IR 05000423/1997082

Insp Rept 50-423/97-82 on 980209-20.Violations Noted.Major Areas Inspected:Corrective Actions Processes

ML20249B085
Person / Time
Site:	Millstone
Issue date:	06/11/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20249B079	List: IR 05000423/1997082 ML20249B078 ML20249B081
References
50-423-97-82, NUDOCS 9806220054
Download: ML20249B085 (59)

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U.S. NUCLEAR REGUI.ATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION SPECIAL PROJECTS OFFICE Docket N Report N License N NPF-49 Licensee: Northeast Nuclear Energy Company P. O. Box 128 Waterford, CT 06385 Facility: Millstone Nuclear Power Station, Unit 3 Inspection at: Waterford, CT Dates: February 9 through February 20,1998 Inspection Team: John T. Shedlosky, Lead, Senior Reactor Analyst, Region i Norman J. Blumberg, Special Projects Office - Region i Edward J. Ford, Quality Assurance and Maintenance Branch (HQMB), Division of Reactor Controls and Human Factors (DRCH), Office of Nuclear Reactor Regulation (NRR)

Robert M. Latta, NRR, DRCH, HOMB Richard A. Rasmussen, Senior Resident inspector - Maine Yankee Garmon West, Jr., Human Factors Assessment Branch, DRCH, NRR Donald A. Beckman, Consultant James C. Higgins, Consultant, Brookhaven National Lab Approved by: Jacque P. Durr, Chief Inspections Branch Special Projects Office, NRR I

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TABLE OF CONTENTS l

EXECUTIVE S U MM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

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Actions.........................................................1

t 1.0 MANAGEMENT PROCESSES AND SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i

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1.1 - Management Directions, Goals, and Expectations . . . . . . . . . . . . . . . . . . . 1 1'.2 Organizational Communications and Teamwork . . . . . . . . . . . . . . . . . . . . 2 i 1.3 Encouragement of Problem Identification by Managers and Supervisors . . . . 3 -

l 1.4 Performance Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . .. ....... 8 1.5 Management's Commitment To Resolve issues .................... 8 2.O CORRECTIVE ACTIONS .........................................12 2.1 Corrective Action Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.2 Corrective Actions - Classification and Root Cause Analysis .......... 14 2.3 Corrective Actions Effectiveness .............................16 3.0 SELF- ASSESS M ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.1 Self-Assessment Program ..................................24 3.2 Self-Assessment of Design Basis issues - ACR 7007................ 27 4.O INDEPENDENT OVERSIG HT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.1 Effectiveness of Nuclear Oversight - Audits and Evaluations ..........34 4.2 Effectiveness of Nuclear Gversight - Quality Control . . . . . . . . . . . . . . . . 41 4.3 Followup of Previously Identified inspection Findings - Nuclear Oversight . 43 4.4 Performance of the Nuclear Safety Assessment Board .............. 45 4.5 Performance of the Plant Operations Review Committee . . . . . . . . . . . . . 46 4.6 Performance of the Site Operations Review Committee . . . . . . . . . . . . . . 47 4.7 Performance of the Independent Safety Engineering Group ...........48 i

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EXECUTIVE SUMMARY )

Millstone Nuclear Power Station 1 Corrective Actions Team inspection 50-423/97-82 {

Operations - Corrective Actions Processes i

Management Processes and Systems ]

• Management's long-term direction of plant personnel is incomplete because the strategic plan (which represents long-term direction) was in draft form. However, the plant staff's clear understanding of management's expectations was considered a management strength (Section 1.1).

• Organizational communications and team work were adequate, with vertical communications between plant organizations considered a strengt Communications between groups and departments in formal meeting settings, such as the daily morning meetings, showed a questioning attitude and command and control by senior managers. Good team work initiatives had been introduced through the first line supervisor level, but not to the overall work force. The team noted ineffective communications occurred between certain elements of the maintenance and oversight organizations. (Section 1.2).

• Plant management was effective in its efforts to encourage plant personnel to identify problems and the plant staff feels that management is receptive to problems brought forward. Individuals generally characterized the environment as improved and currently receptive to problem identification. There is no reluctance or reservation expressed by individuals to identify problems (Section 1.3.1).

• The licensee is adequately responsive to specific harassment, intimidation, retaliation or discrimination (HIRD) case needs. The Employee Concerns Program, the Employee Concerns Oversight Panel and the Safety Conscious Work Environment programs are positive contributions to the overall process (Section 1.3.1). However, NU management has not been fully effective in dealing with trends and common causes for HIRD allegations generated organization-wide to Employee Concerns Program. The Safety Conscious Work Environment processes have not yet been formalized (Section 1.3.2).

• The performance monitoring program was good. The high number of human errors was a weakness that the licensee needs to examine further. (Section 1.4).

• Management's commitment to resolving safety committee recommendations, audit findings, assessment findings and open issues was satisfactory. (Section 1.5)

Corrective Actions

• Overall, the corrective actions program is functioning adequately, but it is clear that the program will continue to require careful monitoring by NU management to ensure sustained performance. The team's general conclusions regarding the iii

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adequacy of the licensee's corrective action program were that the program elements concerning identification, Condition Report (CR) initiation, and CR l processing were performing well. The threshold for identification of issues, deportability reviews, and the assignment of severity level and corrective actions were timely and appropriate. The CR program elements concerning root cause, corrective actions, and failure recurrence were considered to be operating at an acceptable level, but with room for improvement. However, the team found a notable number of process discrepancies, including the boric acid transfer pump air binding issue, in a relatively small sample size of CRs, after NU had completed their own extensive self-assessment preparing for this team inspection. For example:

1) The Corrective Actions Program lacked controls over combining similar Condition Reports such that it was not effective in retaining issue descriptions and significance level. This is considered a program weakness (Section 2.1).

2) Several CRs were inappropriately classified at a lower Significance Leve There were examples of root cause analyses that were narrowly focuse The analysis quality has improved through 1997, but overall performance is still mixed (Section 2.2).

3) Some corrective actions were also narrowly focused which missed the opportunity for the process to detect additional existent problems. A CR documented that a Site Operations Review Committee review of an Independent Safety Engineering Group procedure was not accomplished in accordance with technical specifications. This is a violation (Section 2.3).

4) CRs, which were written for several failed NRC commitments, were not classified appropriately and is a violation (Section 2.3.3).

• Root cause determinations and corrective actions for recurrent boric acid transfer pump problems were inadequate. The deportability evaluations were incomplet Operating experience was not considered. A potential unreviewed safety question resulting from non-conservative boric acid tank level technical specification was not identified. The failure to correct the air binding of the boric acid transfer pumps is a violation (Section 2.3.1).

• Root cause determination and corrective actions for recurring problems with the overload magnetic trip setting for 480 Volt molded case circuit breakers was considered to be acceptable. (Section 2.3.2)

• The licensee's failure to conduct five fire protection surveillance and exclusive reliance on the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> roving fire watch were inappropriate fire protection compensatory measures. (Section 2.3.4)

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• The self-assessment program was being adequately implemented and the associated recommendations were beneficial in identifying areas for enhancement and improved performance. An Automated Work Order (AWO) associated with a modification was closed prior to completion of all specified work; this is a violation (Gection 3.2)- l

• Acceptable processes for Final Safety Analysis Report change control are being l_

applied (Section 3.2).

• A Condition Report Action Request concerning review of Design Change Notices I was closed without accomplishing the specified corrective actions. This is a j violation (Section 3.1).  !

• The Master Setpoint List was found to contain incorrect information for one annunciator panel alarm. Additionally, the methods for control and documentation of setpoint information appeared inconsistent, difficult to retrieve at times and had the potential for allowing incorrect information to persist. The findings of an engineering self-assessment of this area has not been addressed to date. This is a follow-up item (Section 3.2). ,

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• Violations were identified concerning Design Basis Summary documents and the l

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Safety Functional Requirements Manual (Section 3.2).

• Updates of 3,000, Category 2A drawings are planned to take two years. This drawing category has a 90-day guideline for incorporating outstanding changes i (Section 3.2).

• SIL ltem No. 41, concerning the ACR 7007 - Event Response Team findings is closed (Section 3.2).

Independent Oversight

• Nuclear Oversight was effective in performing audits, general plant oversight, and work surveillance activities. Considerable improvement was noted since independent assessments identified weaknesses two ycars ago in the performance of QA activities (Section 4.1).

• Procedures and audits have improved and there is good control in the follow up of audit findings (Section 4.1). Audit findings are stronger and there is good controlin the follow-up of audit findings (Section 4.1).

• - Audit scheduling has, improved bJt there are still weaknesses that have to be resolved (Section 4.1).

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• - zThere is much better communication between the line organization and nuclear oversight (Section 4.1).

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• . Quality Control was generally effective in performing the required in-plant inspections. The QC support group was effective in establishing and standardizing

' the use of OC hold points in work packages (Section 4.2),

• SIL ltem No. 73, concerning audits of the technical specifications, is closed (Section 4.3.1).  !

• 'SIL ltem No. 73, concerning the effectiveness of the Nuclear Oversight Organization, is closed (Section 4.3.1).

• SIL ltem No. 41, concerning trending of Nonconformance Reports (NCR) reports, is closed (Section 4.3.2).

l '* The Nuciaar Safety Assessment Board (NSAB) was effective in reviewing activities on-site and identifying potential nuclear safety issues. The implementation of the NSAB met the technical specification requirements. Flowever, the 1997 resolution of an issue of membership qualifications did not meet technical specifications as presented. This was a weakness in the implementation of technical specification - - section 6.0 requirements (Section 4.4).

I e .The Plant Operations Review Committee (PORC) and the Site Operations Review l Committee (SORC) were effective in accomplishing the reviews required by )

technical specifications (Sections 4.5 and 4.6). l

• The Independent Safety Engineering Group (ISEG) has made considerable progress, albeit at the expense of performing safety reviews,in reducing the backlog of operating experience (OE) reviews. The backlog of OE reviews on Unit 3 has been reduced from several hundred to approximately 40. However, the number of ISEG ],

reviews done in 1997 was only 12, down from 24 the previous year (Section 4.7).

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Report e 3 tails I. Operations U3. 01 Quali:y Assurance in Operations - Corrective Actions (40500)

1.0 MANAGEMENT PROCESSES AND SYSTEMS The team evaluated the processes and systems that Millstone Unit 3 managers use to identify, correct, and prevent problems. The team reviewed management directions, goals, and expectations; organizational communications and teamwork; managerial and supervisory encouragement of problem identification; and performance monitoring. The inspectors conducted interviews, attended meetings, and reviewed licensee document The interviews were with all levels of plant personnel, including senior managers, middle managers, supervisors, and nonmanagerial personne .1 Management Directions, Goals, and Expectations a. Inspection Scope The team assessed the effectiveness of the process by which Northeast Nuclear Energy Company management provides the necessary direction to prevent problems to the plant staff. The team evaluated the organization's high-level goals and expectation b. Observations and Findings i

Licensee management has used numerous means to communicate and reinforce its  !

. expectations of plant personnel. The various means included " Nuclear Group Policies and j Standards;" a daily newsletter, which publishes items of interest, including operating i experience items; posters enumerating management's expectations, which were visible throughout the plant; and daily, weekly, and monthly meetings. The information covered in daily meetings included condition reports, emerging issues, plant status, management expectations, organizational changes, plant modifications, and priorities. The team observed that daily meetings showed good interdepartmental interactions, a questioning attitude by participants, and command and control by senior manager l The interview results showed a common understanding of management's expectations pertaining to identifying and correcting problems and identifying and addressing safety issues. Plant personnel at different levels of the plant organization demonstrated a high degree of compliance with areas of management's expectations, especially in the area of problem identification, j Licensee management's most formal direction to plant personnel is a strategic pian (using a

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top-down approach) and an associated picture that indicates where the plant is heade Both of these items are currently in draft form. The licensee explained that it expects to issue its strategic plan after recovery efforts are completed fer Units 2 and 3. The licensee ,

has issued a draft "Long-Term improvement Plan" (which uses a bottom-up approach), l along with vision, mission, and strategic focus area statements, as part of its periodic  !

report to the NRC titled " Progress Toward Restart Readiness and Long-Term improvement l at Millstone Station - Northeast Utilities Briefing for the U.S. Nuclear Regulatory )

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Commission," dated February 11,1998. The licensee has also issued " Nuclear Group Policies and Standards." The " Policies and Standards" document lacks a " Nuclear Group

. Mission and Vision" statement because it is still being developed. That the vision statement and strategic plan were still draft oocuments was considered a weakness in light of the fact that most of the current management team has been in place since late 199 c. Conclusions The team concluded that the plant staff's clear understanding of management's expectations was considered a strength. The team concluded that because the strategic plan was in draft form, that this was a weakness in managements's long-term direction of plant personne .2 Organizational Communications and Teamwork a. Inspection Scope The team evaluated organizational communications and teamwork, including

. interdepartmental relationships and interfaces. The team assessed both vertical and -.

horizontal communications. It verified that communications are adequate to properly identify and characterize safety-significant issues. It also assessed whether communications between organizations were adequate to properly address safety issue b. Observations and Findings (1) Organizational Communications

(a) Interdepartmental Relationships and Interfaces As noted in paragraph 1.1 above, the team observed that daily meetings showed good interdepartmentalinteractions, a questioning attitude by participants, and command and control by senior managers. However, the team found that horizontal communications  !

were not as effective or as free flowing as vertical communications. The team Bund that I communications within groups were more effective than communications between group Interviews indicated inat communications between groups were greatly dictated by the necessity of completing a task and communications were more free flowing when individuals had some prior relationship with one another. Some impediments to horizontal communications included the following: time constraints, busy schedules, the fact that l communications are often issue driven, the need to resolve an issue, cultural issues, and respect issue The team noted ineffective communications occurred between certain elements of the maintenance and oversight organizations. . The team also found that senior plant

- management was aware of these conflicts and had taken effective, immediate actions to include long-term actions to resolve the conflicts.

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(b) Vertical Communications The team found that vertical communications were a strength of the plant organizatio Interviews indicated that vertical communications were especially strong with respect to using the corrective action system and sending constructive inputs up the chain of command. The team also found that upper management's ability to effectively i communicate its expectations down the chain of command was a management strengt l i

(2) Teamwork The team determined that team-building initiatives had been completed beginning with officers, then directors, followed by managers and first-line supervisors. The team saw evidence of teamwork in several regularly scheduled interdepartmental meetings. Interview results indicated that teamwork initiatives are currently on hold until after startup of Unit Interview results did not identify any teamwork initiatives at the level of the plant worke The interviews indicated that some groups are in conflict (most notably oversight and maintenance), not everyone is familiar with conflict resolution, sometimes interpersonal l conflicts are not resolved, and no formal process exists for rotational assignment c. Conclusions The team concluded that communications between groups and departments was effective l to address safety issues. The team observed that horizontal communications could be improved. The team concluded that teamwork initiatives at the level of the first-line supervisor and higher resulted in good organizational interactions were good initiatives and need continued reinforcement. The team also concluded that teamwork initiatives at the level of the plant worker has not been develope .3 Encouragement of Problem Identifu(Won by Managers and Supervisors a. Inspection Scope The team evaluated whether managers and supervisors encourage employees to identify problems and whether the staff believes that management is receptive to the problems being brought forwar b. Observations and Findings Observations, interviews and plant surveys show that managers and supervisors encourage employees to identify problems. Interviews also indicated that the plant staff believes management is receptive to problems being brought forwar c. Conclusions The team concluded that plant management was effective in its efforts to encourage plant personnel to identify problem !

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1.3.1 Management's Encouragement and Receptiveness in Problem identification by Individuals a. Inspection Scope The team evaluated the NU problem resolution processes (line management practices, corrective action program and others) to determine their effectiveness in providing a vehicle for encouraging identification of, readily accepting and adequately resolving problems and issues identified by individual employees as pathways preferable to the Employees Concern Program (ECP) or the NRC allegation process. The team also attempted to determine if managers and supervisors encourage employees to identify problems, and if the staff feels management is receptive to problems being brought forwar ,

b. Observations and Findings

The team conducted employee interviews at all levels of the organization to determine the i worker's perceptions of management's efforts and communications intended to enhance i problem identification. Performance measurement data for prcblem identification and documentation associated with individual-identified problems (Condition Reports, Employee Concern Case Files, Self Assessment results and others) were reviewe The employee interview results indicated that previously existing barriers to problem identification had been largely eliminated, and no major barriers to problem identification were found. Personnel from organizational units both with and without histories of such barriers to problem identification were interviewed, and generally characterized the environment as improved and currently receptive to problem identification. No reluctance or reservations were expressed by the individuals with regard to their identification of problems to line management, the Condition Reports (CR) process, the Employee Concerns Program (ECP) or the NRC. Most of the individuals interviewed indicated that they had initiated CR's either personally or through referral to their supervision. Several of the individuals stated that the current working environment also supported escalating concerns and problems to management above their direct supervision if believed by the employee to be necessary. The NU performance indicators and condition report program statistics reflect reasonable levels of employee participation, which corroborated the interview result The team also reviewed the Fmployee Concerns Program (ECP) Program, the Safety Conscious Work Environment (SCWE) Program, the Employee Concerns Oversight Panel (ECOP) and other management initiatives for NNECO activities and response actions taken for previously existing barriers to problem identification above. The licensee's handling of individual, alleged cases of harassment, intimidation, retaliation, or discrimination by the ECP and SCWE organizations appears adequately responsive to specific case needs. Both technical and human behavior and performance problems were generally well addressed in the case files reviewe f

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However, a recent NRC inspection 50-423/97-212of these programs in accordance with NRC IP-40001 found that the SCWE processes were not formalized. The lack of SCWE

. program structure (procedures, formal processes and documentation requirements) resulted in many of the management actions being handled directly, on an ad hoc basis, by - .

Recovery Officers and senior managers. The NRC staff found during this inspection that-the program was still not formalize The licensee uses incoming ECP allegations, employee culture surveys, leadership surveys and ECOP activities to identify barriers to problem identification and harassment, intimidation, retaliation or discrimination (HIRD). ECOP oversight activities and surveys are -

used to identify potential or actual HlRD problems or organizational units which exhibit barriers to free identification and reporting of problems, and appears to be a positive contribution to the overall proces While the licensee has become effective in identifying these barriers, they have not been fully effective in addressing a continuing high incidence rate of HIRD allegations coming to tthe Employee Concerns Program. The ECP, ECOP and SCWE mechanisms are effective,

, especially for the more significant issues identified as problem areas. However as

& discussed in report section 1.3.2, NNECO management has not been fully effective in

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a dealing with trends and common causes'for HlRD allegations generated organization-wide ~

'to ECP.- A significant backlog of HIRD allegations was pending investigation and the backlog and emergent HIRD allegations had not been analyzed by NNECO for broad trends or patterns, and common causes. As a result,' the actions taken to date had not been -

effective in assessing the nature and substance of the continuing high incidence rate of HlRD allegations. The licensee's effectiveness in acting to turn the trend of HlRD and prevent its recurrence is further discussed below, c. Conclusions Observations and interviews show that managers and supervisors encourage employees to identify problems. The team determined.that the plant staff feels that management is receptive to problems brought forward, and individuals generally characterized the environment as improved and currently receptive to problem identification. The team observed that there is no reluctance or reservation expressed by individuals to identify problems. The monthly HIRD allegation incidence rate and the frequency of actual and potential chilling effect events at Millstone has not significantly diminished during 1997-

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1998 and continues to represent challenges to the safety conscious work environmen '1.3.2 ' Effectiveness of Problem Resolution Processes in Resolving Employee Identified Concerns -

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- ai Inspection Scope-Assess the effectiveness of the process by which NNECO management provides problem resolution to employee identified concern _

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' Observations and Findings The. team conducted' interviews, reviewed corrective action program data, and reviewed the licensee's technical resolutions, SCWE-related responses, and long term follow-up for

. problems identified by condition reports, employee concerns and Employee Concerns Oversight Panelissues, and NNECO department self assessments. This included a sample of condition reports and their technical and human performance / behavior resolution activities that resulted from employee concern cases and self assessment The team also reviewed ECP cases, SCWE Focus Area response plans, and a sample of CR corrective actions to determined if licensee actions were generally responsive to problems

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'(1) SCWE Focus Areas and Potential Focus Areas

, identification methods for HIRD issues include the NNECO Leadership Surveys, Culture Surveys, ECP case intakes, and ECOP identifications and referrals. These mechanisms appear to be effective, especially for the more egregious issues identified by NNECO

- , management as HlRD," Focus Areas". Documented action plans are used to define and -

- manage the remedial actions for the Focus Areas. - A sampling review of these plans by the team determined that action plans are generally effective at remediating both the technical lasues and human performance and behavior issue I Other HIRD issues that are perceived by NNECO management to be less egregious or that are perceived as minor problems having the potential to eventually become" focus areas" are handled more informally by senior and middle management. These issues are identified by line management via normalline management oversight and activities, review of ECP intakes, and SCWE daily meetings which include dialogue with ECP, ECOP, Human Resources, legal, the ECP/SCWE Independent Third Party Oversight Contractor (Little Harbor Consultants, Inc.) and other program participants. No documented action plans are used, but the team found that management had taken actions in direct response to such HlRD potentialities and were monitoring the effectiveness of the action Notwithstanding the specific HIRD and HIRD-potential management response actions taken to date for HIRD-related problems an average of about 54% of all ECP allegations included HIRD and about 26% of the total allegations involved HlRD associated with nuclear safety related protected activities subject to 10 CFR Part 50.7. The fraction of the total cases not

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involving 10 CFR Part 50.7 typically involve other forms of HIRD, e.g. age, sex or other discrimination, harassment or intimidation over work rules, compensation, benefits, et The overall frequency of personnel behaviors that can be construed as HIRD, regardless of l their source or subject has been identified by NNECO and NRC as a serious concern relative to the establishment of a healthy SCWE. While NNECO notes that only three

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1997-98 ECP cases have involved confirmed 10 CFR Part 50.7 HIRD issues, investigation of many of the alleged cases remained open at the time of the inspection.

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. .NNECO has performed extensive re-evaluation of the data to ensure their categorization is j correct; re-evaluation resulted in no substantive changes in categorization. NNECO

acknowledges that HIRD in non-50.7 environment represents an environment in which

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50.7 HIRD could also emerge. The team discussed and evaluated the management actions planned and taken with the Unit 3 Vice President, the Vice President-Engineering :

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. responsible for SCWE, the SWCE management team, the ECP assistant director, and other (2) Measuring and Test Equipment Program issues Several of the ECP cases reviewed by the team identified programmatic and

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implementation problems in the NNECO programs for control of measuring and test equipment (M&TE). .The inspector initially reviewed the two M&TE related CR's (CR-M3-

' 97-1292,5/2/97 and ACR M3-97-0150,1/15/97) issued in response to the ECP cases, finding that they also related a several other related CR's. Further inspector review of these and a several year history of M&TE-related CR data found numerous NNECO-

_ identified examples of M&TE program deficiencies . Discussions with NNECO and NR _ Region I personnel indicated that the adverse performance history existed at least from 1992. The issues concerned mis-handled M&TE: for example', failure to perform impact

- reviews for M&TE found out of calibration.

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NNECO was requested to provide further information on actions taken to address apparent-adverse trends and a meeting was held on February 18,1998 with the newly appointed -

Metrology Lab Supervisor, the new program owner. The supervisor advised that all M&TE issues had been rolled up into ACR M-1-96-0614 which has been used as a vehicle to address all contemporary problems and drive development of a completely new M&TE progra On February 25,1998, following the end of the inspection, NNECO provided the inspector

- with ACR M-1-96-0614 and its associated corrective action documentation which-indicated that the M&TE Program had been comp;etely re-written and had just become effective in early January 1998. The new program provided for a new Metrology Laboratory to provide central control for equipment and activities, a new training for all program implementors / users. Full implementation had not yet been achieved but was expected to complete soon with an Effectiveness Evaluation planned for mid-199 Based on the long history and broad extent of problems with the M&TE program and the comprehensiveness of the changes in the new progrsm, this item is unresolved pending completion of the licensee's specific corrective actions and fullimplementation of the -

program - (URI 423/97-82-01). _

c. Conclusions-The handling of individual HIRD cases by the Employee Concerns Program and the Safety Conscious Work Environment (SCWE) program is adequately responsive to specific case

. needs.' Both technical and human'-side problems are generally well addressed. The ECP

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case intakes and the Employee Concerns Oversight Panel (ECOP) oversight activities and '

- surveys are effective in identifying potential or actual HIRD problems or organizational units which exhibit barriers to free identification and reporting of problems. These are positive contributions to the overall process. These mechanisms are effective, especially for the more significant issues identified as focus areas. However, NNECO management has not

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been fully effective in assessing trends and common causes for HIRD allegations generated organization-wide to EC .4 Performance Monitoring a. ' Inspection Scope The team evaluated performance monitoring (performance indicators), including management information systems employed to evaluate the following programs: corrective action, root cause analysis, self-assessment, independent oversight, and operating experience. They evaluated the effectiveness of the performance measures process ,and assessed the quality of the information on performance that is given to management. The team verified that the licensee takes action when any of the performance indicators identify areas that warrant management interventio 'b. Observations and Findings

.The various performance indicators used to evaluate the programs of interest were 4  : considered excellent. The licensee has appropriately addressed adverse trends identified i the fourth quarter report concerning compliance of the maintenance department with <

procedures, valve and breaker alignment issues and tagging errors of the operations department, and surveillance testing. The team's analysis of. Licensee Event Reports .

(LERs) found that Unit 3 had 16 human performance related LERs in 1997 versus a national average of six human performance related LER c. Conclusions

- The team concluded that the licensee's performance monitoring program was good. The team also concluded that the high number of LER-related human errors was a potential

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adverse trend that the licensee should examine furthe .5 - Management's Commitment To Resolve issues a. Inspection Scope The team evaluated management's commitment to resolve safety committee recommendations, audit findings, assessment recommendations, and open issue b. Observations and Findings (1) . Resolving Safety Committee Recommendations Plant Operating Review Committee (PORC) Report Number 3 97-256 contained six recommendations, including preparation of a temporary modification and safety evaluation

" report pertaining to the repair of a valve. The staff determined that management had

. addressed all of the subject recommendation I

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(2) ~ Resolving Audit Findings Audit Report (AR) MP-97-A06-02 noted one deficiency regarding reporting by the Unit 3 Radiation Protection Manager (RPM) to the Maintenance Manager. Unit 3 Final Safety Analysis Report (FSAR) Section 12.5.3 states that all health physics procedures and methods for ensuring that occupational radiation exposure is as low as reasonably achievable (ALARA) follow the provisions and suggestions of Regulatory Guide (RG) 8.8, Revision 3; RG 8.10, Revision 1-R; and RG 1.33, Revision 2, as applicable. RG 8.8, Section C.1.b(3), states, in part, as follows:

"The Radiation Protection Manager (RPM) (onsite) has a safety function and responsibility to both employees and management that can best be fulfilled if the individual is independent of station divisions, such as operations, maintenance, or technical support, whose prime responsibility is continuity or improvement of station operability."

Millstone Unit 3 Technical Specification Section 6.2.1.d states, in part, as follows:

"The individuals who train the operating staff and those that carry out health -

physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures."

The licensee closed CR M3-97-1875, noting that the Unit 3 RPM has a direct reporting espability to the Unit Director on radiological issues as noted, by asterisk, on the organization chart. Having the RPM report to the Maintenance Manager is inconsistent with the Unit 3 technical specifications. The licensee stated during the onsite inspection that this issue would be evaluated along with other organizational changes that are currently being considered. Therefore, the team considers this item to be a violation of Technical Specification 6.2.1.d (VIO 423/97-82-02).

(3) Resolving Self-Assessment Recommendations Self-assessment Number 3TS-SA-97-02 involving Unit 3 reactor engineering procedures made five recommendations. The team found that the licensee had addressed each of these recommendation (4) Resolving the Open issue on Independent Verification That Human Errors Have Been

' Reduced

- Sectic,n 6.2.3.4 of the Unit 3 technical specifications makes the following statement:

"The ISEG [ Independent Safety Evaluation Group] shall be responsible for maintaining surveillance of unit activities to provide independent verification that these activities are performed correctly and that human errors are reduced as much as practical."

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t The licenroe's memorandum dated December 23,1997, made the following statement:

"In the event that the HF/E [ human factors / engineering) review effort needs to be audited or become a part of a " Nuclear Oversight" assessment, an independent -

party will be assigned to ensure that independence is maintained. HF/E personnel-may participate in Nuclear Safety Engineering ISEGs and in Nuclear Oversight audits and assessments that do not include previous HF/E involvement in design changes,

. MCB [ main control board] changes, procedure reviews, etc."

The licensee's position on this issue is 'i inconsistent with Technical Specification 6.2. However, there were no examples identified of this position being exercised such that'.

_

there was a lack of organizational independence required by the Technical Specifications -

(TS). Licensee actions on this issue will be reviewed in the future (IFl 423/97-82 03).

(5) Resolving Open issue on Establishing or Applying Appropriate HF/E Guidance for Advanced Computer-Based Displays in the Control Room The team determined that Specification SP-EE-149A details the design requirements for the Safety Parameter Display System (SPDS). The team found that minor modifications to the -

SPDS have been made over the years. However, the guidance references in the specification are out of date in that they do not contain current guidance on computer-based displays. The licensee stated that AR Number 97027958 titled " Review of and Update of Plant Process Computer Specifications" would update the guidance references by December 1,199 (6) Resolving Open issue Regarding the Post Loss-of-Coolant Accident Cooling Status Tree The team determined that the post-LOCA cooling (PLC) status tree is not addressed in the Emergency Operating Procedure (EOP) User's Guide, Section 1.6, " Monitoring Status Trees," or Attachment 4, " Control Room Usage of Status Trees." The licensee ctated that this issue would be addressed by AR Number 07031064, titled "PORC Commitment to

. Investigate the Overall Feasibility of the Post-LOCA Processing" by December 1,1998.The team determined that the navigation scheme for the PLC tree was not the same as for other trees and did not agree with the SPDS specification. The licensee provided i documentation to show that the SPDS specification (i.e., Section 6.6.6 of SP-EE149A, l Revision 6) had been changed to match the navigation scheme for the PLC tree. The team

'

found this change to the specification acceptable for closing this open issu (7) Resolving Engineering Review Recommendations

.The team evaluated the effectiveness of the licensee's management team to resolve

- recommendations,made by Engineer;ng Review 3-ESAR-97-OO8.

E

' One recommendation concerned revising NGP [ Nuclear Group Plan] 5.25 and SP-EE-26 : The licensee documented that the revision to SP-EE-261 will be completed by March 30, p 1998,through AR 96000103,and willinclude a section of standards for computer

. displays and an update pertaining to Revision 1_of NUREG-0700. The licensee _ also K

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11 documented that the revision to NGP 5.25 will be completed by. June 30,1998, which includes a change to NGP 5.25 that would require the Human Factors Specialist rather than

- the Project Engineer to determine whether a detailed control panel design review is

- warranted and an update pertaining to Revision 1 of NUREG-070 The engineering review also raised the question of whether there was a match between the simulator and the control room designs. The licensee's documentation noted the following strengths of the simulator update process:

"

At this time, there are no design changes with simulator impact that have been installed in the plant for [merd than 30 days that have not been incorporated in the j simulator." '

"

L To support the current restart training needs, we have modified our target for.

I incorporating those plant design changes identified by Operations and Operator Training as having simulator impact to have them installed within 30 days of plant installation. This is beyond what is required by ANS-3.5, the standard to which the Millstone 3 simulator is certified . The standard allows 24 months to incorporate such changes."

The licensee's self-assessment (97-004) identified the following weakness of the simulator update process:

"This assessment has demonstrated the absolute need for coordination between site and the simulator support group to ensure that all plant modifications are accounted for in the simulator upkeep."

Overall, the team found that the simulator process, which ensures that the fidelity of the simulator is maintained with regard to the reference plant, was a strengt The team found that there have been several advanced systems added to the control room without consistent design guidelines (e.g., Foxboro intelligent automated (lA) system for the moisture separator reheaters (MSRs), fire protection, environmental qualification (EO)

temperature monitoring, the auto-log system, and SPDS upgrades). There are currently eight different computer-based systems in the main control room. Since many of these systems are significantly different in their human systems interface, displays, and alarms, it was the team's observation that the licensee should evaluate these different systems with respect to unnecessary operator burden and the potential for increasing operator erro Further, it was the team's observation that such an evaluation should consider the criteria in the licensee's revision to SP-EE-26 I

c. Conclusions ' )

J J Overall, the team concluded that management's commitment to resolving safety committee

- recommendations, audit findings, assessment findings, and open issues was satisfactor , . However, the licensee's f ailure to resolve the two open technical specification related issues concerning the independence of the RPM and the independence of the ISEG relative

to reducing human errors was considered a violation and a weakness, respectivel _

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12 2.0 CORRECTIVE ACTIONS 2.1 Corrective Action Program a. Inspection Scope The team assessed the adequacy of the corrective action program. .This assessment included the' evaluation of programs for the identification, analysis, and resolution of plant'

deficiencies, b. Observations and Findings The team conducted discussions with the licensee and performed document reviews of the corrective actions program contained in the recently issued program procedure (September 30,1997) Revision 5 to RP 4, " Corrective Action Program." Prior to_1995, the program,

,

. employing what was known as the Plant incident Report, captured an average of only 300

' items per year for Unit 3 at a high threshold level for events or reportable conditions. This program was superseded by the site-wide Adverse Condition Report which improved the

" capture" threshold to approximately 4,000 items per year. In 1996 and 1997,'the

<

. licensee's program was moved closer to industry practices with Revisions 2,4, and 5 (Revision 3 was never issued). These revisions, among other changes, resulted in a multi-disciplined management review, as well as the requirement that the shift manager review ,

discovered conditions for operability and deportability. The revised condition reporting program also strengthened accountability, provided for enhancement items as well as adverse conditions, and defined management expectation .The team's conclusions regarding the adequacy of the licensee's corrective action program were that the program elements concerning identification, CR initiation, and CR processing were functioning at a good performance level. The CR program elements concerning root cause, corrective actions, and failure recurrence were considered to be operating at an

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acceptable level, but with room for improvement. The remaining element, effectiveness, involves trending and self-assessment, as well as performing effectiveness reviews. The first two attributes (trending and self assessment) show indication of being performed at an acceptable level. However, the effectiveness reviews, by their nature of being performed after the program is functioning for an extended period of time , cannot yet be fully evaluated due to the newness of the current program requirements.

l .The team evaluated the performance of the licensee's program for identification of adverse L conditions. As indicated by the range of CR's reviewed, and discussions with a broad l spectrum'of plant personnel, the team agreed that the licensee has attained a low l threshold for initiating Condition Reports. The fourth quarter continued the high volume trend of CR's initiated for Unit 3, with a total of 1,621 CR's written during the fourth quarter. An increased awareness by plant' staff and a decreased threshold for initiating a -

CRi appear to_be the causal factors for the high volume of CR's initiated. An additional reason for the high volume of CR's is the generation of numerous findings from the Independent Corrective' Action Verification Progra ,

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13 The team observed that the licensee's analysis of reported conditions (CR's),'while

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generally adequate, were at times narrowly focused. Several examples to support this L . ,; observation are discussed below in Section 2.2. ' Among those examples is the resolution'

L ' of closure probl9ms with High Energy Line Break (HELB) door The team also found that the Corrective Actions Program lacked controls over combining similar Condition Reports such that their issues were preserved and that they maintained their appropriate significance level. This weakness was evident in the handling of multiple condition reports concerning deficiencies in the Nonconformance Report process. This control issue was documented in great detail as CR M3-98-0309. In this case, a series of CR's documenting implementation issues within the Non-Conformance Report (NCR)

process were combined into a single Significance Level 1 CR, M3-97-3710. No specific h root cause was performed for the CR, but the analysis was transferred to another Level 1 CR, M3-97-0845,which did not concern NCR implementation issues. Additionally,'CR M3-97-3710 corrective actions were closed to a Level 3 CR, M3-974468,which was written to track recommendations from a self assessmen The team found that the controlling administrative document, RP 4, Revision 5, does not

, mestablish controls over combining CR's to preserve their subject material and Significance a Level. We consider this as a program weakness (IFl 423/97-82-04).

During discussions with the licensee it was learned that the Independent Safety Engineering Group (ISEG) had not been involved in any reviews of the CR program, its status, or the impact of the backlog of high-priority CR's. This determination further

- reinforces the team view that ISEG may need to rebalance its workload and reemphasize a :

focus on plant and management control issue c. Conclusions Overall,'the team has seen evidence that the corrective actions program is functioning adequately, but it is clear that the program will continue to require careful monitoring by NNECO management to ensure sustained performance. For example, the team found a notable number of minor process deficiencies, with the exception of the boric acid pump air binding issue discussed in Section 2.3.1,in a relatively small sample size of Condition Reports, after NNECO had completed their own extensive self-assessment preparing for this team inspectio The licensee has attained a low threshold for initiating Condition Reports. An increased awareness by licensee personnel and a decreased threshold for initiating a CR have resulted in a high volume of CR's initiate ~ The team also found that your Corrective Actions Program lacked controls over combining-similar Condition Reports such that their issues were preserved and that they maintained-their appropriate significance level. This a program weaknes e ,

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2.2 Corrective Actions - Classification and Root Cause Analysis al inspection Scope

The team reviewed a sample of root cause analysis and equipment failure evaluations to determine the adequacy of the process to classify and analyze condition report issues.' For less significant issues, the team reviewed a sampling of the apparent cause determination They also independently verified condition reports for significance, and that apparent' root cause determinations had been performed where require b. Observations and Findings For the Level 2 and' Level 3 CR's reviewed, the team generally agreed with the apparent -

causes. - Although adequate, the inspectors review of Level 1 CR's for root cause analysis found them to be narrowly focused and identified several cases where the root cause analysis was waived. Examples of waived or narrowly focus analysis are provided below:

• In the case of CR M3-97-0652,which described how design interface distribution and transmittal control of design information did not meet the requirements of Criterion lil and V of Appendix B, the root cause analysis was waived by the MRT with no accompanying explanation. The corrective actions involved training Unit 3 4 design engineers on DCR/MMOD requirements and emphasized the need for attention to detail. As determined by the team, the corrective actions appeared to be narrowly confined to the items associated with the DCR deficiencie * For ACR M3-97 0558, dated February 20,1997, the licensee's design basis verification program document for the Chemical and Volume Control System had non-conservative assumptions related to maximum temperatures for the letdown heat exchanger and charging flow. The root cause was waived and referred to other CR's with similar situations. The inspectors regard this waiver as a missed opportunity to thoroughly evaluate the issue. However, the corrective actions appear to adequately address the design deficienc *- Another instance involved ACR M3-97-0409,a Level B, ACR, dated February 4, 1997, that documented concerns for sump water level calculated head losse Although the cause of the event was addressed in an LER (LER 97-015), no root cause was performed for the CR. Nevertheless, the proposed corrective actions appear adequate to resolve identified design deficiencies and a modification review is scheduled following reanalysis with actions to be completed prior to mode change.

l

• The root cause analysis for a diesel low lubricating oil pressure trip addressed in

[-  : ACR 10428, dated June 6,1996, the analysis failed to develop valve manipulation i . ' scenarios that would have caused an engine trip and would have resulted in a 5B

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strainer being one-half full of oi *-.

'The root cause for MOV calculation errors addressed in CR M3-96-0833, dated c February _11,1997, was thorough. However the CR closure, dated December 31, G

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1997, waived the effectiveness review to the MOV program periodic self-l assessments. The team observed that this approach does not ensure that a l . corrective action effectiveness review will be completed during the next self-assessment, for example there was no specific review plan developed which considered the weaknesses and other factors identified in the root cause analysi ~

• . The root cause analysis for loss of the component cooling water system (CCP) ,

addressed in CR MP3 96-0919, dated December 15,1997, failed to analyze and -

address corrective actions for failed barriers concerning informalinstallation testing by the l&C group without a procedure, lack of post maintenance / modification

. testing, and a procedural non-compliance of failure to have vendor participatio Loss of all three of these barriers was evident from the information in the root cause analysis, which failed to develop corrective action '* :The material within seven CR's, concerning multiple electrical separation issues, was gathered under one root cause analysis, CR M3-96-1337, dated January 2, 1997. The documentation of that analysis failed to demonstrate thorough treatment of the issues encompassed by the analysis. For example, the root cause analysis for item number 3 of ACR M3-96-1287 stated that the electrical designer responsible for a design change examined one of the deficient conditions and "he stated that it looks like it was missed." However, there were no corrective actions to establish standards for design engineering field inspection The team noted that most of the root cause analysis associated with the CR's selected for review were well done. Specifically, the team found that the root cause analyses were good for: CR M3-96-0833, dated February 11,1997, CR M3-97-0066, and M3-97-0132, dated January 31,1997, M3-97-0119 and M3-97-0161, dated February 14,1997, M3-97-0709, dated April 1,1997, M3-97-0908, dated April 21,1997, M3 97-1821, dated July 8,1997 and M3-98-0200, dated February 9,1998. In the overall sample of approximately 100 CR's selected by the team, there was a mix in performance although there was a noticeable improvement in the quality of the analyses done in the latter part of 199 i

.The team also noted several instances where CR's were classified at a lower Significance Level than was appropriate. In addition to the issue addressed in report section 2.3.3, they also found that CR M2-97-0510, dated April 1,1997, documented an ISEG activity that identified significant work control issues in the high voltage switchyard. These issues involved personnel safety and loss of off-site power. ISEG promptly addressed the issue by stopping work in the switchyard. However, the CR was not initiated until two weeks late.. Once initiated, the CR was downgraded to a level three. The justification for the level three assignment was that the facts associated with the issue had been reviewed and actions taken prior to the issue of the CR. Further, this previous review determined that no procedural violations actually occurred. The evaluation done in response to the CR minimized the' significance of the issues. Based on the potential safety significance of the issues, and the fact that the ISEG, an oversight group, developed diverging opinions of !

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procedural requirements, the NRC concluded that the assignment of the level three evaluation was inappropriate and served to further minimize the issues. The level of

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review actually performed was well in excess of the review required for a level taree issue and supports the NRC conclusio c. Conclusions Though there were some cases where the licensee classified CR's at a lower level than appropriate, the licensee usually classified CR's properly. The licensee generally performed root cause analyses where applicable. For the Level 2 and Level 3 CR's reviewed,'the team agreed with the apparent root causes identified by the licensee.. Although adequate, the inspectors review of Level 1 CR's noted that the root cause analysis in several cases was narrowly focused, and other cases where the root cause analysis was waive . Although the quality of the root cause analyses had improved through 1997, overall quality -

was somewhat mixe .3 Corrective Actions Effectiveness a. . Inspection Scope The team evaluated the technical resolution of a sample of. safety significant issues for

' timeliness and effectiveness of corrective actions as well as the backlog of open condition'

reports to verify that safety significant items were being tracked to completion. The teams .

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evaluations also included interviews with supervisors regarding the closure rate and reviewing the process to prioritize corrective actions based on ris b. Observations and Findings During the inspection preparation trip in December 1997, the team questioned the apparently low closure rate associated with the licensee-defined, high-priority Level 1 CR' At that time, a review of the approximately 260 items classified as Level 1, and by the licensee's definition, a "Significant Condition Adverse to Quality," showed only 25% of the items to be closed. Subsequently, various supervisors, including the root cause supervisor i and several QA supervisors, were interviewed regarding the CR process in general and the apparent low closure rate in particular. From these discussions and a review of the current status of the Unit 3 CR backlog (as of February 16,1998), the following information was )

obtained:

Section 1.8.5 of the licensee's procedure, RP 4, Corrective Action Program, states that evaluation due dates of 30 days from determination of assignment duties are establishe The number of overdue evaluations in January 1997 was 825 items. This number had l - been reduced to as low as eight items by July of 1997. Since November of 1997 to the

!

present, the average backlog number has been about 20-30 items. This reflects a partial recovery of the management of this portion of the process. A very high volume of CR's )

have been and currently are being initiated; as the volume of CR's being generated I

. moderates, it is anticipated that this average backlog number will be further reduced.

..

Good efforts have also been made in the reduction of restart-related CR's. Starting in the )

. same period, January 1997, restart CR's were at 575 items. Due to the licensee's emphaels and resource focus applied to the reduction of the "over due evaluation" backlog, c

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o d'

.these CR's increased to more than sixteen hundred items during the perio' d when the evaluations were being driven down. Starting from July 1997, with the " freeing up" of resources from the efforts to reduce the "over-30-days evaluations," the licensee was able

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to drive down the restart CR's to less than 400 items currentl The licensee's procedure RP 4 requires CR's to be assigned to one of three significance levels; Significant Adverse Condition (Level 1), Adverse Condition (Level 2), and improvement item (Level 3). The licensee has in use a risk significance classification

. system that further subdivides risk into four categories. Specifically,' Attachment 4, Risk

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Significance, to RP 4 states that in addition to a determination of whether a condition adverse to quality is significant, further attribution is warranted to assist the CR Owner and Multi Discipline Management Review Team (MDMRT) in prioritizing the corrective action The four risk significant categories are defined as: Priority 1, Risk Critical- This category

- would result in consequences that are severe and unacceptable in either human, societal,

, political, or monetary terms; Priority 2, important to Safety - This category can result in -

risk to the reactor systems, industrial safety, public health and safety, or the environment; Priority 3, Compliance - This category corrects common and repetitive non-compliance wit laws, procedures, or normally accepted standards and expectations as defined by station standards and regulatory agencies; Priority 4, Good Management - This category identifies-an isolated condition adverse to quality or an opportunity for program enhancemen The team noted that the CR risk significance is not coupled to the PRA results or to risk ranking by the Maintenance Rule Progra Observations by the team of the MDMRT members review and evaluation of CR's showed that both the level categorization and the risk-significance priority classification were assigned to CR's generated each day. However, discussions, with the Corrective Actions Department and others, revealed that this last classification has not been acted upon beyond the assignment of the classification by the MDMRT. The licensee envisions further implementation of this concept in the futur Also, the inspectors did not discover any evidence, through review of RP 4 or interviews,

that the licensee was availing themselves of information from their IPE, nor from maintenance rule risk rankings, to assign CR classifications. Additionally, there was no evidence that risk-information was utilized for approving extensions to action due date The team selected a sample of CR's, and reviewed the resolution of these CR's to evaluate both timeliness and effectiveness. The inspector noted that the CR process generally

, appeared effective and serving its purpose. Several areas were noted for further follow-up

and are discussed below, i iMany of the'CR's had all of the Action Requests (ARs) completed but the overall CR was still open. The licensee stated that due to the high volume of CR's, as the

- CMP program is coming to a conclusion, there is presently a backlog of 600 to 700 CR's of this type in the system awaiting closure by the Corrective Actions Department, v.

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• In the computerized on-line CR system, all of the pertinent ARs do not appear as associated with their particular CR. The licensee stated that this is an historical .

problem that is in the process of being corrected but will still take more time to-backfit the appropriate link * There were noted to be some gaps in the tracking of Non-Conformance Reports (NCRs) and the ARs associated with NCRs. The licensee was aware of this problem and is in the process of addressing it (reference Memo MP3-CAD-98-004).

• There were several examples where the licensee scheduled some required corrective actions too far into the future, delayed or postponed corrective actions.

,

in addition to the detailed discussions of CR corrective action effectiveness, which follow .

l in the sections below the team observed:

• Corrective actions for unclosed HELB (High Energy.Line Break) doors were not i timely and were narrowly focused. During January 1997, six Level 1 CR's were written on problems involving HELB doors and the failure to close them properl Corrective actions focused on correcting personnel errors as the resolution to the recurring problem of unclosed doors. Subsequently, in August of 1997, CR M3-97-2567 was written. This CR addressed and described the physical deterioration of the doors and stated that problems ranged from missing gasket sections, damaged thresholds, and damaged gasket bars, to large gaps around the gasket seat The origination of this CR (M3-97-2567)was not driven by the HELB door closure problems but rather by a different review (CMP Pl 21, " Engineering Topical Areas i

- Reviews"). This later CR pointed out the lack of a PM program, the lack of regular inspections, and the lack of inspection criteri * CR M3-97-2898, dated September 2,1997, documented that a nuclear oversight audit identified procedures, tools, and equipment needed to support emergency operating procedures were not available in the plant. The corrective actions for this CR were not tied to a key event and were outside of the scheduled restart date at the time they were approved. This issue was also picked up in the NRC review of deferred items and was moved to a mode 2 issu * CR M3-97-3974, dated November 11,1997, documented an audit finding that the ISEG/OE procedure, NOOP 3.04, Nuclear Safety Engineering Group Functions and Responsibilities - ISEG and OE Assessment, was not reviewed by SORC as required i by technical specifications. Technical Specifications 6.2.3.1 and 6.5.2.6 require l that the Independent Safety Engineering Group (ISEG) procedures be reviewed by l

- the Site Operations Review Committee (SORC). Condition Report (CR) M3-97-3974 documented that this review had not been accomplished. The procedure for ISEG

. review of operating experience was reviewed by SORC.. However, other ISEG =

procedures were not. This is a violation of technical specifications (VIO 423/97-82- !

05).

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• - CR M2-98-0419 documented seven pieces of tape that were identified by the NRC during a QC foreign materialinspection of the unit two spent fuel pool. The CR was assigned to reactor engineering as a technical issue and was not evaluated by QC as an inspector performance issu c. Conclusions Efforts have been made to reduce the number of CR evaluations open over 30-days. This -

reflects a recovery of the management of this portion of the process. Good efforts have also been made in the reduction of restart-related CR's. Progress on closing Level 1 CR's

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has accelerated since late last year. However, some corrective actions were narrowly focuse The team concluded that though there were some incidences of weak implementation of corrective actions, the technical resolution of safety significant issues was adequate and that the licensee was adequately tracking open condition reports to closure. They noted the potential confusion that could be generated by the licensee's use of the term " risk'

'

significant" relative to its classification of CR's. The team also noted that the licensee did not apply current risk ranking methodologies (e.g., PRA, Maintenance Rule) to its CR-classification scheme. However, the prioritization of CR's was appropriate in most -

instances reviewed by the tea .3.1 Boric Acid Transfer Pump Air Binding a. Inspection Scope The team assessed the adequacy of the licensee's actions relating to CR's concerning air binding of the Boric Acid Transfer Pumps, which has been a recurrent issu b. Observations and Findings Boric Acid (BA) Transfer Pumps (3CHS-P2A & -P2B) are part of the Technical Specification required reactivity control systems and provide a boron injection flow path to the Reactor Coolant System. CR M3-97-2943, dated 9/4/97, identified a chronic air binding problem with the pumps that periodically rendered them inoperable. The CR noted that the condition had previously been identified in 1992 (PIR 3-92-210); 1995 (ACR 3617); and earlier in 1997 (CR M3 97-0715,CR M3-9701011,CR M3-97-0954)but not fully resolve NNECO's root cause analysis concluded that air entrainment caused the 1992 event,

- vented the pump and required no further cction as it was considered an isolated

. occurrence. No cause was determined for the 1995 ebent other than the presence of

' trapped air in a horizontal pipe run. Internals of the Boric Acid Storage Tank (BAST) batch l

<

atank discharge check valve were removed and 'a modification was proposed to add an

.- ~ isolation and manual vent valve to provide improved venting. CR M3-97-0954 identified .

the recurrence of the problem and initiated actions to revise BAST low tank level operating (: limits and alarm set points upward to prevent air entrainment; revise the system Operating l

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Procedure to vent pumps if BAST levels drop below the increased low limits; and to .

complete the valving modification originally proposed in 199 The corrective actions for M3-97-2943 were essentially the same as the earlier-1997 '

actions above except that they included evaluation of a possible modification to re-route the Boric Acid Gravity Flow Boration Piping. The licensee's root cause had concluded that the height of the BA pump and gravity feed piping connections to the BASTS relative to the minimum TS-allowable tank levels had resulted in air.entrainment and pump binding. Th proposed corrective actions had been approved by management and were awaiting full implementation at the time of the inspectio . The NNECO root cause and corrective actions required implementation of BAST level administrative controls more conservative than the minimum levels required by Technical Specification 3.1.2.5.a, " Borated Water Sources - Shutdown," or 3.1.2.6.a, " Borated Water Sources - Operating." The inspector requested the licensee's 10 CFR Part 50.59

, safety evaluation and 10 CFR Part 50.72 or Part 50.73 deportability evaluations for these

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actions in that they appeared to constitute an unreviewed safety question. Specifically, if the' pumps and tanks were operated in accordance with the Technical Specifications, they might be unable to perform their analyzed safety functions as described in the Final Safety Analysis Report.-

The licensee advised that the M3 97-2943 proposed administrative controls had not been completely prepared and published, consequently the 10 CFR Part 50.59 evaluation had

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not yet been completed and was therefore unavailable. Further, the NNECO Unit 3 Licensing Manager stated the licensee's belief was that the above conditions did not represent an unreviewed safety question, and that a Technical Specification Change Request was planned for later submittal to the NRC. Operation would proceed in accordance with the administrative tank levellimits pending NRC approval of the chang ' The lic'ensee subsequently determined that this issue did not involve an unreviewed safety I question. However, the failure to correct the air binding of the boric acid transfer pumps and take actions to prevent recurrence of a significant deficiency is a violation (VIO 423/97-82-06).

Additionally, the inspector reviewed potential air or gas infiltration paths with the NNECO system engineer. NRC Information Notice 88-23,INPO " Red" Significant Operating

- Experience Report 97-001, and recent events at a Unit 3 sister-plant, Beaver Valley Power ( Station, had each identified the potential for accumulation of substantial amounts of j hydrogen gas in charging pump suction lines. The accumulation resulted from the large J pressure drop across the charging pump mini-flow orifices stripping dissolved hydrogen '

-from the charged coolant and returning it to the pump suctions. The phenomenon has the

- potential for severe charging pump damage. Because of the location of the BA pump

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piping connection to the charging pump suction piping, the inspector had inquired whether the licensee had considered hydrogen accumulation and migration into the affected BA

- pipin The NNECO system engineer, responsible for both the boric acid and charging systems, was unaware of the operating experience information referenced above and the information

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i had'apparently not been considered in the root cause determinations for the CR's. :it was later determined that NNECO Design Engineering had been assigned responsibility for the operating experience items'and had preliminarily determined that Unit 3 did not have the hydrogen problem. The basis for the determinations was not reviewed; the matter was referred for further follow-up to the NRC Operational Safety Team that was concurrently onsit ' On February 19,,1998, NNECO issued CR M3-9'8-0975 which documented inadequacies in the root cause evaluation for CR M3-98-0954in that it did not consider potential

- introduction of air from the BAST batching process.

l The licensee further advised that the deportability evaluation performed with the.CR's had l evaluated the deportability for plant conditions at the times of discovery of the pump inoperabilities but had not adequately evaluated past plant conditions for potential, historical inoperabilities of the BAST /BA pump boration flow paths. On February.18,1998,

the licensee issued CR M3-98-0952 which documented the inadequate historical deportability evaluations; disposition of the CR was in progress at the end of the inspectio '

c. Conclusions J

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The team found that there had been inadequate root cause determinations and corrective actions for recurrent BA pump problems, including two cases under the contemporary CR system. The licensee's deportability evaluations did not evaluate the availability of boric acid flow paths required by the TS for the past events. NNECO did not adequately consider operating experience at other reactor facilities in their evaluation of the proble The team also determined that NNECO did not recognize the potential USQ concerning BA tank level changes and the TS. The failure to correct the air binding of the boric acid transfer pumps and take action to preclude recurrence is a violatio .3.2 480V Molded Case Circuit Breaker Magnetic Trip Setpoints a. Inspection Scope

.The team assessed the adequacy of the licensee's actions relating to CR's concerning the

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overload magnetic trip setting for 480 Volt molded case circuit breakers, which has been a recurrent issu b. Observations and Findings The team reviewed NNECO Engineering Self Assessment 3DE-SA-97-03," Control of

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Magnetic Trip Settings on 480 VAC Motor Starters" which resulted in CR M3-97-3095 documenting that incorrect magnetic trip settings and thermal overloads were found in safety related applications. The self assessment was comprehensive and rigorous and

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represented a quality effort to evaluate the technical issue .

'The self assessment found three of eighty four setpoints to be incorrect. Proposed H corrective actions were generally comprehensive, but required verification of motor name e plates to determine actual required trip settings._ Name plate inspections were deemed

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impractical for a "large" (but undefined) number of breakers due to their inaccessibility. As

. an alternate to the inspections, NNECO determined that the erroneous setpoints occurred when motors were replaced, resulting in new operating currents, but the trip settings were-not appropriately adjusted. Consequently, NNECO reviewed motor maintenance files and

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confirmed that none of the unverified motors had been changed. This caused them to conclude that the setpoints are probably correct and further confirmation is not needed prior to restar Long term planned corrective actions included: revision of the 480V MCB Preventive Maintenance procedure to. include verifications by Unit 3 Refueling Outage 6 (scheduled approximately 10 months after Unit 3 restart) and eventual implementation of the PM procedure, as revised, to perform the inspections and calculations necessary confirm the conclusions of acceptability. The CR documentation provided initially.did.not identify the

. equipment affected by the incorrect settings, did not include the values for the incorrect

. as-found magnetic and thermal overload settings, the amount of deviation from as-required

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' settings, nor the evaluation of the impact of the deviations on the operability of the

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affected motors. ' Further, the documentation did not identify the motors which had not

' been inspected'and verified nor the licensee's basis for the acceptability of delaying inspection and setpoint verificatio As a result of the teams review of the licensee's efforts to assess this issue, the following information was subsequently provided by the licensee between February 25 and March

.18,1998 following the end of the inspection. Only the original three mis-set breakers were found. The reported deviations between the as-required and as-found settings were, in each case, minimal and in the low (conservative) direction to protect the equipment from over current. No other discrepancies were identified in the 121 motor / breaker sets inspected to date. The remaining unverified thirty three breaker loads / settings included four ernergency diesel (EDG) fuel transfer pump motors; various control building, EDG building, and intake structure HVAC motors; and other similar load The licensee concluded that, since the incorrect settings appeared to be the result of improperly documented motor replacements and that no other motor replacements had

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taken place for the thirty three unverified breakers, the probability of incorrect settings is negligible. NNECO further advised that they had examined their records for possible nuisance equipment trips for the affected loads that might have resulted from possibly low trip settings and had found no indication of problems. Most of the equipment was either in continuous or frequent intermittent operation and was functioning properly Upon further review, NNECO advised the team that the plans still include motor name plate verification during the next scheduled PM. - However, in those cases where the next PM is more than 18 months from restart, the inspections will be scheduled to occur during an earlier system outag ;

c . Conclusion

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i The team concluded that the actions proposed by NNECO for this issue were acceptabl ,

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2.3.3 Actions on Condition Reports - Failure to Conform to NRC Commitments a. inspection Scope The team assessed the adequacy of the licensee's actions relating to CR's concerning the previously identified failure to conform to NRC commitments, b. Observations and Findings The team noted four CR's that involved failure to conform to NRC commitments made in Generic Letters (GLs) and that were apparently misclassified as a Level 3 versus a Level 2 CR and/or contained ARs that were deferred post-startup. RP 4, Rev. 5, Attachment 3, CR Initiation and Classification Guidelines, includes the following in the Level 2 guidelines: an external station commitment not adhered to; or a deficiency in material that, if left uncorrected, could affect safe reliable plant operation. The specifics regarding these CR's were as follows:

o CR M1-97-19141s a level 1 CR (related to GL 90-03 and applicable to all three Units). Corrective action #1 (AR 97020123-02)for this CR was inappropriately classified as deferrable to post-startup; however, the portions of this AR that related to compliance with the commitment were previously completed in December,199 * CR M3-97-4672is related to GL 89-13 and was inappropriately classified as Leyel 3. Further, none of the actions were coded as needing to be completed prior to startup. At the time of the inspection allitems but one were noted as complet The one remaining item appears that it should be completed prior to startu * CR M3-97-4346 contains a deficiency in material (inadequate corrosion control)

that, if left uncorrected, could affect safe, reliable plant operation. It is related to GL 89-13 and was inappropriately classified as Level 3. Some of the corrective actions appear that they should be completed before startup. CR M3-97-3501 is also related to GL 89-13 and documents an adverse condition. It is classified as a Level 2, however the corrective actions were not tied to startup. Some of the corrective actions appear that they should be completed before startu CR's M3-97-4672 and M3-97-4346 were inappropriately classified as Significance Level 3 contrary to the requirements of procedure RP 4, Revision 5, Section 1.4 and Attachmen This is a violation of 10 CFR 50, Appendix B, Criterion V (VIO 423/97-82-07).

l The licensee issued CR M3-98-0933 to address these findings and performed a preliminary review to determine the cause. They noted several breakdowns in their program related to l- corrective actions for NRC commitments and began actions to both correct current

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instances and prevent further such occurrences. On March 9,1998, the licensee reclassified CR M3-97-4346 and CR M3-97-4672 as level 2 CR's.

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c. . Conclusion -

The team concluded that NNECO had failed to appropriately classify the CR's which were written for failed NRC commitments, which is a violation of the procedural requirements of RP .3.4 Fire Protection Program Compensatory Measures a. inspection Scope While inspecting the area of Performance Monitoring, Section 1.4 of this report, the inspector identified an issue with NNECO managements acceptance of compensatory measures in lieu of performing surveillanc b. Observations and Findings The team determined that the licensee intentionally does not perform five fire protection surveillance because the operations department does not have the manpower to conduct

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the surveillance. The following condition reports (CR's) have been written in connection with the surveillance: M3-97-3035, M3-97-3981, M3-9'7-4246, M3-97-4394, and M3-97-4618. The licensee has implemented hourly patrols as a compensatory measure. The licensee stated that the subject fire protection surveillance would be performed by Station Fire Brigade personnel when they are traine c. Conclusions The team concluded that the licensee's failure to conduct five fire protection surveillance and exclusive reliance on the 1-hour roving fire watch was an inappropriate deviation from the committed fire protection program. The team also found that relying exclusively on the 1-hour roving fire watch as an interim compensatory measure without other compensatory measures was a weakness, as noted in Information Notice 97-48, " inadequate or inappropriate Interim Fire Protection Compensatory Measures."

3.0 SELF-ASSESSMENTS 3.1 Self-Assessment Program a. inspection Scope The team evaluated the Millstone site and departmental self-assessment programs to determine their. focus on safety. They also evaluated the effectiveness of management to address the findings of self-assessments and performance improvement programs. The

. team verified that significant issues that could impact plant safety are being been addressed.

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b.'. Observations and Findings

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The stated purpose of the licensee's self-assessment program is to identify areas of concern and to improve performance utilizing the established ' corrective action progra Specifically, the programmatic controls described in unit procedure Unit 3 OA 11,?Self-Assessment", Revision 1, directs the performance of pre-planned, department self-assessments by qualified individuals with the objective of achieving higher standards of-

- quality and performance at Millstone Unit 3. This procedure further states that the primary

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g , responsibility for the performance of 'self-assessment activities resides within the line organizations, including identification and resolution of deficiencie In order to evaluate the effectiveness of the self-assessment program, the team reviewed a selected sample of (20 out of approximately 110) recently completed Unit 3 self-assessments. These self-assessments included a cross section of Unit 3 department

! - evaluations as well as follow-up reviews which were performed during 1997 and early

' 1998. As a result of these reviewsithe team determined that there was a significant I variance in the quality and technical depth of these self assessments depending on (1) the l

time frame in which the assessment was performed and (2) the organization performing the activity. .in particular, the self-assessments which were performed earlier in 1997 were in -

a number of instances narrowly focused with limited corrective actions specifie Examples of this category of self-assessments included the following reports:

• 3MNTC-SA-97-17, QC Hold point Requirements, February 5,1997

• 3MNTC-SA-97-01,In-Field Procedure Compliance, May 7,1997

• 3 CAD-SA-97-03, Root Cause Evaluation Quality, April 30,.1397 However, as noted by the team, the technical adequacy of self-assessments were improved in the latter half of 1997 and into the first quarter of 1998 demonstrating a broader perspective and more expansive corrective action recommendations. Self-assessment reports which reflected this improving trend included the following examples: 4

• 3MNTC-SA-97-07, Health Physics Department, June 24,1997

• .3TS-SA-97-13, Functional Requirements in Safety Systems Preoperational Testing, September 30,1997

• 3 CAD-SA-97-14, Assessment of MP3 and Site Level 1 Condition Reports, January 22.,1998

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- As determined by the team, the overall improvement in the quality of Unit 3 self-

- assessments was attributable to the establishment of definitive management expectations

- regarding the need for performance improvement, an emphasis on self-assessment training and enhanced procedural controls. Additionally, at the time of the inspection, the licensee was in the process of implementing a self-assessment program that incorporated both

- station and unit specific elements. Relative to this issue, the team ascertained that the

. station unit and support organizations have developed departmental self-assessment plans

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l with each major support group performing formal assessments using a common approac Accordingly, department self-assessments consist of an in-depth evaluation of significant line and staff activities, which are performed by teams of knowledgeable individuals in accordance with defined assessment plans. The assessment results are used to identify program strengths, findings and areas for improvement. The licensee's program has also been expanded to evaluate the effectiveness of the self-assessments in order to improve the quality and consistency of these activitie The team also evaluated the effectiveness of NNECO's program to identify and correct operator " work-atounds." The purpose of the work-around program is to identify and assess equipment deficiencies that adversely affect plant operations. These deficiencies are characterized as items that may degrade the operator's ability to react to plant transients. The inspection effort involved the review of the procedural controls contained in OP 3260E," Program for Resolution of Operator Work-Arounds", Revision 0, examination of Self-Assessment 3 OPS-SA-97-04," Effectiveness of The Operator Workaround Program" and the evaluation of recently completed modification packages related to operator work-around The inspection team determined that discrepancies associated with operator work-arounds are not documented on CR's nor are they required to be tracked on the licensee's AITTS system, but are administered under a separate program. In particular, procedure OP

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3260E, directs that operator work-arounds, initially identified in the shift turnover log, be screened for their cumulative impact, including operability and deportability consideration This screening process is performed by the Unit Supervisor, the Shift Manager and finally by the Operations Manager. Based on this screening process, the items that are classified as operator burdens or work-arounds are separated from the category of non-conforming conditions which would require corrective actions in accordance with the CR progra Additionally, these items which are tracked using the Operations Performance Database, are prioritized and periodically reviewed by the operations department, and they are reviewed on a biweekly basis by unit management personne Subsequent to reviewing the completed modification work packages for six operator work-arounds, the team performed a system walk-down in order to confirm the implementation of equipment modifications and the adequacy of the completed work activities. Based on the team's reviews, and the results of system walk-downs,it was determined that the equipment modifications associated with five of the operator work-arounds had been appropriately completed or were awaiting testing following component re-work. However, during the review of the documentation related to operator work-around Number 96-03, (correction of flow indication anomalies on service water instrumentation 3SWP-F1-059 A, B, and C), the team determined that the Trouble Report (TR) tags had been removed from the flow indicators and that the automated work order (AWO) associated with this modification had been inappropriately closed, prior to the completion of all specified wor Specifically, the final setpoint calibrations for flow indicators 3SWP-FI-059 A, B and C had not been accomplished prior to removing the TR tags and closing the AWO, which is contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions and also to the work control requirements of Procedure U3 WC 1, " Unit 3 Work Management," Revision 1, Section 1.8.7. This is a violation (VIO 423/97-82-08).

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The team also observed that the action taken by the AWO to correct an operator work l around item had not been identified as a CR. In this case, service water flow instrument operation was hampered by a system process valve position. Thus, the issue did not

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benefit from the controls that would have been imposed by the corrective actions process.

ll Following the identification of this issue, the licensee initiated CR M3 98-0942,in accordance with procedure RP 4, to document this deficiency and to effect resolutio Two contributing factors related to the above noted violation involved the (1) the lack of formal training for operations personnel on procedure OP 3260E and (2) the fact that

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Action Requests (ARs) are not initiated to track the status of each operator work-aroun These specific issues had been previously identified in Self-Assessment Report 3 OPS-SA-97-04 and although a CR (M3-97-3632)had been initiated on October 21,1997 to address the assignment of ARs to track discrete operator work-around items, the corrective actions for this item were not scheduled for completion until September 30,1998. Given that both of these contributing factors had been previously identified in a self-assessment l

.. report, this missed opportunity to avert an item of noncompliance is identified as a weakness within the self-assessment program, c. Conclusions The self-assessment program had established appropriate administrative controls which provided for the tracking of information to detect declining performance and adverse trends. The team concluded that the self-assessment program was being adequately implemented and that the associated recommendations were beneficial in identifying areas for enhancement and improved performanc One violation was identified relating to the implementation of NNECO's program to document and correct operator work-arounds, an AWO associated with a modification was closed prior to completing all specified work. The team also observed that operator work-around issues, which involve plant material deficiencies, were not included in the corrective actions progra .2 Self-Assessment of Design Basis issues - ACR 7007 a. Inspection Scope Evaluate NNECO's actions taken in response to ACR-7007- Event Response Team. This

' issue primarily concerned the historical inaccuracies in the Updated Final Safety Analysis Report (FSAR) and design basis documents. It is designated by NRC as Significant issue List (SIL) Item 4 b. Observations and Findings

- Background:

ACR 7007 was written by the licensee in January,1996 to address the high level concern -

- that "The UFSAR, system descriptions and design basis documents contain inaccuracies."

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An Esent Response Team was chartered to determine the causes of these inaccuracie This team used root cause analysis methods to identify causes and contributing factor This team developed the ACR 7007-Event Response Team Report, dated February 22,'

1996, that identified conclusions, corrective actions, and comments in Sections 5,6, & 7 of the report. It also included a significant amount of backup material that led to these :

conclusions. The report presented a thorough analysis of the problems and a reasonable

. set of conclusions and corrective actions. The identified causes and contributing factors were very broad and far-reaching, necessitating extensive corrective actions. These were further elaborated 'upon by the licensee as described belo In July,1996 the Fundamental Cause Assessment Team (FCAT) and the Nuclear Committee AssessmentTeam (NCAT) determined and reported to the NU Board of Trustees that the root cause of decline in Millstone performance was that senior executives-at NNECO from the CEO to senior nuclear site executives were ineffective over a number of years in providing vision, direction, and leadership necessary for the management of the NNECO nuclear _ power progra Over the last two years the licensee has embarked on a number of broad programs and changes to address these concerns and improve performance. These have included the configuration management program (CMP) and major changes in organization and management at the Millstone sit In order to address the specifics of ACR 7007, the licensee extracted all identified issues from the Event Response Team Report. These were compiled into a main item list of 104 issu'es, dated Nov. 5,1996. In June,1997, Unit 3 issued CR M3-97-1839 to address the Unit 3 aspects of ACR 7007 and the Event Response Team Report. This document sorts the 104 issues into five major areas in order to track and address them effectively. The five areas are: leadership, self-assessment, corrective actions, configuration control, and oversight. These categories are the same that have been identified by the corrective actions department. Additionally, they are the same as five of the eleven key site-wide issues for restart that are being tracked by NNECO and that are addressed in the periodic letter to the NRC titled, Progress Toward Restart Readiness at Millstone Station (NRC Briefing Book).

In order to provide more detailed tracking and specification of corrective actions, subcategories were established below the five categories. All of the 104 issues were then assigned to the .;ategories and subcategories. [However, three of the 104 issues were

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noted to have no action required, and 15 were noted to be applicable to Unit 1 only.] A succinct high levelissue was developed for each of the subcategories, and corrective actions were defined and approved to address each high levelissue. Action request (AR)

. numbers were assigned for all corrective actions. An important point to note is that this

' ACR and the related corrective actions addressed primarily the configuration control aspects of the five major areas and did not try to fully address other aspects of leadership,

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self assessment, corrective actions and oversight There will still be other NRC

. assessmentsisuch as the ICAVP assessment and NRC senior management assessments, of these five issues and the other remaining Key issues.

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As part of the CMP program, under procedure PI 2, " Unit Specific Assessments," Unit performed a self-assessment to determine any other areas similar to those in ACR 7007 that needed to be addressed. ACR .13302 was written to address the findings from that -

assessment. ACR 13302 was included as part of the ACR 7007 package and was also -

reviewed for this inspectio General Discussion:

The inspector reviewed the lists, discussed above, and verified that the issues had been appropriately extracted from the Event Response Team Report and that each of the 104 issues had been assigned to at least one of the categories / subcategories. The decision of -

!.- "no action required" on three issues was judged appropriate. For the 15 issues, designated as applicable only to Unit 1, the documentation.was not clear as to why they were not also applicable to Unit 3.. The licensee acknowledged this, performed further review of these items, and documented the review in Memorandum PES-38-055, Clarification for Cctrective Action Plan for CR M3-97-1839, Feb.12,1998. This

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memorandum provided justification for the 13 of the 15 Unit 1 items. The other two were

- determined to be generic to all units, however, it was further determined that the Unit 3

< corrective actions,~already completed for ACR 7007, encompassed these two items as wel The team also reviewed each subcategory and verified that the issues (assigned to the subcategory's defined as high levelissues) and the specified corrective actions were appropriate. A few areas were noted, requiring further justification, and questions were passed on to the licensee. With the exception of those noted below, all were satisfactorily resolve The NRC also reviewed, on a sampling basis, the documentation provided with the CR for the completion of the defined corrective actions. Some of the reviewed areas were then selected for further in depth follow up. Additionally, the inspector performed an

- independent review in the general area of the high level concerns to verify the thoroughness of the corrective actions. Some areas addressed in more detail included: the FSAR update process, setpoint control, an independent contractor audit of the Design

' Control Manual, and the Design Basis Summarie The inspector also verified that all actions for each of the three CR's were appropriately

- closed or that remaining open items are not significant and scheduled for closure on an acceptable time frame. Some areas designated for completion post startup were questioned. The licensee provided documentation that the work had been completed already or justification that its deferral was appropriate, except as noted belo _

Additionally, the inspector reviewed an internal audit of the corrective actions associated

- with ACR 7007, "The Independent Review Team Report on the Effectiveness of Correctiveness Actions Associated with ACR 7007, Rev.1,6/17/97," and noted that the findings were addressed by ' extensive actions taken by the licensee on ACR 7007 since the audit.~

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Specific Areas and issues:

(1) Final Safety Analysis Report Changes The team discussed the area of Final Safety Analysis Report (FSAR) change identification and control with licensee representatives and reviewed procedure RAC 03, Rev. O, Changes and Revisions to Final Safety Analysis Reports. The inspector also reviewed:

selected FSAR Change Requests, the computer tracking system for FSAR changes, currently related to the FSAR changes in process, the process for initiation of a FSAR change, the review and approval process, the facilitatory / owner of the FSAR and specific FSAR sections, the tracking of in-process changes, and the overall scheme for FSAR updates during the current outage. The licensee has processed a large number of changes (e.g., 596 in 1997) during the current outage due to CMP issue identification (called an updating FSAR change) and due to the large number of modifications being completed (called an up-front FSAR change). Approvals for the up-front changes are processed together with the modification, ensuring that appropriate reviews are given and that the a FSAR is kept current. This large number of changes has resulted in several update d

submittats to the NRC during the past year and has measurably improved the content of the FSA RAC 03, Attachment 7 contains a listing of the manager of the primary responsible discipline for every section of the FSAR, thus facilitating proper review of change request The procedure also has mechanisms for ensuring that in-process changes are consistent with each other. Time limits are set and, if processing time frames of the procedure are not met, then a CR is issued. The inspector reviewed the licensee's tracking and trending data. The area of FSAR change control was judged acceptabl (2) Design Control Manual Review One of the issues associated with ACR 7007 related to design control and the Design Control Manual (DCM). As a result of this, one of several actions was to contract MDM Services Corp. to perform an independent review of the DCM. MDM issued the report,

" Strategic Overview of Millstone Configuration Control Processes by MDM Services Corporation, Final Report," July 15,1997. In this report were a number of recommendations, which the licensee addressed in Memo PES-97-412, dated Dec. 31, 1997. However, a number of the recommendations were rejected with no justification, the licensee revisited the area and provided a new evaluation that adequately addressed all recommendations or provided justification for not addressing the ) NRC Commitments The area of " licensee commitments to NRC" was addressed in ACR 7007. The inspector also noted that Level 1 CR M3-97-1759," Trend identified in the area of NRC commitments," was issued in early 1997 to address then current problems in this are The licensee placed considerable effort into the identification of past regulatory commitments made to the NRC, in docketed correspondence, through the implementation of procedure PI 6, Licensing Reviews. Ongoing controls for commitments were established i

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Regulatory Correspondence Processing and Validation," and RAC 00, " Regulatory Commitment Management Program." The inspector revieined the list of CR's generated as a result of the PI 6 reviews and sampled the actions takeri to resolve them. No problems were identifie (4) Updating of Drawings l The inspector discussed the area of drawing control and updating with the MP3 design engineering personnel. The updating process is controlled by the DCM, Chapter 7, and by ,

EDI 30250. These documents define categories of importance for incorporation of  !

changes into the more important drawings more quickly. Category 1 or Operations Critical drawings require incorporation within 30 days of the DCN being released. Category 2A drawings have a 90 day guideline for incorporating outstanding changes, however, there is currently a backlog of about 3000 drawings. The licensee had no concrete plan for eliminating this backlog. In response to inspector questions, the licensee issued Memorandum M3-DE-98-0090that forecast a plan to work off the backlog over a two year period. This area will be reviewed during the NRC Operational Safety Team Inspection (OSTI). .

(5) Design Change Controlissues Significant items List (SIL),lssue 4, and the similar issues SIL 45 and 59, were not specifically addressed in the ACR 7007 close out package. Thase issues related to a potential weakness in prograrns that may have allowed drawing changes without generating a PDCR or a DCR. The licensee reviewed this area and prepared a documentation package to demonstrate that other controls existed to prevent such an occurrence. Further, the CMP utilized procedure PI-29, Unit 3 P&lD Walkdowns, to review P&lDs for discrepancies, and then evaluate and correct the One closely related concern, documented in ACR 7007, was that the process for controlling drawings had been weak, allowing drawings to be changed without a DCN or DCR and not ensuring that other design documents were updated. As a follow up to this concern, an Engineering Self-Assessment Report (3-ESAR-97-001)was performed to l

assess Category 8 Administrative DCNs for this general problem. The ESAR found that certain DCNs exceeded allowable criteria in that a MMOD or DCR may have been required to properly approve and document the change in question on the administrative DCN. ACR M3-97-0506 was written to take corrective actions on this specific finding. The inspector reviewed the corrective action plan (CAP) and the documented close out actions to address the CAP. The inspector noted that not all of the required actions were clearly documented as being comnlete. The inspector requested additional documentation from the licensee to show that the actions had been completed. The licensee's review then found that all of the actions for one assignment, AR 97003960-05,had in fact not been completed, even though the AR had been closed. This is contrary to RP 4, Step 1.12.4, which states that actions are to be closed out "WHEN assignment is complete." The two actions that did )

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not appear to be completed were: performing an MSEE reconciliation of 199 DCNs that document as-built conditions, and reviewing the 142 DCNs that initiated work and then generating an MSEE, MMOD, or DCR as appropriate. As a result, the licensee issued CR

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~ M3-98-0921,which documents the incorrect closure. AR 97003960-05 specifies four action (6) Setpoint' control The inspector evaluated the design control area cf setpoint control. The licensee does not-maintain all of the setpoints in one common database or area. However, Specification SP .

ST-EE 329, Use and Control of Master Setpoint Index, Rev. 2,10/15/97,provides a roadmap of the various documents, control mechanisms, and databases involved. Change control for the setpoints is typically via the DCM. NGP 5.23, Plant Design Data System

.(PDDS) Data Packages states that the Master Setpoint List provides the setpoint l calculation number and process and instrumentation setpoints for components designated l . by an instrument mark number, in order to evaluate this area, the inspector selected ten l annunciator windows and their respective setpoints, as noted in the corresponding alarm response procedure (ARP). The inspector requested the licensee to provide the design Information for each of the ten instruments involved, that would include the setpoint value and the basis for the setpoint (e.g., the calculation).

Based on the licensee's response, the inspector noted that the information for the ten

< selected setpoints was not contained in one consistent location but rather was in the~

Master Setpoint List, the Technical Specifications, the Westinghouse Precaution, Limitations and Setpoints (PLS) document (WCAP-10072), a Surveillance Procedure, and a calculation. Further, one of the ARPs was found to contaia an incorrect instrument as the initiating device for the alarm. The licensee wrote CR M3-93-0805,"Pzr Level alarm initiating device incorrect in ARP," to address this finding. Also, for one of the alarms, the l Saturation Trouble alarm for the inadequate Core Cooling System, the Master Setpoint List referenced an incorrect calculation. The licensee was unable to find a calculation that provided the basis for the setpoint of 15 degrees. Upon further exploration, the licensee determined that the setpoint should have been changed to 32 degrees (which would agree with a similar alarm in the Safety Parameter Display System - SPDS), as part of PDCR M3-93-121 (and as recommended by Memo NE-93-SAB-263 dated 6/14/93). This 32 degree setpoint is derived in the Millstone 3 Emergency Operating Procedure (EOP) Setpoint Documentation, Calculation # W3-517-981-RE, Rev. 6, dated 9/17/97. The licensee issued, CR M3-98-0935 to address this finding. The Saturation Trouble Alarm in question is described in the FSAR, Section 4.4.6.5, instrumentation for Detection of inadequate Core Cooling, as backups to the primary subcooled/superheat display on SPD ' The inspector noted that the methods for control and documentation of setpoint information appeared inconsistent, difficult to retrieve at times, and had the potential for allowing incorrect information to persist. In addition, the licensee performed an

. Engineering Self-Assessment of the Setpoint Control Topical Area,3-ESAR-97-015, dated 5/2/97 and this self-assessment also noted weaknesses in the Setpoint Control Program

that have not been addressed to date. The team has identified an additional issue to review the quality of setpoint control in the future (IFl 423/97-82-09).

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- (7) ~ Design Basis Summaries a As part of.the changes to the design control process made in addressing ACR 7007, the

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licensee eliminated their Design Basis Documentation Packages (DBDPs) and Syste Descriptions and established Design Basis Summaries (DBSs). These were created per-procedure U3 PI 29, Development of Millstone Unit 3 Design Bases Summary Document U3 Pl 29 states that the objectives of the DBSs, include providing a documented reference:.

for use 10 the design process for future modifications, to support technical reviews and safety evaluations, to support operability evaluations and determinations for continued operations, and to support review of Technical Specification changes and FSAR change It further states that the DBSs are integral to the Unit 3 restart program and are prepared

= for the Unit 3 Maintenance Rule (MR) Group 1 and 2 systems. The inspector reviewed the-set of DBSs against the list of MR systems and reviewed portions of selected DBSs. The

. inspector also reviewed methods for updating the DBSs, which are contained in the Design

Control Manual, Chapter 11, Attachment 5, Instructions for. Controlling and Revisin Design Basis Summaries,' Rev. 6, Change ' The team noted that there was no DBS for the Emergency Lighting System that was a moved from MR Group 3 to Group 2 during the summer of 1997 after the original DBS list-m was developed. He also noted that the licensee was unable to ensure that any changes tov the list of MR systems was reflected into the DBSs. The licensee initiated CR M3-98-0892 to address this issue. Further inspection disclosed a lack of full coverage for the Chemica :& Volume Control System (CVCS) in a DBS since this was a MR Group 1/ Group 2 system.-

Portions of the CVCS were included in the Emergency Core Cooling System (ECCS) DBS, but much of the CVCS system was not included in any DBS. The failure to include the emergency lighting system and the full CVCS in the-design bases summaries is a violation of 10 CFR 50, Appendix B, Criterion V and the requirements of PI 29 (VIO 423/97-82-10).

They noted that the original DBSs were prepared based on design information that had a o freeze date of June,1996. The DCM update controls began sometime after that resulting in a gap of coverage of about one year. This was identified in Memo MP3-DE-97-1616 and is being tracked by De_ sign Engineering, CR's, and by Oversight. While not yet updated, the gaps are clearly identified and are being satisfactorily tracke The team examined the overall control of DBSs, since the PI procedures are being phased out and the DCM only addresses a revision process. The licensee presented the Millstone NPS, Programs and Engineering Standards, Configuration Management Plan, Revision 1,

~ 9/23/97 which addressed the transition process from the CMP and the Pl procedures to a permenent organization. This document is to ensure no gaps in processes and that the going forward products are clearly identifie . (8) Safety Functional Requirements Manual-l 4 .The MP-3. Safety Functional Requirements (SFR) manual was developed to identify the key system level requirements that are reflected in the safety :malysis.- This provides design input and assumption information; primarily for the NSSS equipment. The actual plant NSSS calculations are proprietary and are maintained by Westinghouse. The SFR was l developed by NNECO and was reviewed and commented on by Westinghouse. Chapter 2

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of the manual addresses systems, Chapter 3 addresses the FSAR Chapter 15 safety:

analyses, and Chapter 4 addresses three programs (fire protection, SBO, and safety grade cold shutdown.) :The manual provides a valuable tool for the plant design personne ' The SFR manual, (MP Unit 3, Design Basis Documentation Package, Safety Functional .

' Requirements, DBDP-MP3-SFR)is controlled by procedure NGP 5.28, Design Basis Documentation Packages, Rev. 3,10/15/97. Step 1.1.2 of NGP 5.28 states that if changes are' required they must be documented as DCNs and the DCN numbers entered into GRITS (Generation Records information and Tracking System). GRITS is the on-line interactive database system that provides the current design and revision status for the Millstone facility. The original Revisicn O of the SFR was issued 12/30/94. Revision 1 addressed Westinghouse's review comments and was issued 12/11/96. Revision 2 was issued on 11/20/97 to incorporate CMP changes.- However, Revisions 1 and 2 were issued with Engineering Record Correspondence per NGP 5.31 rather than NGP 5.28, and as a result no DCNs were issued and GRlTS was not updated. As of February,1998, 1 GRITS still showed Revision 0 as the latest version. This constitutes a failure to follow procedure NGP 5.28 and is a violation (VIO 423/97-82-10). The licensee issued CR M3-98-0861 to address this proble c. Conclusions The team identified several violations and problems associated with the licensee's actions taken in response to ACR-7007. However, the overall response has been a thorough and comprehensive effort to address the issues which were extracted from ACR-700 The team concluded that acceptable processes for FSAR change control are being applie An AR was closed without accomplishing the specified corrective actions; this is a violation. The Master Setpoint List was found to contain incorrect information for certain nonsafety systems, the methods for control and documentation of setpoint information appeared inconsistent and the findings of an Engineering Self-Assessment of this area have not been addressed to date. This is a Follow Up item. Violations of design control procedures were identified concerning Design Basis Summary documents and concerning the Safety Functional Requirements Manual. Updates of 3000, Category 2A drawings are planned over a two year period. This category has a 90 day guideline for incorporating outstanding changes. SIL ltem No. 41, concerning the ACR-7007 - Event Response Team findings, is close .0 INDEPENDENT OVERSIGHT 4.1 Effectiveness of Nuclear Oversight - Audits and Evaluations a. Inspection Scope.

I ~ The team' reviewed the activities of the Audits and Evaluation Group of the Nuclear Oversight Organization. This review hvolved discussions with auditors, managers, the

, > director of the audit group and the vice president, Nuclear Oversight. Documents reviewed

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included but were not limited to the bluclear Oversight Group procedures, a sampling of u, recently performed audits, tracking and review of audit findings, adequacy of audit E

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35 findings, scheduling of audits, training and qualifications of auditors, adequacy of staffing ,

adequacy of audit corrective actions, and Nuclear Oversight Group self assessment. In addition to the Audit and Evaluations group, the following activities were also reviewed:

• Nuclear Oversight Group actions to improve on deficiencies identified by independent outside assessments of quality assurance (QA) performed during 199 Specifically addressed were a sampling of actions taken by the licensee to respond to criticisms identified in the Joint Utilities Management Assessment (JUMA)

performed in July,1996;

• Surveillance and quality control inspections performed by the Performance Evaluation Group concerning repairs made to a Unit 3 emergency diesel generator; and,

• Various activitie's performed by Oversight implementing the Nuclear Oversight Restart Verification Plan (NORVP).

b. Observations and Findings The review of Nuclear Oversight involved personnel, programs, procedures, audits, audit

. follow up, audit scheduling, certain aspects of Performance Evaluation and the NORV The observations and findings for each area are discussed separately below:

(1) Personnel Since 1996, the auditor staff has been increased from five to twenty auditors. The audit staff has a significant background in various technical areas including operations. In addition, there is an auditor qualification and training program in place. There are now four audit managers over the areas of operations, maintenance, engineering and technical support, and plant support; and an overall director of Audits and Evaluations. This increase in audit management attention and audit personnel has improved the quality of audit findings. There is now a stronger interface with the line organization. Licensee audit teams typically have four or five members and take up to two weeks to perform. For this reason, support for the audit team is periodically obtained from contractors, the line organization, or other groups within the Nuclear Oversight organizatio (2) Procedures The team reviewed the following procedures:

- NOOP 1.05, Self-Assessment Process, Revision 0, June 30,1997

- NOQP 1.06, Nuclear Oversight Resolutions issues, Revision 0, November 12,1997

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NOQP 2.01, Nuclear Oversight Audits, Revision 2, November 20,1997

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- NOOP 2.02, Qualification / Certification of Audit team Leaders and Orientation of Team Leaders, Revision 2, November 20,1997

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- NOOP 3.02, Analysis of Quality Program Performance, Revision 0, August 29,-

1997

- NOOP'3.03, Nuclear. Oversight Assessments, Revision 0, May 12,1997

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. NGP 3.19, Procedure to Stop work, Revision 1, July 30,1997

- RP.4, Corrective Action Program, Revision 5, September 5,1997 The'te'am observed that all the procedures listed above were established or significantly

- revised during 1997. However, not all were new procedures as they may have existed in another form in the previously established QA program. The procedures appeared to be_-

comprehensive, clearly written and user friendly. There were procedures for all aspects of

' the program. It was determined from the review of the procedures that a new audit program had been established with more strength than the previous QA audit syste (3) Audits The team reviewed the following audits and audit checklists:

- MP-97-A1103, Software QA, performed November 3 - 14,1997

- MP-97-A09-01, Fire Protection, Performed September 8 - 26,1997

- MP-97-AOS-02, Chemistry, Performed May 27 - June 6,1997 I

- MP-97-A10-07," Operating Experience" Program, November 10-17,1997

- . Au'dit checklist for Audit MP-97-A09-02, Security

- ' Audit checklist for Audit MP-97-A06-03, Systematic Approach to Training Procedure NOOP 2.01 has strengthened the audit process by more clearly defining audit expectations, audit checklists and the makeup of audit teams. A review of the above audits and audit checklist indicates significant improvement in the audit process. Audits are generally two weeks in length and performed by a team. This has led to a more in depth look into each are Significant audit findings are issued as Level 1 Condition Reports (CR's) causing them to receive a high level of attention as required by procedure RP 4. An initial 7 day audit report is issued with the CR findings. The audited organization must respond within 30

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- days and the audit group must agree with proposed corrective action as stated on the l initial CR response.1 After agreement on the CR response to the findings, a more complete

- 45 day audit report is issued. 'This process ensures significant QA involvement in the corrective action process._ Less significant audit observations are issued as level 2 CR' ,

p Lin the past, audit exits were apparently given a low level of attention and were meagerly

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l attended by the line staff,' and rarely attended by appropriate management. The team

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reviewed a sampling of audit exit attendance sheets, and observed that the exits were well attended by both line staff and manager (4) Control of Status of Audit Findings:

As stated above; audit findings are controlled by the CR process. It is the responsibility of the line organization to provide corrective actions. The corrective action group for each unit maintains a status of open CR's. However, they do not necessarily track _CR's as audit findings. All due or overdue Level 1 CR's are treated equally. The team had a concern that audit findings may be lost in the large number of open Level 1 CR's.

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,The audit group has established its own computer tracking system of open audit finding This information is provided to audit managers and certain line personnel. It is used as a tool to track overdue or_ inadequate audit corrective actions. The audit open item is tracked by both the CR number and associated assignment request numbers (ARs). Recently, the Treport has been improved to clearly show a summary of the open or incomplete issu L This audit finding tracking system provides an effective mechanism to ensure that audit findings are not lost in the " shuffle" and to ensure that audit managers have a tool for managing the follow up of audit finding ' All audit findings receive a follow-up for adequacy of corrective actions; although this may not be a 100% verification of each action. Audit observations are followed-up on a sampling basis. Audit follow-ups may be done independently or during the next audit of the subject area, inadequate corrective actions are identified by the issuance of a new C The team noted that this was done concerning corrective actions identified in a fire

_ protection audit. The details of these audits are further discussed in NRC Inspection 97-8 (5) ' Audit Scheduling and Planning (Open VIO 50-423/96-05-12)

The team observed that an audit schedule had been established for 1998 and a projected schedule for 1999. These schedules just covered general areas to be audited and not detailed audit objectives. Actual audit planning is done by the team leader in conjunction with the manager for that area. Audit planning is established in NOOP 2.01. If not already done, the adequacy of corrective actions for previous audits in the area to be audited. The team verified that such audit planning is accomplished for each audi ~

' On May 29,1996, the licensee wrote an ACR that audits of TS 6.8.4.e, " Accident Motoring Instrumentation," may have been missed. Based on this ACR, a review of audits

. of the technical specifications, and a concern that TS audits were neither comprehensive nor well documented was previously detailed in inspection Report 50-423/94-2 Violation 50-423/96-05-12,was issued stating that "The failure to audit technical specification section 6.8.4.e within a five year is a violation of TS 6.5.3.7..." In its

. response dated September 16,1996, the licensee stated in part that, "The process for the

independent review of "The TS Audit Matrix" will be revised to require a review of "what

' has been done" versus "what has been scheduled." In the future, persennel will enter

- information into a tracking database after the audit has been completed and after the

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independent reviewer has checked "what has been done." Procedure...will be revised byL

- October 31,1996."

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The NRC team verified NOOP 2.01 has been revised to require an audit commitment database ~. A database has been established for general audit commitments and a separate one for all technical specifications. Further discussions indicated that while a data base had been established to determine what technical specifications had been audited and what needed to be audited, it was difficult to use as a " tracking and scheduling" tool. The Llicensee had already discovered the same thing during a recent self assessment of th audit progra ~

Memorandum AE-98-4037, dated January 19,1998, issued report no. 97-AE-08, "Self- .

Assessment of the Audits and Evaluation Group's Audit Commitment Tracking Process."

'.The self-assessment was performed December 1-5,1997. The executive summary'of this assessment stated, in part, the following:

"The audit commitment database, as'it is presently configured, does not meet the needs of the Audits & Evaluation (A&E) Group. This database is not a tracking database but is strictly a database for recording information. As a result, the A& Group can not use this database to track when a commitment was addressed, by

.what audit and the deadline for addressing the commitment again. There is not a high level of confidence that all audit commitments listed in the 45 day audit report have been completely addressed...ATLs laudit team leaders] and Managers are not consistently using the database and, in some cases, do not completely understand the purpose of the database. Finally, the database is missing some data and contains some data that is no longer relevant."

It should be noted that the commitment database includes all audit commitment sources such as NRC regulations, the FSAR,10 CFR Part 50 Appendix B, ANSI Standards, regulatory guides, INPO, etc. and not just technical specifications. The licensee was ultimately able to' generate a list of all TS in their database and those already audited; however, this was not being used as a scheduling tool. The Director of A&E stated there is a commitment to NSAB to resolve this issue by June,1998. Although much of the ,

corrective' action for the violation has been accomplished, this violation remains open

. pending the development of an effective scheduling too: to ensure that all technical specifications will be audited within a five year period in addition, the remaining commitment database is lengthy. and cumbersome. The licensee stated they intended to

" scrub" the database of unnecessary and obsolete commitment (6) Nuclear Oversight Review Plan (NORVP)

The NORVP was established to provide independent oversight for readiness to restart and for heat up (mode 4). The NORVP meets at least weekly and frequently interacts with the-line organization. Among the documents reviewed were the following:

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• - , Work Control and Planning Brief (2/5/98)- Presents work control and planning for restart from a Nuclear Oversight perspective

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• Nuclear Oversight Restart Verification Plan - Presents Nuclear Oversight numerical evaluation of progress in the areas leadership, corrective action, NSAB/ oversight, configuration management, engineering, maintenance /l&C, regulatory compliance, radiation protection, conduct of operations, security, procedure quality & adherence, work control & planning, training, conduct of operations, and fire protection

• Millstone 3 Mode 4 Attribute Summary,2/9/98

• 50.54(f) Recovery Oversight status report for the week ending 2/7/98 Review of the above documents and discussions with Nuclear Oversight personnel indicate significant oversight involvement in the Millstone recovery process. The reports were indicative of an extensive Oversight review. Nuclear Oversight meets three times a week to discuss station progress toward restart. Deficiencies observed in the startup process are discussed with the line organization. This process appears to be effective to ensure that senior management is provided with -

an independent evaluation of the restart proces (7) Self-Assessment Self-assessments have become a significant portion of the licensee's self improvement program. All Millstone organizations are required to perform self assessments including the Nuclear Oversight Group. As noted above the Oversight self assessment procedure is NOOP 1.05, Self Assessment Process, Revision 0, June 30,1997. The team reviewed the following self assessments:

• AE-97-S2-02- Final Self Assessment of : Audits and Evaluations Compliance With NOOP 2.01, Rev.0, " Nuclear Oversight audits", Dated May 16 1997

• AE-97-S2-03- Self-Assessment Report of Training and Certification of Lead Auditors / Auditors in Training and Technical Specialists, Dated May 19,1997

• Self-Assessment Report of Performance Evaluation Strategic Plan Development and Assessment of Organizational Effectiveness

• 97-PE-01 - Self Assessment Report.of Applying PE QC Hold Points and AWOs by PE QC AWO Reviewers, Dated January 6,1998

• 97-PE-05 - Self-Assessment Report of Performance Evaluation's Work Process, Dated January 12,1998 l * 97-AE-08 - Self-Assessment of the Audits & Evaluation Groups's Audit Commitment Tracking Process

.The team observed that the self assessments were in-depth and effective. The self assessment process allows for process deficiencies to be identified and brought fo. ward in l' ' a formal manner. A response to the NSAB is required and corrective action commitments must be made. The self-assessment process is proactive rather waiting for problems to

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arise or to be discovered by outside assessment groups. A's noted above self assessment 97-AE-08 noted significant weaknesses in the audit scheduling process. " DRAFT" Nuclear Oversight Plans for Audits and Evaluation for 1996-2OOO, identifies future planned self assessment (8) Performance Evaluation The Nuclear Oversight Performance Evaluation Group performs surveillance of work

- activities and quality control hold point inspections for specific work activities. The surveillance technicians and quality control inspectors comprise two separate groups. The term surveillance as described in this section applies to periodic oversight of various work activities and does not re'v to surveillance testing as required by the Technical

Specifications. Surveillance activities are controlled by the following procedures:

• NOOP - 4.02, Performance, Reporting and Follow-up of Surveillance Activities and Field Observations at the Millstone Station, Revision 1, Dated May 20,1997

• NOOP - 4.09, Planning, Scheduling, and Administration of Quality Surveillance Activities, Revision 1, Dated December 19,1997

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There are 27 surveillance technicians , of which, nine are assigned to Millstone Unit 3 i activities. There are 23 QC inspectors. For both the technicians and inspectors, there is a

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mix of permanent employees and contract specialists. Most technicians and OC inspectors were former maintenance workers or l&C technicians thus giving them credibility with the personnel performing the maintenance , ,

The team reviewed documentation of the following completed surveillance activities:

• - MP3-P-97-132, System Engineering Communication, September 23,1997

• MP3-P-97-105, Cortical Maintenance, Dated September 23,1997

• MP3-P-97-123, Shielding Program, Dated October 7,1997 l

• MP3-P-97-136, Control of Overtime, Dated November 3,1997

• MP3-P-97-149, Procurement - Vendor Control, Dated December 4,1997 i,

j '* MP3-P-97153, AWO Quarterly Review, Dated December 6,1997 L

.* MP3 P-97-118, Material Condition - Field Walkdown - Housekeeping / Material l

. storage, Dated December 30,1997 L

• MP3-97-154, Minor Modifications, Dated January 8,1998

• - MP3-P-98-OO1, Conduct of Operations, Dated January 13,1998

• ' MP3-P-98-OO5, AWO Quarterly Review, Dated January 19,199 .

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m. Surveillance activities for major areas and planned activities are~ scheduled approximatel six months in advance. - The team reviewed the current six month schedule which called L for 27 surveillance for Unit 3 alone. Surveillance for emergent maintenance activities are -

scheduled as they occur. Not all jobs are reviewed; but, there is an attempt to provide

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some oversight of most major work activitie c. Conclusions l

l, The team determined that the Nuclear Oversight organization is effective in performing l audits, general plant oversight,~ and work surveillance activities. Considerable improvement is noted since independent assessments identified substantial weaknesses two years ago in

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.the performance of QA activities. Procedures and audit quality have improved. Audit '

findings are more meaningful and there is good control in the follow-up of audit finding Audit scheduling has improved but there are still weaknesses that have to be resolve There is much better communication between the line organization and nuclear oversigh : 4.2 Effectiveness of Nuclear Oversight - Quality Control a. inspection Scope The effectiveness of the Quality Control (QC) department was assessed by reviewing .

procedures, interviewing personnel, and accompanying inspectors on inspection b. Observations and Findings The QC Inspectors interviewed were experienced and qualified in their areas of expertis The licensee utilized a mix of staff and contract personnel to perform the QC inspection function. OC inspectors were knowledgeable of the site work control process and documentation. All of the inspectors stated they would stop work if required. During-observations of inspections in the field, two inspectors stopped jobs due to questions with proposed signoffs. All of the interviewed QC inspectors stated that they now felt they had {

management support to stop jobs if require ]

During this inspection, the team watched one surveillance activity in the field.' Testing of the Unit 3 "B" emergency diesel generator was stopped due to a leak on the air start lin The leak had been caused by chafing between the air line and the fuel line. AWO M3 -98 -

2900 was issued to replace and reroute both lines to eliminate the tubing contact. While there was no specific procedure for this job, generic procedure CMP 721 A, " Installation of instrument Tubing, Fittings and Supports" was used as the basis for the tubing installatio OC hold points were established by this procedur ' The surveillance' was comprehensive and reviewed all aspects of the job. There was good interface between the surveillance technician and the maintenance workers performing the jjob.: Although not witnessed by the team, the surveillance technician briefed the workers

- as to the results of his oversight at job completion. Field observation checklist MP3-P-98-004-F18 was issued on February 12,1998, giving the results of the surveillance. The surveillance identified one deficiency concerning the fact that a systems engineer at the job l

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sight did not have a controlled copy of the drawing in use. However, the maintenance workers di The team also observed a QC inspection of portions of the foregoing job. As part of the hold point, the' inspector verified tubing connections and tubing bend radii; and, then

. signed off _a verification sheet. The QC inspector stated that he'also looks at more than

.Just what is called for in the hold point. Overall, the QC inspection appeared to be acceptabl The teams evaluation of both the surveillance and QC inspection activity is that they.were effective and comprehensiv ' Although the observed inspections were g6nerally rigorous, the NRC observed one instance that indicated an inspection performed as part of the spent fuel pool foreign material-exclusion (FME) program was not adequately performed. The purpose of the inspection

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was a monthly requirement to sign for a list of objects allowed within the FME barrier an to determine that no other objects existed within the exclusion zone.' The NRC inspector noted a row of tape pieces hanging from the underside of the spent fuel pool bridge. The tape was not listed as authorized and was not noticed by the QC inspector. Although the -

tape was not noted by the QC inspector, the QC inspector did raise other concerns with staging that was in the area and not tracked with the list. A condition report was written to document the tape; however, the response went to the reactor engineer and was being treated as only a technical issue. Through interviews of the QC manager, the inspector identified that the QC inspector issue was not being addressed. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio The QC support group, which was developed to standardize the inspections performed by the QC inspectors, was a strength. - The QC support group reviewed all work packeges to identify hold points prior to the packages going to the field. The group developed standardized inspection points for many routine work activities and group members coordinate closely with each other to assure standardization. Additionally, the pup had a rotational assignment that was filled by an inspector from the QC inspection staff. The purpose of this position was twofold. It brought the QC inspector's perspective into the process and trained the inspectors on the group's policies and procedures. This was essential because emergent work packages developed after normal working hours were processed by the field inspectors in the absence of the QC support grou c. Conclusions Quality Control was generally effective in performing the required in-plant inspection However, the spent fuel pool FME inspection was noted as an inspection that was not adequately performed. The QC support group was effective in establishing and

. standardizing the use of OC hold points in work package ,

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43 4.3 Followup of Previously identified Inspection Findings - Nuclear Oversight 4.3.1 (Closed) Inspector Follow Item (IFI) - 96-06-17, Joint Utilities Management Assessment (JUMA) Concerning the Effectiveness of QA; (Open) Violation - 50-423/96-05-12; Failure to Audit All Technical Specifications Within a Five Year Period; (Closed) MC 0350, Restart Checklist items C.1.4.a,b &c and C.2.1.c; l (Closed - SIL ltem No. 73) 'I The JUMA was performed during June,1996. A subsequent NRC inspection Report,96-06, referring to the JUMA results stated, in part, that:

"The JUMA team concluded that the audit, surveillance and inspection programs at Millstone were not effective in the implementation of their mission statement and the resolution of identified problems. The [JUMA) team attributed these problems to:

• Lack of support for the OA organization by the executive and line management

• Lack of an effective action program" The NRC made the licensee JUMA corrective actions an IFl in inspection Report 96-0 Section 4.1 of this report noted an increase in the effectiveness of the Nuclear Oversight Organization and increased support for the Nuclear Oversight organization by the NNECO President and CEO. There is increased interaction with the line organization and Oversight is involved with corrective actions to deficiencies identified by them. The licensee showed the team documentation that they had been responsive to all JUMA concerns. Section of this report demonstrates improvement in the licensee's corrective action process. It is not the intent of this inspection to make judgement as to the adequacy of the licenser's response to the JUMA. However, because of NRC observations made during this inspection, IFl 96-06-17 is considered close The licensee's response to violation 96-05-12 concerning the scheduling of audits of the technical specifications is discussed in report section 4.1. Although, the licensee has made considerable progress in this area, they are not yet effectively using their database to schedule TS audits. As stated in section 4.1, this violation remains open pending completion of all licensee corrective actions, but, based on corrective actions already taken, this aspect of SIL ltem No. 73 is close An issue identified by the JUMA audit team involved OC inconsistently assigning hold points in work packages. This issue was addressed by the licensee by establishing the QC support group. The procedures for reviewing work packages and assigning hold points were formalized. The QC support group reviewed all work packages to identify hold points l prior to the packages going to the field. The group developed standardized inspection points for many routine work activities and work members coordinate closely with each other to assure standardization. Additionally, the group had a rotational assignment that

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L .'was filled by an inspector from the QC inspection staff. The purpose of this position was i

- twofold. It brought the QC inspectors perspective into the process and trained the

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. inspectors on the group's policies and procedures. This was essential because emergent L work packages developed after hours were' processed by the field inspectors in the absence of the QC support group.

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The inspector reviewed several work packages and concluded that the inspection points

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were appropriately selected. Based on the program improvements and implementation of L the process by the' OC support group, the inspector concluded that the issues identified in l '

the JUMA report were adequately addressed.

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NRC Manual Chapter 0350, " Restart Approval Checklist" includes the following it ,ms:

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• -C.1.4.'a Effectiveness of quality assurance program

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• C.1. Effectiveness of industry experience review program

• . C.1. Effectiveness of licensee's independent review group

l *- C.2. Management involvement in self-assessment and independent self-assessment capability Each of the above areas are addressed in various parts of this inspection report and have been found to be sufficiently acceptable to close these areas. Based on the above, SIL ltem No. 73 is close .3.2 (Closed) Unresolved item (URl) 50-423/95-81-01- Lack of Trending of Non-conformance Reports (NCRs) and Level "D" ACRs; (Closed - SIL ltem No. 41)

Inspection Report 95-81 stated, in part, the following: "...QA did not trend any NCR or

- look for adverse trends which may be discernible from such data... the lack of level "D" ACRs may mask a recurring problem and its significance. Formalizing the practice of trending NCRs and level "D" ACRs would address a potential weakness in the program, and make the process less reliant on individual analysis and perception of the significance of the recorded proble The acceptability of the lack of trending of NCRs and verifying the effectiveness and adequacy of the ACR database by the Quality Assurance Department remains unresolved..."

Since Inspection Report 95-85 was performed CRs have taken the place of ACRs and level

"D" ACRs have been ' eliminated. The team verified that NGP 3.05, "Non-Conformance reports" has been revii.ed to have NCRs trended in accordance with procedure RP 4,

" Corrective Action Programs." Also 'a CR is written for each NCR issued. Procedure RP 4, Paragraph 1.17, states ."at least quarterly, PERFORM trend analysis of AITTS database related to CRs and ISSUE trend report within 30 days of the end of the of the quarte Trend analysis shall include human errors associated with NOVs and LERs...BRIEF

- appropriate levels of organizational management of results of trend analyzed trend

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data... ENSURE CRs are initiated for any adverse trends identified through periodic trending"- The team reviewed recent trend reports and verified that CRs and NCRs (which are trended separately) are now trended at least quarterly. Based on the above review, URI 95-81-01 is close . The licensees actions taken in response to the findings of their ACR-7007 task force was-addressed in report section 3.2, above. This closes SIL ltem No. 4 .4 Performance of the Nuclear Safety Assessment Board a. Inspection Scope

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The performance of the Nuclear Safety Assessment Board (NSAB) was evaluated by reviewing board meeting minutes, observing a board meeting, and interviewing board member 'b. Observations and Findings

' The NSAB consisted of senior plant managers and two independent contractors. All of the board members met the Technical Specification (TS) requirements for education and background. Section 6.5.3 of technical specifications specifically stated the areas of expertise required for representation on the NSAB. To assure compliance with these requirements the board secretary maintained a matrix of qualifications of the board members. The list, as originally presented to the NRC, did not indicate any regular

. members as having methilurgy experienc As a result, an alternate member with specific metallurgy experience was added to the list

- of alternate board members in February,1997. However, based on the review of the qualification matrix the NRC concluded that the technical specification was not met because no regular member possessed metallurgy experienc The licensee re-evaluated the qualification matrix and concluded that their previous screening of qualifications was too conservative. The previous screening criteria required an academic degree or direct experience to be counted as a metallurgist. However, this screening was in excess of the technical specification requirements. Based on the new screening criteria two current members were identified with metallurgy experienc Although the NRC did not contest the actual board member's qualifications, the evaluation and resolution of this discrepancy in.1996/1997did not meet the literal interpretation of the technical specifications. This was an example of non-conservative interpretation and implementation of technical specifications. This is failure constitutes a violation of minor

, significance and is not subject to formal enforcement actio y

. Portions of a NSAB meeting were observed on January 29,1998. The meeting consisted largely of pr'esentations to the board by various managers regarding departmental readiness for restart. Several board members asked probing questions and displayed significant knowledge of the issues and obviously were prepared for the meeting; while, in contrast, l there was only limited participation by a few of the members.' The board identified a potential safety issue with fire protection systems that have outstanding surveillance tests, m

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4 The board questioned the acceptability of prolonged use of compensatory measures. The fire protection program manager was requested to respond to the board at the next meeting. Action items assigned during meetings were tracked by the NSAB secretary and the status of open items were included as part of the meeting minute packages. _ Closure .

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of open items required consensus of the board membership.

i Another example of the NSAB providing appropriate oversight of plant activities was weaknesses identified in the area of training. As a result of NSAB intervention, the training on site was stopped pending program improvements.' Additionally, the board thoroughly probed the area of operational experience (OE). To validate a presentation by the OE program manager lthe board questioned various department managers on the use of OE by their departments during subsequent presentation ,The NSAB audit program was appropriately implemented by the Audits and Evaluation Department. The scope of the audit plan was reviewed by the board, and audit results-were presented to the board by the auditor .c. Conclusions The NSAB was effective in reviewing activities on site and identifying potential nuclear safety issues. The implementation of the NSAB met the technical specification requirements, however the 1997 resolution of an issue of membership qualifications did not meet technical specifications as presente ' 4.5 Performance of the Plant Operations Review Committee a. Inspection Scope The performance of the Plant Operations Review Committee (PORC) was assessed by li reviewing committee meeting minutes, observing board meetings, and interviewing board

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members, b. Observations and Findings

= During observed PORC meetings, the membership met the technical specification requirements. PORC members were prepared for the issues on the agenda and asked technical questions of the presenters. The questions focused on safety and compliance with regulatory issues. The PORC meetings were conducted in a professional manner with g an emphasis on clear communications between the committee and the presenter Station expectations for the conduct of PORC were high as demonstrated by several assessments of PORC performance. :The PORC process was being evaluated at the station leyel as including the performance of the committee as well as the personnel presenting-items to the committee. Instances such as presenters being unable to answer PORC

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questions and PORC-rejected documents were considered weaknesses by the licensee. An i effort was underway to develop a feedback process to track and reduce poor presentations to POR .

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PORC issues were tracked in the plant action request tracking system by the PORC secretary. A review of the backlog indicated timely resolution of PORC issue c. . Conclusions

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PORC was effective in accomplishing the reviews required by technical specifications.-

Station evaluation of PORC performance and the standards being demonstrated set high standards for the documentation and review of technicalissues.-

4.6 Performance of the Site Operations Review Committee a. Inspection Scope The performance of the Site Operations Review Committee (SORC) was assessed by reviewing committee meeting minutes, observing board meetings, and interviewing board

. member b. Observations and Findings During observed SORC meetings, SORC membership met the technical specification requirements. However, technical specifications designate that the Senior Vice President and Chief Nuclear Officer (CNO) shall be the SORC chairperson. A review of SORC meeting minutes demonstrated that the chairperson responsibility was normally delegated to the Director of Unit Operations. The basis given for this delegation was the fact that the Chief Executive Officer (CEO) was acting as the CNO and a potential conflict existed .

because technical specifications state that disagreements between the CNO and the SORC "

shall be brought to the attention of the CEO. The team did not take issue with delegation of SORC chairperson responsibilitie SORC members were prepared for the issues on the agenda and asked technical questions of the presenters. Members adequately represented the site-wide perspective of SOR This site presents a particular challenge with the differences in license requirements i between the units. SORC members displayed a combination of knowledge to integrate site-wide license and technical requirements. This was evident in a discussion of fire

. protection issues which required detailed knowledge related to all three units. During one meeting, the inspector noted four occasions that safety issues which required further evaluation were identified by the committe SORC issues were tracked in the plant action request tracking system by the SORC

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secretary. A review of the backlog indicated timely resolution of SORC issue c. Conclusion . SORC met the technical specification requirements and was effective in identifying

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potential safety issue j

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4.7 Performance of the independent Safety Engineering Group a. Inspection Scope The performance of the Independent Safety Engineering Group (ISEG) was evaluated by reviewing ISEG reports, reviewing the operating experience (OE) program implementation, reviewing ISEG backlogs, and interviews of personne b. Observations and Findings Millstone unit 3 technical specifications required an ISEG consisting of four full-time personnel to perform independent reviews of plant operations. The Unit 3 ISEG consisted of three full-time engineers, one part-time contractor, and a supervisor. The ISEG charter required the group to use operating experience when reviewing plant operations, to make detailed recommendations to improve safety and reduce human errors. In implementing

this charter, the licensee elected to make the ISEG the site group responsible for reviewing and implementing OE. OE typically comes in the form of industry group reports, vendor notifications, and NRC informatio Over the past two years, the ISEG focus appears to have shifted from the ISEG activity of independent reviews of plant operations to the performance of OE reviews. A large backlog of unreviewed OE items was processed which resulted in a reduction of the number of ISEG activities performed. The number of ISEG reviews done in 1997 was only 12, down from 24 the previous yea Reviews of plant operations performed by the ISEG group were documented in report Reports reviewed by the inspector indicated that the ISEG performed critical reviews and made appropriate recommendations. Examples included an ISEG review of high voltage switchyard work, and an evaluation of work activities in the spent fuel pool. The ISEG group used the action request (AR) process to track the implementation of recommendations and performed closeout inspections of each ite The inspector followed up on the condition report from the ISEG review of the high voltage switchyard work. This ISEG evaluation identified several significant issues which involved personnel safety and the potential loss of off-site power. The initial response to the issue

! was good, however the condition report was processed as only a level three. This was not identified or challenged by the ISEG group. NHC review of this issue indicated that the condition report should have been processed at a higher leve By focusing on operatirig experience, the ISEG group reduced the backlog of OE issues from several hundred to approximately 40 for unit 3. A sample of OE packages was reviewed and were to be found generally thorough and complete. However, one weakness was identified in the review conducted for NRC IN 97-14 related to spent fuel pool coolin The information notice suggested reviewing siphon breakers used to prevent spent fuel l pool draining. The OE review confirmed that the siphon breakers were installed, however

. the reviewer stated no future inspection or maintenance was required. The inspector considered that the issue of potential fouling of the siphon break orifices or the need for

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periodic inspections was not adequately evaluated. The licensee was reviewing this issue at the close of the inspectio Although the backlog of OE issues had been significantly reduced over the past year, the amount of work represented by the remaining issue's was significant. Several of the issues remaining in the backlog had a high probability of identifying safety significant issues. One example is NRC IN 97-78, which involved identifying unreviewed safety questions related to the use of manual actions in place of automatic actions in emergency operating procedures (EOP's). The review of this issue was completed for unit 2 which identified

- several areas of concern. Additionally, a previous NRC inspection identified an issue

- involving manual action for a control room ventilation system that cannot be accomplished within the time assumed in the safety analysis. Based on these factors the priority of evaluating this issue appears inappropriat Site implementation and use of OE was mixed. OE is not consistently being used by the working groups at this time. A key reason is that the site-wide procedure to establish the expectations for the use of OE was in development and not yet issued. Once a site-wide procedure is issued, the departments still have to develop implementing procedures. The OE Minute, a daily publication listing relevant OE items, appeared effective in disseminating OE to the site on a daily basis. However, the information was not easily retrieved after the fact. Industry and NRC information was sent to appropriate personnel (system engineers, operations) for information, however the site responsibility for evaluating and identifying actions was with the ISEG group. Access to the industry nuclear network data base was improving, but was still lacking in some areas such as system engineerin c. . Conclusions The ISEG was staffed and met the technical specification requirements. However, the implementation of the OE reviews by the ISEG was limiting the number of independent reviews of plant operations performed by this group. ISEG independent reviews and ISEG i reviews of OE were generally thorough. Corrective actions for ISEG items were tracked and verified by the ISEG prior to closur .!

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.e:

e INSPECTION PROCEDURES USED IP 40500: Licensee Self Assessments Related to Safety issues inspections IP 37550: Engineering IP 71750: Plant Support Activities IP 92720: Corrective Actions IP 92903: Follow-up Engineering ITEMS OPENED, CLOSED and DISCUSSED Opened I

URI 423/97-82-01 Review of measuring and test equipment program after 1 implementation (Section 1.3.2).

. VIO 423/97-82-02 Radiation Protection Manager reporting to the Maintenance Manager is inconsistent with the Unit 3 Technical Specifications (Section 1.5).

IFl 423/97-82-03 HF/E personnel may lack organizational independence required by Technical Specification 6.2.3.3 for ISEG assessments (Section 1.5).

IFl 423/97-82-04 Corrective Actions Program lacked controls over combining CR to preserve issues and significance level (Section 2.1).

VIO 423/97-82-05 CR M3-97-3974 documented an audit finding that the ISEG/OE procedure, NOOP 3.04, was not reviewed by SORC as required by ,

Technical Specifications 6.2.3.1 and 6.5.2.6 (Section 2.3). l I

VIO 423/97-82-06 Failure to identify and correct the air binding of the boric acid transfer pumps rendered them inoperable (Section 2.3.1).

VIO 423/97-82-07 Inappropriate significance level assigned to CR's M3-97-4672 and M3-97-4346 (Section 2.3.3). ,

VIO 423/97-82-08 Failure to complete corrective actions for SW flow indicators 3SWP-F1-059 A, B and C (Section 3.1).

IFl 423/97-82-09 Quality lacking in the Setpoint Control Program (Section 3.2.b.6).

l VIO 423/97-82-10 Failure to follow the requirements of PI 29 and develop Design 8 asis l

Summaries for the Emergency Lighting System and CVCS (Section 3.2.b.7).

i VIO 423/97-82-10 Failure to follow the requirements of NGP 5.28 to maintain the Safety j Functional Requirements manual (Section 3.2.b.8).

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4 Closed IFl 50-423/96-06-17 4. UR! 50-423/95-81-01 4. Discussed VIO 50-423/96-05-12 4.1. _ - - - - _ - - _ - - _ - -

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, LIST OF ACRONYMS USED ACR adverse condition report (s)

AITTS action item Tracking and Trending System AR(s) action request (s)

ARP(s) alarm response procedure (s)

ALARA as low as reasonably achievable AOP(s) abnormal operating procedure (s)

ATL(s) audit team leader (s)

AWO(s) automated work order (s)

BAST (s) Boric Acid Storage Tank (S)

CAP corrective action plan CCP Component Cooling Water System CFR Code of Federal Regulations CMP configuration management plan CNO Chief Nuclear Officer CR(s) condition report (s)

CVCS Chemical & Volume Control System DBDP(s) design basis documentation package (s)

DCM Design Control Manual DCN(s) design change notice (s)

DCR design change record DRS Division of Reactor Safety ECOP Employee Concerns Oversight Panel ECP Employee Concerns Program EDG(s) emergency diesel generator (s)

EOP(s) emergency operation procedure (s)

EQ environmental qualification FCAT Fundamental Cause Assessment Team FME foreign material exclusion FSAR Final Safety Analysis Report GL Generic Letter l HELB high energy line break HF/E human f actors / engineering HVAC heating, ventilation and air-conditioning IFl inspector follow item INPO Institute of Nuclear Plant Operations IPE Individual Pisnt Evaluation ISEG Independent Safety Engineering Group JUMA Joint Utilities Management Assessment LER(s) licensee event report (s)

LOCA loss of coolant accident i M&TE measuring and test equipment j MCB main control board j MMOD minor modification j MRT Management Review Team i MSE maintenance support engineering evaluation MSR moisture separator reheaters NCAT Nuclear Committee Assessment Team

~ NCR(s) nonconformance report (s)

NGP(s) nuclear guidance procedure (s)

NNECO Northeast Nuclear Energy Company

.NORVP Nuclear Oversight Verification Plan

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l l [_ NOV(s) Notice of Violation l

NRR Nuclear Reactor Regulation NSAB nuclear safety assessment board NSIC' Nuclear Safety information Center NSSS nuclear steam system supply NUMARC Nuclear Management and Resources Council NUQAP - Northeast Utilities Quality Assurance Program l

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NUREG Nuclear Regulation NUSCO Northeast Utilities Service Company OCA Office of Congressional Affairs l OEDO Office of Executive Director for Operations

! OSTI Operational Safety Team inspection P&lD(s) Piping and instrument Drawing PAO Public Affairs Office l

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PDCR plant design change request PDDS Plant Design Data System PDR Public Document Room PIR plant incident report PLC post LOCA cooling PLS Westinghouse Precaution, Limitations and Setpoints PMMS production maintenance management system PORC plant operation review committee Pzr pressurizer QA quality assurance QAS Quality and Assessment Services RG Regulatory Guide SBO station blackout SCWE Safety Conscious Work Environment Program 3 SFR Safety Functional Requirements Manual SIL significant item list SORC site operations review committee SPDS Safety Parameter Display System SPO Special Projects Office SW service water-TS(s) technical specification (s)

UFSAR updated final safety analysis report URl(s) unresolved item (s)

USO(s) unresolved safety question (s)

.Vdc volts, direct-current Vac volts, alternating-current VIO violation VTM vendor technical manual WC work control WP& OM work planning & outage management ... _