveness reviews. The inspectors reviewed 43 NRC information notices,4 NRC bulletins, and 9 INPO significant operating experience reports, which are idnntified in the supplemental information attachment to this inspection report, Observations and Findinas The inspectors determiried that the operating experience feedback program procedures provided appropriate controls for forwarding information regarding events to the appropriate licensee review personnel. The inspectors determined that corrective actions resulting from the review of information for operational events were planned, implemented, and tracked to completion through the condition report proces The inspectors noted_that, as a result of the root cause analysis of a extraction steam line break on April 21,1997, the operating experience feedback group personnel conducted an assessment of the operating experience feedback progra This assessment, completed on June 20,1997, evaluated the adequacy of the internal and external experience to Fort Calhoun Station programs to prevent incidents / events and to improve the programs. As a result of that assessment, the following changes occurred:

• Fort Calhoun Station initiated daily discussion of nuclear operating experience at the morning meeting. Previously, Fort Calhoun Station did not discuss nuclear operating experience at the morning meetin * Awareness of ownership and the need to use all available information, external, as well as internal, was increase * Industry sources of information were identified for owner, management, and self assessment us _ _ .._

.

1*

i The inspectors determined that the above improvements were occurring at the Fort Calhoun Station. The inspectors attended several morning meetings, and the operating experience group presented other nuclear plant operating events that had occurred receritly. The inspectors noted increased communications between the participants at those meeting '

,

The inspectors determined that operating experience evaluations were assigned according to a priority classification system. If the classification was urgent, then the evaluations would br, completed within 30 days, and a condition report was generated. if the classification was routine, then the evaluation would be comple.ted within 90 days, if the classification was specified as generic letters and bulletins, then the evaluation wtsuld be performed within the time frame specifie The inspectors verified that all classes of evaluations were normally completed !

within thelt time limi ! Conclusions  :

.

The inspectors concluded that the operating experience information was being disseminated appropriately. The team concluded that evaluation reviews and

- corrective actions for operating experience were being controlled. The team also noted that the licenses had identified examples of improvements of the program in a self assessment, completed in June 20,1997. The team concluded that these improvements were effective in presenting plant events information to the plant staf E2.2 Enaineerina Sunoort of Condition Reoorts IDsDeClion ScoDe (40500)

The inspectors reviewed 44 condition reports that were assigned to engineering for disposition. A list of condition reports reviewed by the inspectors is included in the attachment at the end of this repor Observations and Findinait The inspectors found that most of the condition reports were adequately addressed by licensee engineering groups. The dispositions included a good problem description, a cause determination, immediate and long term corrective actions, and a consideration of generic ramifications, where applicable. For significant issues, the licensee performed a root-cause analysis, which was evaluated by the inspectors in each case to be satisfactory. Several problems observed by the inspectors are discussed'in the following section =

7 ,

I

.

b.1 Preconditionina Concern During review of Condition Report 190001179, the l inspectors identified a concern related to preconditioning of safety related

! equipment prior to surveillance testing. The licensee had self identified the I

'

preconditioning potential of this scenario, but eventually concluded, based on the design of the circuitry and limitations in the way the valves could be tested, that no preconditioning concern existe The raw water to component cooling water heat exchangers (four total) each have a component cooling water inlet valve (489A/490A/491 A/492A) and outlet valve (4898/4908/4918/492B) controlled by a single handswitch on the main control room panel board. The safety function of these alt operated valves was to open or ,

remain open in response to an accident signal. An Inservice Testing Surveillance l Test OP ST CCW 3001, " Component Cooling Category B Valve Exercise Test,"

Revision 16, was performed quarterly to ensure that the opening function was intact and was not degrading. During the test, with the valves closed, the handswitch was taken to open and the opening stroke time for the "A" valve was recorded. At the same time, the "B" valve cycled open. Af ter both valves were closed, the handswitch was again taen to open, and the opening time for the "B" valve was recorded. The inspectors considered the manner in which this test was conducted to constitute preconditioning of the stroke time test of the "B" valve Since the stroke time test of the "A" valve was performed first during each test, the stroke time of the "B" valve was always checked a short time af ter the "B" valve had been cycled. Therefore, a stroke time anomaly for one of the *B" valves may not be detected if the pretest stroke eliminated the anomaly. The licensee stated that they were reluctant to block the "B" valve during the test since it would be taking the system out of its normal operating configuratio CFR Part 50, Appendix B, Criterion XI, states, in part, that "lal test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisf actorily inservice is identified and performed in accordance with written test procedures which !

incorporate the requirements and acceptance limits contained in applicable design documents." The inspectors concluded that the licensee's test procedure failed to establish a suitable means by which to ensure that the valves in question would perform satisfactorily inservice. This was identified as a violation of Criterion XI (50 285/9713 02). Generic Consecuences Not Considered The following were examples where the inspectors observed that the generic consequences of adverse conditions were either not considered or were not documented within the condition repor During review of Condition Report 19960771, the inspectors noted a weak

_

response during review of an inservice examination problem. The licensee had found that inspection of the pressurizer manway cover studs, as directed by Procedure PE RR RC 0103, was not consistent with ASME Section XI. According .

to ASME Section XI, the cover stud nuts required a visual VT 1 examination, but Procedure PE RR RC-0103 specified no examination of the nuts. The oversight occurred during the drafting, reviewing, and approving of the procedure in that the

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _-__

.

>

requirement for visually i.aspecting the nuts was not incorporated. The licensee took actions to perform the visual examination of the nuts. However, the licensee dici not extend the corrective actions to include a review of related inservice examination proceduret to determine whether the observed oversight was isolate In response to the inspectors' concern, the licensee reviewed the related procedures and found no similar error During review of Condition Report 199601396, the inspectors observed a weakness in the licensee's response to a misconfigured piping support, in the process of conducting a coatings inspection prior to exiting containment, the licensee found the spring can on Pipe Hanger ACH 310 not centered on the bottom of the pip The piping was part of the component cooling water system. The licensee speculated that the spring can had been mispositioned during a past work activity in the overhead. Although the licensee repaired the pipe support and evaluated its operability in the as found condition, no inspection was conducted in some vicinity to determine whether other equipment may have been damaged by tSe same work activity or activities. This concern was subsequently mitigated by the licensee's position that this pipe support was unique in its configuration, that its misconfiguration was not considered damaged, and that other misalignments in the area would probably have been noticed by the same quality control inspectors that found this proble During review of Condition Report 199600479, the inspectors identified a weakness in the licensee's response to their identification of a configuration control problem, in 1991, the licensee initiated an engineering change notice to upgrade the material used in the casing of the four raw water pumps (AC 10A/B/C/D). The manufacturer sent ASTM A 487 GR CA6NM materialinstead of the specified ASTM 743 GR CA6NM. The system engineer accepted the material, but a formal equivalency evaluation was not performed by engineerin After identifying the problem in 1996, the licensee performed the missing equivalency evaluation (which determined that the two materials were equivalent)

but the corrective action process was not expanded to consider the generic implications of the event, such as whether other contemporary engineering change notices (or those performed by the same system engineer) were similarly mishandle The inspectors identified these three examples as a weakness in the licensee's corrective action program to fully consider the generic implications of identified discrepancies. The licensee acknowledged the need for improvement in this are Motor-Operated Butterfiv Valve Overtoraue During review of Condition Report 199600507, the inspectors identified an issue for further review. This condition report addressed on overtorque condition that had occurred with Motor-Operated Butterfly Valve HCV 383 4-0, " Containment Sump Isolation Valve." This valve had been tested twice using MOVATS BARTS equipment, during which the normallimit switch controls were bypassed. Based on updated uncertainty data published by MOVATS in MOVATS User Technical Notice 96 01, the licensee discovered that the total torque applied to the actuator for these two strokes may

_ _ _ _ _ -

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - ___ ___ _ _ ____ _ __- _ ____ _ _ _ _ _ _ ______ _ _ ____ _ _ _ __-

.

.

have been as high as approximately 130 percent of the rated torque limit for the actuator, as published by Limitorque, the actuator manuf acturer. Based on previous guidance from Limitorque, the licensee would be required to open and inspect the actuator for indications of damage on torque bearing components. However, in an effort to avoid an operational impact, the licensee obtained from Limitorq;n an exemption from these guidelines for the two strokes at 130 percent of the ratin Previously, limitorque had granted an exemption for the same valve allowing 100 strokes at 125 percent of the torque rating, but based on the information in the MOVATS User Technical Notice 90 01, an additional exemption was needed. The licensee stated that the actuator for Valve HCV 383 4 0 would be inspected and overhauled during the Spring 1998 refueling outag The inspectors could not determine the bases under which Limitorque granted the exemptions, (i.e., were they test-based, analytically derived, or based on engineering judgments). In addition, the inspectors noted that while this condition was identified in April 1996, the licensee did not address the concern during the Fall 1996 outage. The licensee did not approach the exemption from the perspective of a short term basis for continued operation, but rather as a permanent exemption from the inspection requirements. The inspectors considered that the nature of the exemption was such that an inspection at the earliest available opportunity would have been prudent. The licensee stated that the refurbishment of HCV 383 4-0 had originally been part of the Fall 1996 outage scope but had been removed when schedular pressures intervened. The licensee stated that this task had received high priority for the Spring 1998 refueling outag The inspectors identified these issues as an inspection followup item (50 285/9713-03), specifically to review the Limitorque exemptions both in a specific and generic sense, and to verify that the licensee performs the inspection and refurbishment of HCV 383 4-0 during the Spring 1998 refueling outage, Lack of Cause identification - Condinon Report 199601251 was written following a f ailure of Valve HCV 489A, " Component Cooling Heat Exchanger AC 1 A, Component Cooling Water inlet Valve," to open during the performance of Surveillance Test OP ST CCW 3001. The failure rendered raw water component cooling water heat exchanger AC 1 A inoperable. In the investigation, the licensee could not identify the cause of the failure or make the opening failure recur. All subsequent stroke demands were successful. As a consequence, no corrective actions were taken. The inspectors reviewed the stroke time trend for this valve and did not observe any indication suggesting that the valve was approaching f ailure. The inspectors considered the lack of cause identification to be a concem because of the safety significance of the valve's failure to open (the loss of a raw water / component cooling water heat exchanger), The inspectors considered this to be an example of weak engineerin . . _

.

_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

.

. Conclusions Engineering performance in the disposition of condition reports was generally goo However, several problems were noted, including one instance whers a precondition concern was not addressed, three instances where generic implications were not adequately considered, and one instance where the cause of an event was not determined. The inspectors also identified a concern related to the manner in which l

the licensee and the vendor handled a butterfly valve overtorquing even E2.3 Enaineerina Sucoort of Enaineerina Assist Reauests Inspection Scoce (40500)

The inspectors reviewed the engineering assist requests listed in the attachment to assess engineering's performance in supporting operation Observations and Findinat, For the most part, the inspectors considered engineering to have performed wellin assessing the issues addressed by the engineering assist requests. Each was a comprehensive analysis of the technical concerns, with a satisfactory level of detail explaining the bases under which the various determinations were mad Within Engineering Action Request 06166, " Replacement Solenoid Valves for FCV 1904A/B/C," the following statement was made: " Based on the information l collected it appears that the 4 way solenoid for FCV 1904C valv6 has been replaced since the original plant startup, but no engineerhg change notice or drawing change was processed to document the change." The licensee issued an engineering change notice to correct this deficiency, but did not initiate a condition report, as required by Procedure SC R 2, " Condition Reporting and Corrective Action," Revision 4, to activate the full scope of the corrective action process, 10 CFR Part 50, Appendix B, Criterion V, states, in part, that " activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings."

Procedure SO R 2 requires a condition report if an alteration in configuration was found, but was not provided for by an authorized change process, such as a modification engineering change notice, or temporary modification documen The failure to initiate a condition report was considered as one example of a violation of 10 CFR Part 50, Appendix B, Criterion V (50 285/9713 04).

_ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ ______ _ _ .

. Cpnclusiqnli licensee engineering exhibited good performance in the disposition of the engineering assist requests reviewed by the inspectors. One example was identified where a condition adverse to quality was not reported on a condition repor E7 Qua!ity Assurance in Engineering Activities E Quality Assurance Audits and Self Assessments IDsoection Scone f40500)

The inspectors reviewed one self assessrnent related to engineering, " Configuration Control Self Assessment Final Report," dated April 11,1997. This was the only self assessment performed by the licensee in the area of engineering in the past 2 years. The inspectors alsu reviewed several engineering related audit and surveillance reports as listed in the attachment, Observations and Findinas Within the configuration control self assessment, the licensee reviewed procedures associated with configuration control, reviewed relevant condition reports, performed in plant walkdowns against plant drawings, and reviewed applicable industry experience in this area. The licensee determined that mostly minor configuration problems existed and that most of items were pre existing problems that have been recently identified because of increasing sensitivity to configuration control issues. Several recommendations were made to enhance procedures for better overall control in this are The inspectors noted that the licensee had not initiated any condition reports as a direct consequence of the self assessment. Instead, commitment identifications were generated to addresu vrarious observations and weaknesses. The licensee stated that commitment identifications were used exclusively to track recommendations and enhancements, but were not used for conditions adverse to quality. The inspectors considered at least one of the findings of the self assessment to be a condition adverse to quality, that being several procedural inadequacies that allowed for configuration inconsistencies to exist for short periods of time both before and after field installation of a modification. The procedures in question were S0-G 21, " Modification Control," Revisun 62, and PED gel 56,

" Configuration Control Closeou'," Revision 4. WitNn Procedure S0-0 21, the licensee identified that certain documents were not required to be updated until modification completion, which is often several months after the modification becomes operable. Conversely, no clear restrictions existed to prevent a procedure, drawing, or other document from being submitted, processed, and updated prior to the configuration change being made operable. In Procedure PED gel 56, the licensee identified that, except for drawings submitted to drafting, drawings

_ _ _ . - _ _ _

_ -

-- _-

_ _ _ _ _ _ _ _ _ _

_

-0 received no engineering review upon issuance, in addition, the licensee found that revisions to other configuration related documents, such as fuse control and breaker programs were submitted prior to modification closcout. The inspectors considered this to represent a condition adverse to quality, for which the licensee should have initiated a condition repor CFR Part 50, Appendix B, Criterion V, states, in part, that " activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accornplished in accordance with their instructions, procedures, or drawings."

Procedure S0 R 2, " Condition Report and Corrective Action," Revision 4, requires origir'ation of a condition report for the following:

"

. . . equipment related events, documentatloa, deficiencies, non routine outside agency notifications, operational events, testing deficiencies, security infractions, human performance errors, personnel safety issues, radiological l occurrences, or other circumstances which impact or potentially impact safe /or reliable operation of Fort Calhoun Station . . . . "

The inspectors considered this issue to constitute a circumstance for which a condition report was required since the safe operation of the plant could potentially be impacted. This was considered a second example of a violation (50 285/9713 04) of 10 CFR Part 50, Appendix B, Criterion V, for failure to follow Procedure S0 R The inspectors found that the joint utilities management audit, conducted December 1996, and the safety audit and review committee audit, conducted May 1997, had both identified similar t. blems in the corrective action process involving overdue responses or actions to condition reports. For example, Joint Utilities Management Audit Finding 96 CA 6F documented continuing violations of Standing Order R 2 condition report response timeliness requirements with overdue responses to condition reports. The safety audit and review committee audit documented problems in Condition Report 199700714 associated with the timely closure of condition reports. The inspectors were concerned that the licensee had not taken appropriate corrective actions to reduce the number of overdue condition reports since first identified by the joint utilities management audit. The inspectors reviewed information regarding the total number of overdue condition report responses from October 1995, through June 1997. The inspectors found that the average number of overdue responses to condition reports prior to the Joint Utilities Management Audit was five per month. The inspectors found that, during the period January through June 1997, the average number of overdue responses to condition reports was one per month. The inspectors found that the licensee's goal was to have zero overdue responses and that this was what the safety audit and review committee audit was emphasizing in their report to the licensee. The inspectors noted that during this inspection the total number of condition report overdue responses was zero,

. Conclusions The licensee had adequately assessed its engineering processes over the past ,

2 years. Within an audit of configuration control, the licensee identified a condition '

adverse to quality, but failed to initiate a condition report in accordance with procedure. The inspectors also concluded that the licensee had taken effective cortcetive actions to reduce the number of overdue responses to condition report E7.2 Interview with Desian Enaineers Insnection Scone (40500)

The inspectors interviewed two design engineers concerning their interface with, and impressions of, the corrective action program, Observations and Findinas The inspectors found the engineers were involved in responding to action items that were generated from condition reports and that both had gene.ated new condition reports on their own. They considered the threshold for writing condition reports to be low and were never discouraged from writing a condition report. They had ample time to comprehensively address the condition report action items assigned to them. In general, the engineers had positive impressions of the corrective action program, Conclusions The inspectors concluded that engineers were supportive of the corrective action program and confident in its effectivenes E7.3 Boot-Cause Analvsis insoection Scone (405001 For the condition reports listed in the attachment to this inspection report, the inspectors reviewed the root-cause analyses performed by the licensee.

! Observations and Findinas Overall, the inspectors found the licensee's root cause analyses were technically

! thorough and broad scoped. The identified root causes and the corrective actions taken to prevent recurrence were appropriate with the reported facts, Conclusions I

'

The inspectors concluded that the licensee was effectively implementing a good root-cause analysis process.

.

b

l I

e---r,- r --

,,. -

- - . . . - , - - - , , --- ----

. , , , , , . - , n- -

, - - - - -,,

-_

o

l

!

E8 Miscellaneous Engineering issues 192903)

E8.1 iClosed) Inspection Followup Item 285/9405 01: Toraue Switch Repeatability for Total Thrust Measurefnent

' Backaround The licensee had not accounted for torque switch repeatability within its analysis of the maximum thrust applied to a motor operated valve during operation. The repeatability of the torque switch, most often taken as 5 percent, can range as high as 20 percent for certain actuator sizes and torque switch settings. Therefore, although a given diagnostic test may have shown a motor operated valve's maximum thrust to have been within limits, when stroked again without diagnosdes, the thrust may have exceeded these limits without detection. In situations where actuator torque was calculated using the thrust measurement, the same issue applied to the analysis of maximum torque. Because the torque switch was always bypassed in the opening direction, this issue applied only to the closing strok I Insoector Followgg The licensee reviewed the calculations for all safety-related motor operated valves and detarmined that application of torque switch repeatability to the total thrust measurement did not result in any overthrust or overtorque conditions. The licensee revised Calt,ulation FC06203, "MOV Testing -Instrument Errors, Setpoint Calculation," Revision 1, to include within the maximum closing thrust and maximum closing torque analyses an error associated with torque switch repeatability, Conclusions The inspectors reviewed Calculation FC08203 and observed that torque switch repeatability had been included within the error applied in the calculation of maximum closing thrust and torque. The inspectors concluded that the licensee had adequately addressed the immediate and long term implications of this issu VI. Manaaement Meetinos X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 18,1997. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ .

.

.

ATTACHMENl SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee J. Adams, Design Engineer R. Andrews, Manager, Nuclear Assessments

,

M. Bare, System Engineer J. Chase, Plant Manager M. Core, Manager, System Engineering T. Daily, System Engineer T. Dukarski, Supervisor, System Chemistry M. Ellis, Supervisor, Maintenance Support C. Fritts, System Engineer l J. Gasper, Manager, Nuclear Projects 4 G. Gates, Vice President, Nuclear K. Hyde, Senior Nuclear Design Engineer L. Kusek, Alternate for Manager Quality Assuranc;/ Quality Control C. Lloyd, Relief Valve Program Engineer

'

E. Matzke, Station Licensing J. McManis, Supervisor, Mechanical Engineering R. Mehaffey, Principle Electrical Engineer B. Mierzejewski, System Engineer D. Motzer, Design Engineer R. Phelps, Manager, Station Engineering R. Plath, Supervisor, Electrical Design A. Richard, Supervisor, Mechanical Systems S. Resch, Motor Operator Relief Valve Program Engineer J. Spilker, Operations Engineer M. Tesar, Manager, Corrective Action J. Tills, Manager, Nuclear Licensing D. Trausch, Manager, Nuclear Safety Review K. Woods, Nuclear Design Engineer NflQ W. Walker, Senior Resident inspector ITEMS OPENED, CLOSED, AND DISCUSSED Onened 50 285/9713-01 VIO Failure to Maintain the Senior Vice President as Chairperson of Safety Audit and Review Committee in Accordance with Technical Specification

, _ , _ _ _ ._

.

.

50-285/9713 02 VIO Inadequate Test Procedure 50-285/9713 03 IFl Motor Operated Butterfly Valve Overtorque Concerns r0-285/9713-04 2 VIO Failure to initiate Condition Reports Clos /9405-01 IFl Torque Switch Repeatability for Total Thrust Measurement 50-285/9713 01 VIO Failure to Maintain the Senior Vice President as Chairperson of Safety Audit and Review Committee per Tech Spec LIST OF DOCUMENTS REVIEWED Plant Procedures Procedure Revision Iltlg Number SO-R-2 4 Condition Report and Corrective Action SOR3 12 Reportable Occurr6nces

"

SO R 11 32 Notification of Significent Events S0G5 121 Fort Calhoun Station Plant Review Committee S0 G 94 7 Material Nonconformance Control SO-G 111 2 Performance of Solf Assessments SO O 1 32 Conduct of Operations OPD-416 2 Operations Control Center Description OPD-417 4 Control Room Deficiencies and Operator Work Arounds" NOD-QP 3 17 10 CFR 50.59 Safety Evaluations NOD OP-19 18 Root Cause Analysis Program NOD OP-20 7 Human Performance Enhancement System Program NOD OP-21 6 Operating Experience Review Program

>

---

.

.

NOD QP 22 11 Preparation and Approval of a Safety Analysis for Operability (SAO)

NOD OP 23 10 Commitment Action Tracking System (CATS)

NOD-QP-31 12 Operability and Reportability Determinations l

Conditions Reports Condition Reoort N IWg Enoineerina q

19PbOOO94 Trip Coil Relay Failed to Trip  !

I 199600196 Three Filet Welds Cracked 199601082 Degraded Circuit Breaker 199601179 AOV Failed to Close 199601217 Part 21 on Circuit Board 199600774 Check Valve Installed in Reversed Orientation 199600771 Pressurizer Manway Cover Stud Inspection Does Not Meet ASME Requirements 199700036 Packing Cooling Pump Discharge Low 199600060 Raw Water Flush Line Clogged 199700255 Inspect and Repair Boric Acid Pumps 199500011 Problems with 10 CFR 50.59 Evaluations

'199500065 Seven Pressure Control Valves Out of Tolerance 199500099 Motor Operated Valve Failed to Open 199500105 Diaphragm Valve Failure 199500221 Same Part Number not Like for Like 199500285 : Failure to inspect Piping Segment-

'

-,

.

199600103 Valves Removed Under Maintenance Work Order Instead of Construction Work Order 199601211 A0V Manual Operator Gearbox Housing Damaged 199500075 SOV not Replaced on Schedule 199500113 Relief Valve out of Tolerance 199600385: Wire Burnt in Circuit Breaker 199600414 AOV Failed to Open 199601018 Overcurrent Trip Device-199601251 AOV Stuck in Closed Position 199601499 1989 ASME Section XIinformation not included in Surveillance Test Procedure 199700042' ice Formation in Condensate Storage Tank 199500192 Relay Coil Circuit Wiring incorrect 199600104 Circuit Breaker Tripping Problem 199600507_ Increased MOV Test Equipment Inaccuracles 199600654 Valve Excessive Vibration 199601024 Check Valve Discrepancies-199601273 Pressurizer Relief Valve Out of Tolerance 199601285 Main Steam Line Relief Valve Discharge Stack Wald Cracked 199601326 Relief Valve Failure 199601444 - No Seismic Qualification for Racked Out Circuit Breaker 199600062 High Energy Line Break Concerns 199600296 Seismic Restraint not Properly Mounted on Wall 199600479 Wrong Material Raw Water Pumps Bowl Assemblies 199601007 Mismatch Between Charging Pump Flow Rate and Safety Analysis

. .

.

.

O

,

199601396 Spring Can Not Located Properly 199601419 Snubber Interference 199601443 Steam Migration Through Ductwork 199700134 Error in Surveillance Test Setpoint Calculation 199700232 Error in Heatup Rate Maintenance

!

199500030 Motor driven fire pump FP 1 A failed full flow test 199500042 Out of position hand hole cover 199500082 Diesel driven fire pump FP 1B failed to meet ac::eptance criteria 199500058 Leak rate test limit exceeded 199500446 Check valve inservice test pegged high 199500119 Undersized limit switch washer 199500091 Acid regenerant pump 2 isolation valve fracture 199500220 Auxiliary feedwater pump FW 10 liquid penetrant test 199600011 Wild roster does not indicate welder qualifications 199600025 Wrong pipe material welded to feed Pump FW 4C suction drain valve 199600166 Walds made cyclone separator CW 5 drein valve by unqualified welder 199601372 HCV 318 breaker would not pullin at required voltage 199601402 Failure to return weld rod 199600856 HCV 400C valve stroke time above required range 199600165 Raw water head shaft nuts made of wrong material 199600304 Hoist not inspected prior to use 199600397 Steam generator conductivity rising 199601592 Disabled post accident sampling system valves 199601545 Incorrect radiation monitor setpoints 199600066 Incorrect end cap position on valve operator 199600256 Excessive water running from end bell of condensate cooler 1J9601079 Misinterpretation of fuel off-load bridge and trolley coordinates

{

199700134 Survoillance test error i 199700138 New fuel storage rack enrichment concerns 199700040 Condensate pump motor shaft run out issue ,

199700019 Engineering evaluation not performed on replacement valves l Operations 199601148 MS 265 found open and no danger tag attached 199601334 On 10/28/96, intake tunnel was found full of water 199601354 Two warning tags were found on floor 199601592 I&C tech informed SS that two PASS valves were disabled 199601346 Fuses were pulled on YCV 1045 to maintenance activities 199601203 LSO failed to provide timely direction to Ros

l l

_ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ .

.

. ..

.

.-

'

.199601052 H2 cooler outlet temp alarms Hi-Hi are sr : at 50 deg 199601342 Lo an operable boric acid flowpath in > the RCS 199600233 CA 1 A af tercooler war. not functioni 199600234 Compensatory measures were se- Annun was not operable 199600757 Control room entered AOP 5, rea, t coolant pump vi' .ating

'

199600251 Preo and planning not well coordint 'd for the job 199700079 Monthly ST lockout relay failed to tr.. on demand 199700479 A NOUE due to a leak in the CCW system /

199700602 Two examples of work without complying with S0-G 20A

!

199700301 Clearance tag hanging on breaker for FW 52 199700384 15 CRs written on same day for Configuration control 199700210 Waste decay tanks A&C were slowly depressuring l 199700307 No equipmnt out of service log as required by Maint rule 199700306 OCC performance needs improvement 199700305 Occasions to improve operator performance were not exploited 199700301 Clearance tag on breaker for FW 52

,

l 199700300 No turnover signs posted on contrni room entrances 199700299 Access to control room not strictly enforced 199700030 Containment Isolation valves do not have accumulator

!

Miscellaneous L

I 199601598 Circuit drawings not marked to indicate open modification '

199700079 Lockout relay failed to trip on demand ,

199700098 Update steam generator blowdown drawings >

199700106 DC circuit problems 199700111 Rockshaft for HCV 1042A received without set screw holes Engineering Assistence Request FEAR)

EAR No lijlg 95-033 Diesel Generator Temperature Control Valve 95 094 Containment Spray Pump Piping 95 071 Removal of Steam Generator Orifice Plates 96 032 Evaluation of CCW Heat Exchangers Post DBA Performance with 5 Percent Plugged Tubes 96 134 Evaluation of Hydraulic Snubber inspection Results 95 063 Clarification of Design Basis for HCV-1103/04 96 068 Covering RCS, CVCS Openings After Removal of Equipment

,

.

e 90 102 Maximum Allowable Stroke Times 96166 Replacement Solenoid Valves for FCV 190A/B/C 96 168 CA 288/289/200 Globe Valve Replacement 96 178 HCV 1040 Actuator Modification 97 092 Replacement Springpack for Limitorque Operator SMB 00 Oberability Determinations Reviewed NOD OP 31 for Condition Report 199700523 NOD QP 31 for Condition Report 199700570 NOD-OP 31 for Condition Report 199700652 NOD OP 31 for Condition Report 199700664 nob-OP 31 for Condition Report 199700732 NOD Olb.$1 for Condition Report 199700752 NOD CP 3 for Condition Report 199700818 l

Wore, Docume its MWR 9602195 MWR 9603400 MWR 9700399 MWR 9700424 MWR 9700465 MWR 9700493 MWR 9700602 MWR 9700902 MWR 9701268 MWR 9701384 MWR 9701397 MWR 9701457 MWR 97C1615 MWR 9701719 MWR 9701908 MWR 9700723 Miscellaneous Docaments Safety Audit and Review Committee Charter, Revision 23, dated March 21,1997-Ooeratina Exoerience Review Information Notice 97-01, " Improper Electrical Grounding Results in Simultaneous Fires in the Control Roorn and the Safe Shutdown Equipment Room"

o

'

,e information Notice 97 09, " Inadequate Main Steam Safety Valve Sotpoints and Performance issues Associated with Long MSSV Inlet Piping" Information Notice 9712. " Potential Armature Binding in General Electric Type HGA Relays'

l information Notice 97 25, " Dynamic Range Uncertainties in the Reactor Vessel i

level Instrumentation" Information Notice 96-03, " Main Steam Safety Valve Setpoint Variation as a Result of l

Thermal Effects" j

information Notice 9615, " Unexpected Plant Performance During Performance of New Surveillance Tests" i Bulletin 96 02, " Movement of Heavy Loads over Spent Fuel, Over Fuelin Reactor Core, or over Safety Related Equipment" INPO Significant Event Report 0196, " Transformer Explosion and Loss of Offsite Power" INPO Significant Event Report 03 96, " Failure to Perform Reactor Scram and Turbine Trip When Test Limits were exceeded" INPO Significant Event Report 09 96, " Interrupted Control Rod Insertion on a Reactor Scram as a Result of Inappropriate Personnel Action" INPO Significant Event Report 14 96, " Operation with Reversed NIindications" INPO Significant Event Report 03 97, " Inadvertent Nitrogen Leak into Reactor Coolant

_

_ system Results in Lowered Reactor Vessel Water Level" -

Safetv Related Ooerator Work Arounds

43 A/B CIAS Override Switches Condition Report 199700607 Expected Resolution date: August 1,1997 This OWA requires restricted operation of steam generator blow down isolation valves in post accident scenario, until dose assessment is resolved 2 -MS 291 & 292 Ops Memo 97 06 & ( .ndition Report 199700744 Expected -

Resolution date: July 18,1997 Thw OWA states that MS 291 and 292 may not work below 750 psig 3 Potential N2 pockets in LPSI header Ops Memo 97 05 & EAR 97-155 Expected Resolution date: July 30,1997 This OWA states to maintain LPSI header pressure

> 210 psig by cycling valves or pumps if needed 4 Potential Raw Water Pump Sanding Problems-FC OPS 104 97 & Condition _ Report 199700655-Expected Resolution date: July 31,1997 This OWA states that

d

.

.

e operations rotating raw wa'ar pumps twice per week until further notice from system engineering 5 Fire pumps are experiencing sanding problems Condition Report 199700655-Expected Resolution date: July 31,1997 This OWA states that operations perform monthly fire pump ST twice per month for Fire Pump 1 A until 7/24/97 6 Inadequate Ventilation Design during design basis tornado in switchgear and battery rooms SE PM AE 1001 & Condition Report 199600787 Expected Resolution date:

August 1,1997 This OWA requires that operations must block open numerous fire doors and post a fire watch during a tornado watch 7 Tripping breakers with 69 switches in AOP 06-Ops Memorandum 97 07 &

Condition Report 199700772 Expected Resolution date: August 1,1997 This OWA require tripping breakers with 69 switch rather than pushbutton during ;

AOP 06 i 8 Containment Spray Pump recirculation valves locked closed EAR 95111 Expected Resolution date: April 30,1998 This OWA requires that the CE containment pressure analysis must be redone, due to pump instrument uncertainty issue 9 HCV 133 is leaking by valve and requires rebuild MWO-970349 Expected Resolution date: 98 RFO This OWA requires the control room operator must log the downstream pressure and report values to Supervisor Operations for evaluation 10 CCW pressure must be maintained > 34 psig MR FC 97 007 Expected Resolution date: 98 RFO This OWA requires the EONA to monitor / log the CCW tank pressure, _

LO must log once per shift, various administrative controls / caution tags on 6 valves

-- 1 1 On a post D8A reset of CIAS, PCV 2909,2929,2949, & 2969 may open, diverting flow to RCDT MR FC 97 004 Expected Resciution dste: 98 RFO This OWA reouires guidance be added to applicable EOP/AOPs to prevent the potential diversion of SI flow 12 Inadequate NPSI for LPSI pump under certain accident / single failure conditions MR-FC 9718 Expected Resolution date: 99 RFO This OWA requires under certain conditions, throttling of LPSI valves required if only one pump is operating Audits Reviewed Quality Assurance Audit Report 68, " Station Engineering," July 12,1995 Joint Utilities Management Audit of the Correct Action Program, December 16,1997-SARC Audit Report 45, " Corrective Action," June 26,1997 Surveillances Reviewed E 95 3,'" SRI ECN 10 CFR 50.59 Evaluation," October 18,1995

. E3 951, "MOV Program," May 9,1995 i

~

_ = _ _ l

--

e G-

.

i E4 951, " Modifications During Outages," March 13,1995 EO-951, " Design Basis Document / Drawing Venfication," March 8,1995 E6 95 2, "DDD/ Drawing Control," June 7,1995 E7 95 3, " Station Engineering," December 14,1995 E 96 2, "lNPO SOER Disposition Review," April 9,1996 E3 961, "MOV Program," December 20,1996 E4 961, " Modifications During Outages," December 30,1996 E 97 2, " Calculation Update," April 29,1997 E6 971, "DBD/ Drawing Control," March 5,1997

!

10