ML18101B166
| ML18101B166 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/15/1995 |
| From: | Eliason L Public Service Enterprise Group |
| To: | Lieberman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML18101B165 | List: |
| References | |
| NUDOCS 9601160030 | |
| Download: ML18101B166 (50) | |
See also: IR 05000272/1995002
Text
Public Service Electric and Gas Company Leon R. Eliason Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1100
Chief Nuclear Officer & President
Nuclear Business Unit NOV 1 5 1995 * * LR-N95196
United States Nuclear Regulatory
Commission
Document Control Desk Washington, DC 20555 Attn: Mr. James Lieberman
Director -Off ice of Enforcement
Gentlemen:
RESPONSE TO NRC NOTICE OF VIOLATION
INSPECTION
REPORT NOS. 50-272/311/94-32, 50-272/311/95-02, 50-272/311/95-07
AND 50-272/311/95-10
SALEM GENERATING
STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 On October 16, 1995, the Nuclear Regulatory
Commission (NRC) issued a Notice of Violation (NOV) and proposed a $600,000civil
penalty for violations
identified
by the NRC during four inspections
that occurred between December 5, 1994 and June 23, 1995. The NRC issued to Public Service Electric & Gas (PSE&G) reports for these inspections
on March 30, April 7, May 24, and July 14, 1995. A predecisional
enforcement
conference
was held on July 28, 1995. PSE&G does not dispute the violations
cited in the October 16, 1995 NOV. Therefore, pursuant to 10CFR2.201, PSE&G submits its reply to the October 16, 1995 NOV. An electronic
transfer of funds payable to the Treasurer
of the United States in the amount of the proposed civil penalty will be made on November 15, 1995. As the NRC is aware, PSE&G management
realized that significant
steps were necessary
to reverse the P,erformance
decline at Salem. Therefore, on June 7, 1995, a decision was made to maintain Salem Unit Nos. 1 and 2 shutdown -until performance
improves to acceptable
levels. The self-imposed
shut down sent a significant
message to PSE&G employees.
PSE&G management
is 9601160030
951227 PDR ADOCK 05000272 G PDR
I . ** * * Document Control Desk LR-N95196 -2 -i\!OV-1 5 1995 serious about the changes necessary
for plant safety, personnel
performance, and process improvement.
PSE&G evaluated
the apparent violations
and broader concerns identified
in the four inspection
reports. Based on this evaluation, our July enforcement
conference
presentation
focused on three critical broad areas that had to be improved before acceptable
and long-lasting
changes at Salem could occur. These areas are: (1) establishment
of a culture that will facilitate
improvement, (2) improvement
of self-assessment
capabilities, and (3) ensuring timely and thorough problem assessment
and resolution.
These focus areas and their underlying
problems are a subset of concerns being addressed
in the Salem Restart Plan. The details of the Restart Plan will be formally submitted
on the
and discussed
with you during the public meeting presently
scheduled
for December 1995. In addition to our response contained
in Attachments
1 through 5, we provide below a discussion
of our progress in addressing
the three focus areas . Culture Change Improved personnel
and organizational
performance
is currently
and *will continue to be a focal point for the new management
team and is considered
essential
in establishing
the proper safety culture within the Nuclear Business Unit (NBU) . To aggressively
change the culture of the NBU, most of its top management
has been replaced.
This change signals the most important
factor that distinguishes*present
activities
from those of the past. One of the key characteristics
of the new managers is the ability to lead by example. Personnel
selected for this team have demonstrated
the necessary
leadership
capabilities
as well as the high standards
necessary
to develop a quality organization.
Most of the individuals
come from nuclear units which have had successful
performance
turn-arounds
and operate at an excellent
level. NBU management
has placed an emphasis on the development
and communication
of roles and responsibilities
to the organization, as well as establishing
expectations
for individual
performance.
The following
are examples of initiatives
which have been established
to drive the process of change. 95-4933
- * Document Control Desk LR-N95196 -3 -NOV 15 1995 First, several of the action plans developed
to support restart recognize
the need for improved definitions
of organizational
and individual
roles and responsibilities.
For example, a Conduct of Operations
document is now being finalized
which communicates
management
expectations
and, as importantly, establishes
the ethic of the Operations
organization.
Roles and responsibilities
for system engineers
have already been defined and communicated
to support the System Readiness
Review Process and system engineer improvement
initiatives.
The goal of these communications
is to establish
the necessary
standards
against which personnel
and organizational
performance
can be measured and held accountable.
Secondly, a Performance
Ranking process has been instituted
to assess individual
performance
in the following
behavioral
areas;
and Leadership, Initiative
and Results Achievement, Job Knowledge, Communication, and Adaptability
and Flexibility.
Individuals
will develop improvement
plans appropriate
to their overall standing.
This process is designed to identify and confront substandard
performance
that has gone undetected
or unchallenged
to date. In addition, personnel
who fail to make prescribed
improvements
will be held accountable, up to and including
discharge.
This ranking process represents
the first of four performance
review efforts to be conducted
within the NBU over the next 18 months. This focus on performance
is intended to re-emphasize
the responsibility
of managers and supervisors
to set and enforce proper
s.tandards
and revise substantially
the quality and productivity
of the workforce.
Finally, managers and supervisors
are being provided training to assist them in identifying, confronting
and correcting
performance
issues. The process being utilized has been implemented
successfully
at other nuclear plants, as well as non-nuclear
companies.
NBU management
has established
the expectation
that line managers and supervisors
attend this training and utilize this process. Two protocol groups have been established
to ensure the process is being ir.,plemented
uniformly
and consistently.
The Managers protocol group has recently developed
the course content and identified
significant
issues to be addressed.
The Executive
protocol group has evaluated
the course content and training to ensure that expectations
for this process have been satisfied.
In the longer term, the Managers protocol group will evaluate . * implementation
of the process to promote consistency
and make r I . 95.4933
- * * Document Control Desk LR-N95196 -4 -NOV 1 5 1995 appropriate
recommendations
on policy issues to the Executive
protocol group. These actions, effectively
implemented, are expected to improve individual
and organizational
performance
and will provide the infrastructure
for the proper safety culture within the NBU. As the impacts of these actions are measured, appropriate
changes in approach and method will be made to achieve the lasting and profound changes being targeted.
Self-Assessment
Improvement
The long-term
objective
for this focus area is to develop an organization
which instinctively
takes necessary
steps to improve
through effective
self-assessment
and timely corrective
action. A program defining expectations
for assessment
during routine operations
has been developed.
Each Salem department
has identified
specific representatives
to support this program. These representatives
have been trained on the program and its expectations.
To date, all but one Salem Station department
has performed
a self-assessment
using this program. The remaining
departmental
self-assessment
will be completed
in the near future. Issues identified
during these assessments
will be reviewed and incorporated
into the Salem Restart Plan, as appropriate.
A second program, which provides guidance on conducting
self-assessments
for readiness
to return to operation
following
refueling
outages, is being developed.
Salem personnel
are demonstrating
their willingness
to identify deficiencies
and to initiate actions necessary
for correction.
Indications
of this can be seen in the March 24, 1995, "Organizational
Effectiveness
Assessment
Report for Salem Nuclear Generating
Station," and our presentation
during the July 28, 1995 enforcement
conference.
This continues
to be shown by system walkdown results, backlog review, and most notably, the number of condition
reports being generated
on a daily basis. NBU management
has and will continue to monitor, and to the extent necessary
intervene, when self-assessment
related expectations
are not met . 95.4933
- * * Document Control Desk LR-N95196
Timely/Appropriate
Resolution -5 -NOV 1 5 1995 A consolidated
Corrective
Action Program (CAP) has been implemented
to communicate
NBU management
expectations
on timely problem identification
and resolution
and provides clear definition
of roles and responsibilities.
The CAP was designed using input from other utilities
which have effectively
managed program consolidations
as measured by improved program and station performance.
The consolidated
program includes a low threshold
for reporting
problems, provides aggressive
problem assessment/root
cause determination
expectations
and places management
in charge of root cause and corrective
action completion
times. Results to-date indicate that personnel
are not hesitant to raise issues through the process. The Director -Quality Assurance/Nuclear
Safety Review has oversight
responsibility
for the CAP. He has dedicated
resources, *_under the Manager -Corrective
Action and Quality Services, to fulfill that responsibility.
Measures have been established
to monitor the performance
of the corrective
action process. Recent data indicate overall improvement
in evaluation
completion
times and a reduction
in overdue corrective
actions. Station management
receives daily reports on overdue evaluations
-most of which have resulted from the volume of issues generated
by system walkdowns.
Accountability
for CAP implementation
rests with station line management.
As such, station managers review root cause evaluations
for completeness
and adequacy.
A Corrective
Action Review Board (CARB) has been established
at Salem and the General Manager -Salem Operations
is its chairman.
Completed
root cause assessments
for significant
issues are presented
to the CARB where the adequacy of the cause determination
and selected corrective
actions are evaluated:
A performance
measure has been established
which tracks the acceptance/rejection
rate for CARB presentations.
This indicator
is included in the monthly report to senior management . 95-4933
- * * Document Control Desk LR-N95196 -6 -NOV 15 1995 A new element, being incorporated
under the CAP improvement
area, is the Operational
Experience
Feedback (OEF) Program. This program is under review to identify needed improvements
in the processing
of internal and external OEF information.
This review includes a validation
of actions taken in response to past OEF items. Improvements
to the OEF process itself will include the establishment
of well defined roles and responsibilities, and standards
of performance
for implementing
organizations.
Performance
measures will also be established
to allow NBU management
to monitor program effectiveness
and assign accountability
if performance
standards
are not satisfied.
These changes are being made in order to better integrate
the OEF program into the operation
of the stations.
NBU management
recognizes
that, in addition to the changes already described, culture improvements
and self-assessment
capability
improvements
are essential
to anchoring
the CAP as an integral part of sustained
performance
improvement.
We will establish
and achieve appropriate
performance
standards
for the CAP at Salem prior to restart . Summary We agree with the NRC. that performance
within the NBU must improve. Our commitment
to maintain the Salem Units shutdown until required performance
improvements
are demonstrated, changes to the NBU management
team, and our aggressive
actions to strengthen
the safety culture within the NBU., illustrate
the fundamental
differences
between our present actions and those of the past. We will not restart the Salem Units until the hardware, important
processes
and programs, and organizational
and individual
performance
reach acceptable
levels. Changes will continue, as needed, to ensure that expectations
continue to be met after resumption
of power operation . 95-4933
I . ** * * Document Control Desk LR-N95196 -7 -NOV 1 5 1995 If you have any questions
regarding
this submittal, please do not hesitate to contact me. Sincerely, Attachments
95-4933
- * Document Control Desk LR-N95196 -8 -C /Mr. T. T. Martin, Administrator
-Region I U. S. Nuclear Regulatory
Commission
475 Allendale
Road King of Prussia, PA 19406 NOV 15 1995 Mr. L. N. Olshan, Licensing
Project Manager -Salem U. S. Nuclear Regulatory
Commission
One White Flint North 11555 Rockville
Pike Mail Stop 14E21 Rockville, MD 20852 Mr. C. Marschall
-Salem (S09) USNRC Senior Resident Inspector
Mr. K. Tosch, Manager, IV NJ Department
of Environmental
Protection
Division of Environmental
Quality Bureau of Nuclear Engineering
CN 415 Trenton, NJ 08625 95-4933
- * REF: LR-N95196
STATE OF NEW JERSEY SS. COUNTY OF SALEM L. Eliason, being duly sworn according
to law deposes and says: I am Chief Nuclear Officer & President
-Nuclear Business Unit of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced
letter, concerning
the Salem Generating
Station, Unit Nos. 1 and 2, are true to the best of my knowledge, information
and belief .
and
before me this /5-t.h day of /puvm/JJ.A_ . , 1995
/biGW£. Notary
Jersey My Commission
expires on KIMBERLY JO BROWN NOTf\RY PUBLIC OF NEW JERSEY My Commission
Expires April* 21, 1998
I I " ** * * Document Control Desk LR-N95196 -1 -ATTACHMENT
1 VIOLATION
I. 10 CFR Part 50, Appendix B, Criterion
XVI, Corrective
Action, requires, in part, that conditions
adverse to quality are promptly identified
and corrected; .and in the case of significant
conditions
adverse to quality, the cause of the condition
shall be documented, appropriately
reported to levels of management, and corrective
action taken to preclude repetition.
- A. Contrary to the above, a significant
condition
adverse to quality existed at the Salem Unit 2 facility from January 26, 1995, until June 7, 1995, in that the Licensee was aware that the No. 22 Residual Heat Removal (RHR) pump minimum recirculation
flow valve would not open on low RHR flow as required to prevent pump failure. Similarly, the Licensee was aware that the same significant
condition
adverse to.quality
existed at the facility from February 9, 1995, until June 7, 1995, for the No. 21 RHR pump minimum recirculation
flow valve. However, prior to June 7, 1995, the Licensee failed to determine
the cause of the valve failures or initiate corrective
measures.
(01013) This is a Severity Level III Violation (Supplement
1) Civil
$100,000 RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
- psE&G does not dispute the violation.
On January 26, 1995, and
9, 1995, different
operating
crews identified
failure of the automatic
open feature for the Residual Heat Removal (RHR) pumps minimum flow recirculation
valves 21RH29 and 22RH29. Both failures occurred as Salem Unit 2 was nearing completion
of its eighth refueling
outage (2R8) . Each failure was observed while the console *operator (a licensed Reactor Operator (RO)) was reducing RHR flow in preparation
to align RHR as an Emergency
Core Cooling System (ECCS) flowpath.
In both cases, the operating
crew initiated
an Action Request (AR) . Troubleshooting
for these valves was subsequently
scheduled
for August 2, 1995 and June 27, 1995, respectively.
Although Operations
personnel
recognized
that valve operability
was Mode-dependent, they did not establish
mode change constraints
when the failure of the automatic
open feature was recognized .
- * * Document Control Desk LR-N95196 -2 -Attachment
1 (cont'd) In June, 1995, following
Operations
department
identification
of 54 open work orders with potential
operability
concerns, these valves were targeted for immediate
operability
assessment.
Once valve operability
was questioned, the RHR system was operated to test and evaluate valve response.
Valve 21RH29 failed to operate and was declared inoperable.
When tested, valve 22RH29 opened on pump start. The Engineering
Analysis Group (EAG) was tasked with performing
a follow-up
operability
assessment.
The results of follow-up
engineering
evaluations
did not provide sufficient
basis to confirm 22 RHR loop operability.
As a result, with both RHR loops inoperable, at 18:27 hours on June 7, 1995, the operating
crew entered Technical
Specification
3.0.3 and commenced
shutdown of Salem Unit 2. At the time of the initial valve misoperation
events, an Operations
Standing Order and Operability
Determination (OD) Flowchart
were in place to guide Operations
personnel
in making Operability
Determinations.
Licensed operators
had received training on the use of the OD flowchart
during the 1994 fall training segment. Although the Standing Order and OD Flowchart
were available
on January 26, 1995, and February 9, 1995, the operating.crews
did not perform an Operability
Determination
when the operation
of the RH29 valves came into question . ROOT CAUSE ASSESSMENT
The RH29 valve control relays were tested and the most probable cause for
misoperation
was attributed
to failure of the Struthers-Dunn
low flow interlock
relay. PSE&G has determined
that the root cause of the failure to *identify
and correct this condition
adverse to quality was inadequate
management
commitment
to the Operability
Determination
process. This was demonstrated
by the following:
1. The implementation
of NRC Generic Letter (GL) 91-18 operating
philosophy
was not timely and effective
in improving
Operability
Determinations.
2. The implementation
of Operations
Department
procedures (Operability
Flowchart
and Operations
Department
Directive
SC. OP-DD. ZZ-OD02 (Q) (OD-2) , 11 Operability
Determinations")
to improve Operability
Determinations
was ineffective.
3. Less-than-adequate
safety culture within the Operations, Technical
Engineering, and Station Planning organizations, whi.ch was manifested
by a tolerance
for equipment
problems and insufficient
follow-through
to correct these problems .
- * * Document Control Desk LR-N95196 -3 -CORRECTIVE
STEPS THAT HAVE BEEN TAKEN Attachment
1 (cont'd) The Struthers-Dunn
valve control relays for valve 21RH29 were replaced.
The 22RH29 valve control relays passed in situ functional
testing and will be replaced prior to Unit restart. Salem Unit 2 was shutdown to comply with Technical
Specification
requirements.
To address the less-than-adequate
safety culture issues, PSE&G management
decided that Salem Units 1 and 2 will remain shutdown until performance
improves.
The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
OD-2, "Operability
Determinations" has been revised to provide better guidance and expectations
for performance
of Operability
Determinations.
Operator awareness
of NRC GL 91-18 is being reinforced
during Salem licensed operator training.
These actions assure that management
expectations
regarding
roles and responsibilities
in the Operability
Determination
process are clearly understood
and consistently
applied. As an interim measure, the Operations
department
reviews
Operability
Determinations (OD's) periodically
to ensure that actions and contingencies
are progressing
and/or completed.
The review process is directed by Operations
procedure
OD40 {Q), "Shift Routines." To assess the effectiveness
of the OD process, the Safety Review Group (SRG) is, on an interim basis, independently
evaluating
the OD's and providing
feedback to Operations
management.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
The Operability
Determination
process, including
the OD-2 procedure, is being further enhanced to: 1) improve the Engineering
and Operations
departmental
interface;
2) ensure consistency
between OD-2 and NC.NA-AP.ZZ-0006(Q) (NAP-6) "Corrective
Action Program";
and 3) ensure tracking of Operability
Determination
status. These improvements
to the process will be completed
by March 1, 1996 . f
- * * Document Control Desk LR-N95196 -4 -DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED Attachment
1 (cont 1 d) PSE&G has identified
and corrected
the cause of the valve failures.
PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -ATTACHMENT
2 VIOLATION
B. Contrary to the above, a significant
condition
adverse* to quality existed at the Salem Unit 1 facility from December 12, 1994, until May 16, 1995, in that the No. 12 safety related switchgear
ventilation
supply fan failed on December 12, 1994, and the Licensee did not initiate resolution
of the condition
or effect any corrective
measures to resolve the condition
promptly. ( 02013) This is a Severity Level III Violation (Supplement
1) . Civil Penalty -$100,000 RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
In Deceinber, 1994, the No. 12 Switchgear
Area Ventilation
System (SPAVS) supply fan tripped on overload protection.
Further investigation
revealed that the fan motor bearings had failed. Repair of the fan motor bearings necessitated
that the fan motor assembly be removed from the system. A Temporary
Modification (T-Mod) was required to maintain system/plenum
integrity
with the fan motor assembly removed. As a result of poor planning and lack of communication, corrective
actions had not been taken to repair the No. 12 SPAVS supply fan when the No. 13 SPAVS supply failed on May 12, 1995. At the time of these failures, no spare supply fan motors were available.
Troubleshooting
reve.aled
that the second fan motor had developed
an internal short to ground. In accordance
with the Salem Updated Final Safety Analysis Report (UFSAR) , normal system operation
requires two of the three 50% capacity SPAVS supply fans to be in service, with the third fan available
in a standby mode to accommodate
failures.
With the failure of No. 12 SPAVS supply fan motor in December, 1994, .station personnel
failed to recognize
that SPAVS was operating
outside the UFSAR assumptions.
On May 12, 1995, two of the three supply fans became
and System Engineering
personnel
were unable to clearly establish
the system's ability to fulfill its intended safety function.
A shutdown of Salem Unit 1 was initiated
on May 16, 1995 .
- * * Document Control Desk LR-N95196
ROOT CAUSE ASSESSMENT -2 -Attachment
2 (cont'd) PSE&G has determined
that the root cause of this event was ineffective
corrective
action. Involved personnel
failed to recognize
the significance
of losing redundant, important
to safety components.
Due to a less-than
adequate safety culture, prompt corrective
actions, consistent
with the safety significance
of the equipment, were not initiated
as evidenced
by: 1. Failure to repair the first failed SPAVS supply fan motor in a timely manner. 2. Lack of communication
in the System Engineering
organization.
3. Failure to complete the work planning for repair by the issuance of a T-Mod which was not accomplished
prior to the second SPAVS supply fan motor failure. The
Action Program (CAP), in effect at that time, lacked sufficiently
low thresholds
to ensure that conditions
adverse to quality would be identified
and resolved in a timely manner. That same program did not provide clear guidance on the need to perform nor the required content of assessments
to support continued
assurance
of equipment
operability . The following
contributing
factors were also identified:
1. Adequate Preventative
Maintenance (PM) program tasks were not established
for these fan motors. Opportunities
to establish
appropriate
PM's were missed due to lack of follow-through
with regard to industry experience
notifications
and a previous SPAVS fan motor failure. 2. Lack of clear understanding
by Operations
and Engineering
personnel
of the SPAVS design basis. 3. Operations
did not have a tracking system to assure that inoperable
Technical
Specification
systems or support systems would be corrected
in a timely manner. CORRECTIVE
STEPS THAT HAVE BEEN TAKEN On May 16, 1995, Salem Unit 1 was shutdown to comply with Technical
Specification
requirements
when reasonable
assurance
of system operability
could not be established .
- * * Document Control Desk LR-N95196 -3 -Attachment
2 (cont'd) The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
Preventive
Maintenance
Change Requests (PMCR's) were generated
to create new PM Recurring
Tasks to replace the SPAVS fan motor bearings on a regular basis. All three Salem Unit 1 SPAVS supply fans were inspected
and the fan motors replaced.
- operations
Department
procedure
SC.OP-DD.ZZ-ODlO(Q} "Removal and Return of Nuclear Safety Equipment" has been issued. This procedure
provides guidelines
for removal and return to service of all Technical
Specification
related equipment.
OD-2, "Operability
Determinations" has been revised to provide better guidance and expectations
for performance
of Operability
Determinations.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
A_ll SPAVS supply fans on Salem Unit 2 will be inspected
and the fan motor bearings will be replaced, on an as-needed
basis, prior to unit restart . Process improvements
for the Operating
Experience
Feedback Program (OEF) are presently
under evaluation.
This activity is being managed under the Corrective
Action Program element of the Salem Restart Plan. The Technical
Specification
Action Tracking procedure
has been revised to require the NSS to verify and initial for completed
Technical
Specification
actions and allow for tracking of potential
Technical
Specification
entries. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The No. 12 and 13 SPAVS supply fans were repaired.
PSE&G
have achieve.d
compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
--
I . ** * ** Document Control Desk LR-N95196 -1 -ATTACHMENT
3 VIOLATION
C. The Licensee was informed by Westinghouse
on March 15, 1993, of a significant
condition
adverse to quality involving
nonconservatisms
in the setpoint methodology
for the Pressurizer
Overpressure
Protection
System (POPS) for low temperature
overpressure*
conditions.
1. Contrary to Criterion
XVI, the Licensee took nine months of analysis, from March 1993 to December 1993, to conclude that the corrected
peak transient
pressure would exceed pressure/temperature (P/T) limits as described
in each unit's technical
specifications
limits. After completing
the analysis, from December 30, 1993, and continuing
for approximately
one month, the Licensee dispositioned
the matter of the nonconservatism
in the setpoint methodology
for the POPS by 1) administratively
limiting RCS operation
to two reactor coolant pumps when the RCS was less than 200° F and 2) increasing
each unit's P/T limit by 10%; the latter corrective
action was inadequate
because it utilized as a basis an unauthorized
ASME Code Case (N-514), which the Licensee was aware was not acceptable
pursuant to 10 CFR 50. 55 (a) . ( 03 013) This is a Severity Level III Violation (Supplement
1) Civil Penalty -. $100,000 2. Contrary to Criterion
XVI, in January 1994, following
the Licensee recognizing
the unacceptability
of using unauthorized
Code Case N-514 as a corrective
action to disposition
the POPS setpoint methodology, the Licensee elected to implement
corrective
action by taking credit for the relief capacity provided by RHR system suction relief valve RH3 to augment POPS relief capacity .
I . ** * * Document Control Desk LR-N95196 -2 -Attachment
3 (cont'd) However, as the Salem FSAR (Section 7.6.3.2) describes
the POPS system to include two Power Operated Relief Valves (PORVs) and does not describe Valve RH3, this corrective
action was inadequate
because an evaluation
was not performed
to determine
the acceptability
of the use of Valve RH3 as part of the POPS system. In addition, the Licensee failed to identify that on the receipt of a safety injection (SI) signal, a previously
operating
positive displacement
charging pump's discharge, combined with the discharge
from the high head safety injection
pump that starts on receipt of the SI signal, could have injected water mass into the RCS at a rate that could have prevented
POPS from performing
its function.
(04013) This is a Severity Level III Violation (Supplement
I) Civil Penalty -$100,000 RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
On March 15, 1993, Public Service Electric & Gas (PSE&G) was advised by Westinghouse
of a generic issue involving
a conservative
setpoint calculation
in the analysis of the Pressurizer
Overpressure
Protection
System (POPS) . The Low Temperature
Overpressure
Protection
System (LTOPS) protects the Reactor Pressure Vessel (RPV) against pressurized
thermal shock events as required to comply with 10CFR50 Appendix G criteria ("Fracture
Toughness
Requirements").
PSE&G requested
to perform a Salem plant-specific
analysis for the cases of one, two or four reactor coolant pumps running. On September
29, 1993, PSE&G received the plant*-specif
ic Westinghouse
results and had evidence that a non-conservative
setpoint (375 psig) could lead to violating
the Technical
Specifications
and Appendix G pressure/temperature
limits. Over a period of three months (September
to December, 1993), Nuclear Engineering
personnel
performed
calculations
to address this concern .
I . ** Document Control Desk LR-N95196 -3 -Attachment
3 (cont'd) On December 30, 1993, the Nuclear Engineering
department
issued an evaluation (MEC-93-917)
which restricted
operations
in Mode 5 to two reactor coolant pumps. The recommended
restrictions
were implemented
via revisions
to the plant's Integrated
Operating
Procedures (IOP's). Nuclear Engineering
personnel
improperly
took credit for American Society of Mechanical
Engineers (ASME) Code Case N-514 (which had not yet received NRC approval)
as part of dispositioning
this issue. Despite the involvement
of multiple departments
during this evaluation
process, numerous opportunities
to recognize
the reportability
requirements
for this issue were missed and, as a consequence, the condition
was not reported to the NRC. On May 26, 1994, another evaluation (MEC-94-630)
was issued whi.ch further restricted
the number of operating
Reactor Coolant pumps from two to one pump in Mode S. The new calculated
values showed that Salem Unit 2 pressure did not exceed specified
limits. However, it was recognized
that Salem Unit 1 could exceed its pressure limit during a mass addition transient
below 2oo*degrees
F. Involved personnel
failed to evaluate the calculated
deviation
from the specified
limit against reportability
requirements.
Likewise, there was a failure to recognize
the need to establish
justification
for continued
operation
while this condition
existed and the need to report * that justification
to the NRC. * On June 13, 1994, Nuclear Engineering
issued calculation
S-C-RC-MDC-1358.
This calculation
inappropriately
took credit for use of a relief valve (RH3) in the Residual ijeat Removal (RHR) system. On November 17, 1994, it was determined
that Salem Unit 1 could operate outside of the design/licensing
basis for the POPS analysis if the following
conditions
existed: 1) a Safety Injection (SI) signal was initiated;
2) Reactor Coolant System (RCS) temperature
was below 200 degrees F; 3) a Reactor Coolant pump was in service; 4) a Positive Displacement
Charging Pump was in service; and 5) power remained available
to a maximum of one Centrifugal
Charging Pump. This discovery
resulted in the issuance of Licensee Event Report (LER) 272/94-017 .
I . ** * * Document Control Desk LR-N95196 -4 -Attachment
3 (cont'd) On February 7, 1995, the NRC approved PSE&G's use of ASME Code Case N-514. At that time, appropriate
Safety Evaluations
were performed
for the resultant
changes to the Updated Final Safety Analysis Report (UFSAR) . Implementation
of the ASME Code Case provided additional
margin (10%) and higher pressure/temperature
limits for POPS during the LTOP conditions
and re-established
plant operation
within its design and licensing
bases. In April, 1995, PSE&G issued Incident Reports to identify and evaluate the organization's
inappropriate
actions and their causal factors. ROOT CAUSE ASSESSMENT
PSE&G has determined
that the root causes of this event were: 1. 2. 3. Lack of understanding
of the regulatory
significance
and reportability
implications
of the Westinghouse
analysis results. Specifically, the organization
became too focused on the technical
resolution
aspects of the issue without adequate consideration
of regulatory
requirements.
Lack of supervisor/management
sensitivity
to the need to impJement
existing procedures
and processes
which require timely entry of issues into the Corrective
Action Program (CAP) . Monitoring
of the Corrective
Action process by
was insufficient.
Inadequate
training of engineering
personnel
on the use of ASME Code Cases, requirements
of lOCFRS0.59
and requirements
for regulatory
reporting.
CORRECTIVE
STEPS THAT HAVE BEEN TAKEN The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment .
- * * Document Control Desk LR-N95196 -5 -Attachment
3 (cont'd) NBU Management
has re-emphasized
the expectation
that supervisory
personnel
must assess issues objectively.
Specifically, supervisory
personnel
must maintain their oversight
role. The Manager -Nuclear Engineering
Design (NED) has verbally reinforced
this expectation
to the engineering
design organization.
The Nuclear Engineering
Design organization
was surveyed relative to any past reliance on unapproved
ASME Code Cases. Based on this survey, no other instances
of unapproved
ASME Code. Case use were identified.
Personnel
involved in this occurrence
have received appropriate
reinforcement
on procedure
compliance, their responsibility
for compliance
with regulatory
requirements, and problem reporting.
Management
has re-emphasized
by internal memorandum
and follow-up
review with engineering
personnel
that the potential
impact on the UFSAR must be considered
whenever design basis calculations, evaluations
or assumptions
are revised. Departmental
procedures
provide clear guidance on these requirements.
The expectation
- for procedural
adherence
was also reinforced.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
Engineering
Design and Licensing
&_Regulation
management
will reinforce
expectations
for organizational
interface
to their personnel.
This will be completed
by March 15, 1996. Lessons learned from this issue will be disseminated
to Engineering
Support personnel
during 4th quarter Operating
Experience
Feedback (OEF) training.
This will be completed
by January 15, 1996. Process improvements
for the Operating
Experience
Feedback Program (OEF) are presently
under evaluation.
- This activity is being managed under the Corrective
Action Program element of the Salem Restart Plan. Specific training on the ASME Code and NRC restrictions
on its use will be provided to appropriate
engineering
support personnel.
This will be completed
by January 31, 1996 .
- * * Document Control Desk LR-N95196 -6 -Attachment
3 (cont'd) The Engineering
Qualification
training program is being revised to assure that job qualifications
are consistent
with job requirements
and that Engineering
personnel
are trained consistently.
Required personnel
training in Code Job Packages will be incorporated
into the Engineering
Qualification
Guide. This* training will include ASME Code Cases, NC.NA-AP.ZZ-002B(Q) "Code Job Packages" procedure
requirements
and regulatory
reporting
requirements.
The revised Engineering
Qualification
Guides will be completed
by January 31, 1996. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The request to use ASME Code Case N-514 at Salem station was approved by the NRC. PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
I Document Control Desk .* LR-N95196 -1 -* * ATTACHMENT 4 -lST EXAMPLE VIOLATION
D. Contrary to the above, on several occasions, conditions
adverse to quality existed, but were not identified
and promptly corrected, as evidenced
by the following
examples:
1. On June 7, 1994, the Licensee identified
that material management
documentation
for limit switches related to the reactor head vent valves, improperly
classified
the components
as non-safety
related. A nuclear design discrepancy
evaluation
form (DEF) identified
that a switch short circuit could render two head vent valves inoperable
since the components
were powered from the same common circuit. Notwithstanding, the DEF did not identify any concern relative to operability
or safety. In February 1995, the Licensee determined
that non-safety
related limit switches were actually installed
in reactor head vent valves 1RC41 and 1RC43 at Salem Unit 1. Subsequently, the Licensee failed to* perform and document an engineering
evaluation
to demonstrate
the acceptability
of continued
Salem Unit 1 operation
with non-safety-related
parts installed
in a safety-related
application.
RESPONSE -DESCRIPTION.OF
CIRCUMSTANCES
PSE&G does not dispute the violation.
On June 7, 1994, a Discrepancy
Evaluation (DEF) was written to resolve an apparent conflict in safety classification
between the Reactor Head vent valves and their corresponding
position indicating
limit switches for Salem Units 1 and 2. On March 3, 1995, it was determined
that
related limit switches were installed
in two Salem Unit 1 Reactor Head vent valves. Investigation
into this occurrence
indicates
that, in April, 1992, an opportunity
to resolve the noted discrepancy
was missed when a different
DEF on the same subject was dispositioned.
The identified
corrective
actions in that DEF were not carried through to completion .
- * * Document Control Desk LR-N95196
ROOT CAUSE ASSESSMENT -2 -Attachment
4 (cont'd) PSE&G has determined
that the root cause of this occurrence
was the erroneous
classification
of the Reactor Head vent valve limit switches as non-safety
related. Due to personnel
error, these switches were incorrectly
assigned a non-safety
related purchase class during a spare part Folio Classification
initiative
in 1986. This error in classification
initiated
a sequence of events which resulted in the installation
of non-safety
related limit switches in an application
originally
designed to use safety related components.
The root cause of the failure to resolve this condition
adverse to quality in a timely manner is attributed
to an inadequate
Corrective
Action Program (CAP). The CAP, in effect at that time, lacked sufficiently
low thresholds
to ensure that conditiqns
adverse to quality would be identified
and resolved in a timely manner. That same program lacked centraliz'ed
oversight
of the various mechanisms
to identify and resolve discrepancies.
CORRECTIVE
STEPS THAT HAVE BEEN TAKEN In March, 1995, Nuclear Engineering
Design issued an assessment
to resolve the outstanding
DEF. This assessment
concluded
that the non-qualified
switches did not affect the operability
of the Reactor Head vent valves. * The following
changes were made in the Nuclear Procurement
and Material Management (NP&MM) system: The Purchase Class 4 (PC4) Limit Switch Folio parts were put 11 0n Hold" and were re-classified
as 11 obsolete.11 A New Purchase Class 1 Limit Switch Folio was created. The Reactor Head vent valve limit switch component
ID's were removed from the computerized
Managed Maintenance
Information
System (MMIS). Separate component
ID's were determined
to be unnecessary
as the Bill of Materials (BOM) for the valves contains the Folio information
for the limit switches .
- * * Document Control Desk LR-N95196 -3 -Attachment
4 (cont'd) The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
Non-safety
related limit switches in the Reactor Head vent valves will be replaced prior to restart of Salem Unit 1. Outstanding
DEF's are being reviewed for impact on plant systems, including
operability
issues. This will be completed
prior to restart of Salem Units 1 and 2. PSE&G is currently
conducting
a review of the MMIS database to determine
if there have been other occurrences
of safety related components
being purchased
as non-safety
related. The scope of _this review will include components
acquired under purchase class "PC4" (non-safety
related).
Any additional
occurrence(s)
of safety related parts in safety related applications, discovered
during this review, will be dispositioned
under the current CAP
which include documentation
of Operability
Determination
and evaluation
for reportability, when appropriate . . This review will be completed
prior to restart of either Salem Unit 1 or 2. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The Engineering
department
has dispositioned
the outstanding
DEF. PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
I . ** * * Document Control Desk LR-N95196
VIOLATION
ATTACHMENT 4 -2ND EXAMPLE 2. On February 24, 1995, Unit No. 1 operators
placed control of a PORV in the manual mode, rendering
it inoperable, and failed to adhere to the Technical
Specification
3.4.3 action statement
which required operators
to close the block valve within one hour. A shift supervisor
discovered
that the PORV had been erroneously
placed in the manual mode and corrected
it on February 25, 1995, about 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> later. RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
On February 24, 1995, Salem Unit 1 was in the process of raising Reactor.Coolant
System (RCS) pressure using Integrated
Operating
Procedure
2 (IOP-2). To support a controller
inspection, the Pressurizer
pressure master controller
was removed and pressure control was placed in manual. This action rendered Operated Relief Valve (PORV) 1PR2 inoperable
and required closing of PORV block valve 1PR7. The operator did not close valve 1PR7 and the oversight
went unnoticed
for approximately
22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />. Although a pre-job brief was performed
prior to this evolution, it did not cover all TS required actions. Specifically, the briefing did not discuss closing valve 1PR7. The Nuclear Control Operator (NCO) and the Nuclear Shift Supervisor (NSS) failed to conduct adequate self-checking.
The NSS failed to maintain the proper supervisory
overview to insure that the Technical
Specification
action was completed.
ROOT CAUSE ASSESSMENT
The root cause of this event has been attributed
to personnel
error on the part of the 'supervisor (NSS) and the control operator (NCO) . A contributing
cause to this event was inadequate
guidance in the Technical
Specification
Action Tracking Log. This log did not prompt operators
to verify that TSAS are completed
when the action statement
is entered .
- * * Document Control Desk LR-N95196 -2 -CORRECTIVE
STEPS THAT HAVE BEEN TAKEN Attachment
4-2ND example (cont'd) Appropriate
disciplinary
actions were taken for the individuals
involved.
The NSS primary work location has been moved into the respective
Control Room area as of March 3, 1995, to improve oversight
and management
of control room activities.
The Technical
Specification
Action Tracking procedure
has been revised to require the NSS to verify and initial for completed
Technical
Specification
actions and allow for tracking of potential
Technical
Specification
entries. An Information
Directive 95-017 and two separate shift briefings
were completed
for each of the Operations
crews. The Operations
Department
has re-emphasized
the use of checking techniques, peer verification
and expecta.tions
for NSS oversight.
Operations
management
re-emphasized
the conditions
under which the PORV's should be declared inoperable
during Licensed Operator Requalification (LOR) training in segment 4, 1995. Understanding
of Technical
Specification
actions by Operations
personnel
were verified through LOR examinations.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
The Operations
Department
is developing
an Operations
Standards
document which will reference
appropriate
procedure
guidance for conducting
pre-job briefings.
The Operations
Standards
document will be implemented
by November 21, 1995. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The PORV block valve was closed to comply with TS requirements.
PSE&G will have achieved compliance
with 10CFR50 Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -Attachment
4-3RD example (cont'd) VIOLATION
ATTACHMENT 4 -3RD EXAMPLE 3. On July 6, 1994, safety-related
reactor head vent valve 2RC40 failed to operate (stroke open) during testing while Unit No. 2 was in cold shutdown.
Subsequently, the valve was returned to normal service on July 10, 1994, without any review or assessment
in accordance
with established
procedures;
that is, the Licensee failed to process this occurrence
in accordance
with the applicable "Work Control Process" procedure.
Consequently, this failure of a safety related component
was never documented
and formally assessed relative to preventive
maintenance, operability, actions to prevent recurrence, or generic implications.
RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
On July 6, 1994, the 2RC40 valve failed its post-maintenance
testing due to indications
of reduced flow and dual position indication
problems.
Subsequent
investigation, including
consultation
with the vendor, indicated
that the most probable cause of the valve failing to stroke open was due to boric acid solidification
around the pilot plug. The boron solidification
was suspected
to be the result of valve seat leakage. The Maintenance
Engineer recommended
backflushing
of the valve with demineralized
water and increasing
the Reactor Coolant System (RCS) temperature
to 180 °F to dissolve the boron. This resulted in proper valve operation
and supported
the original root cause supposition.
Therefore, it was concluded
that a temporary
condition
could develop at low RCS temperatures
and pressures
that could result in boric acid binding of the valve. On July 8, 1994, the valve was placed back into service with a recommendation
to evaluate the need for additional
valve preventive
maintenance .
- * * Document Control Desk LR-N95196 -2 -Attachment
4-3RD example (cont'd) In December, 1994, the Salem Unit 2 head vent valves were replaced as a result of excessive
seat leakage. Similar conditions
had been previously
observed on the Salem Unit 1 valves 1RC40 and 1RC42 and prompted their replacement
in May, 1994. In May, 1995, the vendor disassembled
and inspected
valve 2RC40, which had been removed in December, 1994, to identify any material condition
that could have caused the valve's failure to open. The test results on valve binding were inconclusive
but indicated
that the reported leaking of the valve could be attributed
to steam cutting between the valve and the pilot valve disc due to normal wear. On April 5, 1995, an Incident Report was initiated
and a root cause analysis undertaken
which arrived at much the same conclusions
as that of the vendor. The root cause of the failure of valve 2RC40 to stroke was indeterminate.
The type of degradation
experienced
by the head vent valve would not have alone prevented
it from stroking.
Degradation
of the valve internals, however, was identified
as a causal factor in both the valve leakage and failure to stroke and was attributed
to a lack of preventive
maintenance.
PSE&G has determined
that this failure mode is applicable
only to the Reactor Head vent valves. In April, 1995, a re-analysis
of the Preventive
Maintenance*
requirements
for these valve internals
was completed
and a 54-month inspection
was recommended.
ROOT CAUSE ASSESSMENT
The final root cause for the failure of valve 2RC40 to open was inconclusive.
Probable causal factors include: 1. Lack of preventive
maintenance
on valve internal components.
2. Accumulation
of boric acid precipitate
on valve pilot plug. The root causes for the failure to identify and correct this condition
adverse to quality were: 1. An inadequate
Corrective
Action Program* (CAP). The CAP, in effect at that time, failed to establish
sufficiently
low reporting
thresholds
to ensure that conditions
adverse to quality would be identified
and resolved in a timely manner. 2. Management
failure to establish
and enforce high expectations
for equipment
and personnel
performance .
- * * Document Control Desk LR-N95196 -3 -CORRECTIVE
STEPS THAT HAVE BEEN TAKEN Attachment
4-3RD example (cont'd) The Reactor Head vent valves 1RC40/42 and 2RC40/41/42/43
have been replaced in May, 1994, and December, 1994, respectively.
Valves 1RC41 and 1RC43 are being replaced during the current outage. The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
Appropriate
Operations
Department
procedures
have been revised. These revisions
include guidance to preclude boric acid accumulation
in the valve body. An Action Request to identify any solenoid operated valves other than the reactor head vent valves that serve as a Reactor Coolant System (RCS) pressure boundary and could potentially
be subject to the same or a similar failure mode, such as boric acid binding to*seat leakage, has been completed.
PSE&G has determined
that this failure mode is applicable
only to the Reactor Head vent valves. NC.NA-BP.ZZ-0002(Z), "Root Cause Analysis Guidelines," has been developed
to provide additional
information
and guidance in the use of various root cause analysis techniques
which have been proven effective
in resolving
both human and equipment
performance
problems.
Within the Salem Maintenance
Department, PSE&G has established
dedicated
resources
to conduct required root cause analyses, develop _and recommend
appropriate
corrective
actions, and assure their proper implementation
and overall effectiveness
through followup assessments.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
New PM Recurring
Tasks (RT's) have been initiated
to implement
a 54-month PM to open and inspect the Reactor.Head
vent valve internals
and to repair as needed. A new Maintenance
Department
procedure
has been issued to provide guidance on the disassembly, inspection
and refurbishment
of the Reactor Head vent valves. These corrective
actions will be completed
prior to restart of the affected unit .
- * ** Document Control Desk LR-N95196 -4 -Attachment
4-3RD example (cont'd) DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED This condition
was documented
and a root cause analysis was completed.
PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -VIOLATION
ATTACHMENT 4 -4TH & STH EXAMPLES 4. An oil sample laboratory
report, dated August 4, 1994, recommended
resampling
and changing the oil on the No. 21 high-head
safety injection
pump based upon a ten-fold increase in wear particle concentration.
An oil analysis, dated November 28, 1994, identified
high wear particle concentration
in the No. 22 high-head
safety injection
pump speed increaser
oil. In both these cases, the system engineer, though aware of the findings of the lab reports, did not initiate any follow-up
evaluation
or corrective
measure, nor establish
a bases for operability
or reliability
in view of the apparent degraded condition
of the equipment.
The degraded nature of the equipment
was not entered into the Equipment
Malfunction
Identification
System (EMIS) until March 20, 1995. 5. A lab report, dated October 6, 1994, recommended
resampling
the No. 23 Auxiliary
Feedwater (AFW) turbine lube oil due to a detectable
amount of water contamination
and an increase in wear particle concentration.
However, the degraded nature of the equipment
was not entered into the EMIS until March 27, 1995, and the system engineer did not initiate review, and evaluation, or establish
any basis for equipment
operability
or reliability.
RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
PSE&G acknowledges
the issues identified
in this violation
were not addressed
in a timely fashion. Documentation
of equipment
status was deficient
and inadequately
maintained .
- * * Document Control Desk LR-N95196 -2 -Attachment
4-4TH & STH (cont'd) ROOT CAUSE ASSESSMENT
PSE&G attributes
the root cause of these occurrences
to: 1. Management's
failure to enforce expectations
regarding
individual's
responsibilities
for the Performance
Monitoring
program. 2. The lengthy turnaround .time for laboratory
analyses (including
radioactive
material handling)
challenged
the ability of the System Engineer to make timely decisions.
The root cause for the failure to identify and correct these conditions
adverse to quality is: 1. An inadequate
Corrective
Action Program (CAP). The CAP, in effect at that time, lacked sufficiently
low thresholds
to ensure that conditions
adverse to quality would be resolved in a timely manner. That same program did not provide clear guidance on the need to perform nor the required content of assessments
to support continued
assurance
of equipment
operability.
CORRECTIVE
STEPS THAT HAVE BEEN TAKEN The 23 AFW Pump was declared inoperable.
This action was completed
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of when PSE&G received notification
from the laboratory-
that the follow-up
oil sample had been confirmed
to be the wrong grade for the component.
The Corrective
Program (CAP) has been revised as described
in the cover letter to this Attachment.
Roles and responsibilit.ies
within System Engineering
have been defined and communicated
as described
in the cover letter to this Attachment.
Within System Engineering, a component
reliability
group was established
to provide improved focus on equipment
performance
and reliability
issues. The Manager -Component
Reliability
will define and communicate
roles and responsibilities
for tracking
I . ** * * Document Control Desk LR-N95196 -3 -Attachment
4-4TH & STH (cont'd) and trending of performance
monitoring
data. This will be completed
by January 15, 1996. Lube oil abnormalities
from this occurrence
have been documented
by an Abnormal Condition
Report to the System Manager from the Lube Oil Analysis Program Manager. This process will remain in place to document future reports of abnormal indications.
PSE&G has contracted
with a lube oil analysis laboratory
capable of handling radioactively-contaminated
lube oil samples. The laboratory's
ability to handle contaminated
material will reduce the time from sample collection
to condition
determination
by reducing the required count time per sample. CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
The Lubricating
Oil Program is being assessed to identify recommendations
on a comprehensive
lube oil program. The program recommendations
are due by the end of the fourth quarter, 1995. These
will be evaluated
and an Implementation
Plan for approved recommendations
will be established
by the end of first quarter, 1996. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The abnormal Lube oil conditions
were documented, reviewed and evaluated
for operability
impact. PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related -processes
have been
effective
at identifying
and
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -VIOLATION
ATTACHMENT 4 -6TH EXAMPLE 6. LER 95-05 identified
seven instances, between May 8, 1990 and January 14, 1995, of Pressurizer
safety valves (PSVS) being beyond the 1% tolerance
required by TS 4.0.5 for Unit 1. Four instances
were identified
between November 14, 1994, and January 14, 1995, which involved 2 of the 3 installed
PSVS. In all instances, the vendor notified the appropriate
system engineer by telephone
and written follow-up
reports. However, the responsible
system engineer never initiated
an Incident Report. Consequently, root cause, operability, and reportability
actions were not accomplished.
-DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
Beginning
in May 8, 1990, eight (8) occurrences (total for both Salem Units) of Pressurizer
code safety valves (PSV's) exceeding
the 2485 psig +/-1% lift set pressure were identified.
Seven of those instances
were cited within the Notice of Violation.
An eighth occurrence
was self-identified
and reported to the NRC via LER Supplement
272/95-05-01, dated October 31, 1995. These occurrences
were identified
during testing required by Technical
Specification (TS) 4.0.5. Failure to report these anomalies
resulted from personnel
error in that Incident Reports (IR's) were not written in accordance
with NC.NA-AP.ZZ-0006(Q)
procedure
requirements.
ROOT CAUSE ASSESSMENT
The causes of the lift setpoint variances
are a combination
of variability
due to individual
valve performance
characteristics
and random test variations
which are common for these valves. The specific causes for Salem Station's
variation
are: 1. Minor test loop instrument
error. 2. Valve design limitations.
3. Applied loads from the discharge
piping .
' . ** Document Control Desk LR-N95196 -2 -Attachment 4 -6th example (cont'd) PSE&G has determined
that the programmatic
root cause of this violation
was management's
failure to clearly and adequately
communicate
expectations
regarding
when an IR was required.
Specifically, the System Engineers
did not recognize
the requirement
to initiate IR's for these lift setpoint anomalies, in accordance
with Nuclear Administrative
Procedure
NC.NA-AP.ZZ-
0006 (Q), "Corrective
Action Program" in effect at the time. They also failed to recognize
the reportability
implications
for the out-of-tolerance
valve performance
data. Consequently, the testing anomalies
were not reviewed against the lOCFRS0.73 "Licensee
Event Report" (LER) reporting
criteria and required LER reporting
did not occur. CORRECTIVE
STEPS THAT HAVE BEEN TAKEN PSE&G has performed
engineering
evaluations
as part of the Fuel Upgrade Margin Recovery program. The thermal-hydraulic
analysis indicates
that PSV lift setpoint variances
up to 3% are acceptaple.
The structural
analysis is more limiting, indicating
that variances
of +2.2 to -3.0% are acceptable.
Analyses within the Fuel Upgrade Margin Recovery program are continuing.
The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
,_., Lessons learned from this violation
were incorporated
into the * third quarter Operating
Experience
Feedback (OEF) training for Engineering
Support personnel.
Appropriate
discipline
was taken with personnel
involved in the failure to initiate IR's. Applied loading on the PSV's from the discharge
piping has been* reduced. CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
A single point of contact within the PSE&G organization
will be established
to ensure coordination
of activities
associated
with PSV testing. This will be completed
by March 29, 1996 .
- * * Document Control Desk LR-N95196 -3 -Attachment 4 -6th example (cont'd) DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The lift setpoint variance conditions
were documented, reviewed and assessed to demonstrate
acceptability.
PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved . l I
I . * * Document Control Desk LR-N95196 -l -VIOLATION
ATTACHMENT 4 -7TH EXAMPLE 7. On March 6, 1995, May 3, 1995, and May 8, 1995, the Salem Unit l staff failed to determine
the cause, correct, or prevent recurrence
of failure of the Containment
100 foot elevation
personnel
airlock to pass its local leak rate test. RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
On March 6, 1995, the Salem Unit l Containment
personnel
airlock on the 100 foot elevation
failed its local leak rate test (LLRT) A work request was initiated
for the Maintenance
Department
to investigate
and correct the problem. Maintenance
technicians
inspected
the door seals and identified
no obvious seal damage but noted that dirt had accumulated
on the seal surf ace near the bottom of the door. The door seal was wiped clean with a damp rag and the LLRT was successfully
rerun. Following
the incident, Maintenance
and Operations
agreed to have Operations
personnel
wipe down the door seal and retest the airlock in the event of another airlock failure prior to contacting
the Maintenance
Department.
This was noted as the corrective
action in Incident Report (IR) #95-204. On May 3, 1995, the airlock again failed the local leak rate test. Operations
personnel
wiped down the door seal and satisfactorily
retested the airlock. The Operations
Department
initiated
IR #95-518 and Action Request (AR) #950503088
to evaluate and document the occurrence.
On May 5, 1995, an LLRT was satisfactorily
conducted indicated
an elevated leakrate.
On May 8, 1995, the airlock failed its LLRT for the third time. The door seal was wiped down and the LLRT was successfully
rerun. Operations
initiated
IR #95-551. and AR #950508110
to troubleshoot
and correct the recurring
condition.
Subsequent
investigation
revealed a .significant
buildup of dirt and hardened grease in the groove on the seal surface which was caused by the gasket set. Seal surface wipedown would not have been effective
in removing * this buildup .
- * * Document Control Desk LR-N95196
ROOT CAUSE ASSESSMENT
-2 -Attachment 4 -7TH example (cont'd) The root cause of this event has been attributed
to
adequate management
expectations
of system performance
as demonstrated
by: 1. Inexperienced
personnel
were assigned to perform the initial inspection
and corrective
actions. 2. Inexperienced
personnel
were assigned to perform the initial root cause evaluation.
Deficiencies
in both the preventive
maintenance
and surveillance
test procedures
also contributed
to this event. CORRECTIVE
STEPS THAT HAVE BEEN TAKEN The gaskets have been replaced on the Salem Unit 1 containment
personnel
airlock on the 100 foot elevation
and the leakage test was
performed.
The Salem Unit 1 containment
personnel
airlock (130 foot * elevation)
and equipment
hatch gaskets will be replaced prior to restart of Salem Unit 1 . The Salem Unit 2 gaskets on the containment
personnel
airlocks and equipment
hatch will be replaced prior to restart of Salem Unit 2. The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
Procedure
NC.NA-BP.ZZ-0002(Z), "Root Cause Analysis Guidelines" has been developed
to provide additional
information
and guidance in the use of various root cause analysis techniques
proven effective
in resolving
both human and equipment
performance
problems.
Within the Salem Maintenance
Department, PSE&G has dedicated
resources
to conduct required root cause analyses, develop and recommend
appropriate
corrective
actions, and assure their proper implementation
and overall effectiveness
through followup assessments.
Appropriate
Salem Maintenance
procedures
have been revised to include specific guidance on seal inspection, cleaning, and maintenance
to assist in troubleshooting
of leakage problems .
- * * Document Control Desk LR-N95196 -3 -Attachment 4 -7TH example (cont'd) Appropriate
Maintenance
procedures
will be revised to change the airlock seal lubricant
specification
from Dow Corning 111 to Dow Corning 3451 in accordance
with Nuclear Engineering
recommendations.
The above procedure
revisions
will be completed
prior to restart of either Salem Unit 1 or 2. CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
A Preventive
Maintenance
Change Request (PMCR) has been initiated
to evaluate the need for additional
Preventive
Maintenance (PM) tasks for the containment
airlock gaskets. The PMCR recommends
PM's following
the six-month
Structural
Integrity
Test and gasket replacement
at the end of each refueling
cycle. Appropriate
Operations
Department
procedures
will be revised to provide guidance on maintaining
seal surf ace cleanliness
and for performing
leak rate testing, including
discrete leakage criteria for determining
when additional
corrective
action is required.
These corrective
actions will be completed
prior to restart of the affected unit . DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The cause of the airlock seal f ailu*re occurrences
was documented
and evaluated, and the condition
was corrected.
PSE&G will have achieved compliance
with lOCFRSO Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit 1 or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -VIOLATION
ATTACHMENT 4 -BTH EXAMPLE 8. From February 29, 1992 until June 7, 1995, Salem Unit 1 staff failed to correctly
determine
the cause or take action to preclude recurrence
of failures of instrument
lines connected
to the jacket water cooling system for the No. lB and No. lC emergency
diesel generators.
RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
On June 1, 1995, during a lB Emergency
Diesel Generator (EDG) Surveillance
Test, a jacket water leak was identified
at the threaded connection
of a 1/4" pipe nipple to an elbow upstream of instrument
root valve 1DA46B. The failed component
was subsequently
replaced in kind. As part of the root cause analysis for the 1/4" nipple failure, natural
frequency
tests were performed
on all EDG's (at both Salem Unit 1 and 2) at locations
congruent
to-this failure. The'test results showed that piping at specific locations
on this and other EDG units could potentially
experience
damage or fail in response to induced vibrational
stresses.
The testing indicated
locations
with natural vibration
resonance
frequencies
very close to an integer multiple of the frequency
which corresponds
to the EDG shaft operating
speed. The affected EDG's were declared inoperable
pending further analysis.
A review of the past failure and maintenance
history of the Salem Unit 1 and 2 EDG's was performed
to identify occurrences
of similar failures.
PSE&G's analysis indicates
that there have been repeated failures due to yibration-induced
fatigue and the recurrent
nature of these failures was not recognized.
Failure to recognize
this repetitive
problem was due to inadequate
root cause analyses and the fact that the failures were attributed
to a wide variety of causes. Recommendations
stemming from this analysis included design change activities
t6 create more vibration-tolerant
configurations
and maintaining
failed components
for subsequent
laboratory
analysis.
Corrective
actions taken in the past were ineffective
at resolving
the vibration-induced
component
failures, as evidenced
by the recurring
nature of these problems .
- * * Document Control Desk LR-N95196
ROOT CAUSE ASSESSMENT -2 -Attachment 4 -BTH example (cont'd) PSE&G attributes
the root cause of the piping nipple failure to a design which did not adequately
include tolerance
for vibrational
stresses.
A contributing
cause was a lack of specifications
for dimensions
potentially
critical to vibration
tolerance
in the manufacturer's
documentation.
The root cause of the failure to identify and correct these component
failures was an inadequate
Corrective
Action Program (CAP). The CAP, in effect at that time, had numerous program elements which lacked adequate capacity for integration
and oversight.
The CAP did not facilitate
detection
of common failure elements nor did it ensure that conditions
adverse to quality were assessed for impact on Operability
in a timely manner. CORRECTIVE
STEPS THAT HAVE BEEN TAKEN All affected EDG's were declared inoperable
but available, pending resolution
of the potential
for vibration-induced
failure. Interim contingency
plan guidance was provided to the Operations*Department.
This guidance established
requirements
to maximize the availability
of the demineralized
water supply to fill the EDG jacket water system in the event of a postulated
failure. A short-term
adjustment
to the cantilever
length of the affected piping was made. This action reduced the potential
for resonance
between this piping and the engine/header.
A vibration
tolerance
design review of the EDG's and peripheral
equipment
was conducted.
This review resulted in recommendations
for appropriate
enhancement
modifications
to harden the diesel engines against vibration-related
concerns.
The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
A "lessons-learned" memorandum
relative to this issue was issued by the Manager -Nuclear Engineering
Design (NED) to all appropriate
NED personnel.
rolldowns
to NED personnel
have communicated
the "lessons learned."
- * * Document Control Desk LR-N95196 -3 -Attachment 4 -BTH example (cont'd) CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
Jacket Water Pressure transmitting
tubing runs will be redesigned
to eliminate
the piping nipples and associated
piping isolation
valves. This work has been completed
for lB EDG. Modification
packages have been prepared for the remaining
five EDG's. The Design Change Process (DCP) checklists
will be revised to include specialty
engineering
review for vibration-induced
failure issues. These changes will be incorporated
at the next revision to the appropriate
procedures.
Maintenance
and Planning Department
training programs will be revised to include specific information
regarding
the general nature of fatigue failure and system vibratory
response.
These corrective
actions will be completed
prior to restart of either Salem Unit 1 or 2. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED The cause of the instrument
line failures was identified
and actions were taken to reduce their susceptibility
to
induced failures . PSE&G will have achieved compliance
with 10CFR50 Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart Salem Units 1 and 2 until performance
in this and other areas has improved.
i ! .
- * * Document Control Desk LR-N95196 -1 -VIOLATION
ATTACHMENT 4 -9TH EXAMPLE 9. From July 11, 1992 until June 10, 1995, Salem staff failed to determine
the cause, evaluate the potential
safety consequences, and establish
corrective
action for an abnormal condition
affecting
the No. 21 Residual Heat Removal discharge
manual isolation
valve (21RH10) associated
with impact noise from the interior of the valve. (05013) RESPONSE -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
On July 11, 1992, during mode 5 operation, unusual noise was identified
coming from Residual Heat Removal (RHR) system valve 21RH10. A slightly lesser noise was heard from valve 22RH10 and from the Unit 1 operating
RHR loop No. 21. A Salem Technical
Department
Memo (92-138) was issued to inform the Operations
Department
that this noise may be caused by flow-induced
vibrations
from existing play in*the male/female
discs and/or disc arm. On April 21, 1993, a maintenance
activity to open and inspect the valve was completed
and wear marks were found on the downstream
seat in two locations.
The cause of the wear marks was attributed
to "wedge banging against seat ring." No internal pa.rts were found in need of replacement
with the exception
of packing and a gasket. On June 10, 1995, valve 21RH10 was again reported making metallic banging noise internally.
A maintenance
supervisor
did an in-field observation
of the valve and concluded
that the noise was abnormal.
An Action Request (AR) was written documenting
the noise; however, a formal Operability
Determination
to assess the impact of the noise on system functional
capability
was not documented
prior to June, 1995 .
- * ** Document Control Desk LR-N95196
ROOT CAUSE ASSESSMENT -2 -Attachment 4 -9TH example (cont'd) PSE&G has determined
that the root causes of this event were: 1. Inadequate
performance
by the System Engineer regarding
record keeping and tracking/trending
of equipment
malfunctions. . 2. Inadequate
Corrective
Action Program (CAP) as indicated
by: -Inadequate
management
and supervisory
oversight
of equipment
failure follow-up.
-Lack of documented
engineering
analysis of the physical condition
of the valve. The Corrective
Action Program (CAP), in effect at that time, lacked sufficiently
low thresholds
to ensure that conditions
adverse to quality would be identified
and resolved in a timely manner. That same program did not provide clear guidance on the need to perform nor the required content of assessments
to support continued
assurance
of equipment
operability.
CORRECTIVE
STEPS THAT HAVE BEEN TAKEN On June 15, 1995, Salem System Engineering
completed
a Follow-up
Assessment
of Operability
and determined
that the valve noise did not *adversely
affect the functional
capability
of the RHR system. Work Orders have been issued to open and inspect valve 21RH10 to determine
the reasons for the noise currently
being experienced.
Valve 21RH10 is scheduled
to be opened and inspected
after Unit 2 core off-load.
For the purpose of trending, vibration
data on valve 21RH10 is being taken periodically
and reviewed by the System Manager. This will continue until the loop is taken out of service following
core off-load.
The vendor for the valve was contacted
to obtain recommendations
on actions to be taken. The vendor stated
because of the valve design and its location in a turbulent
flow area, impact noises can be expected.
The vendor did not recommend
any periodic preventive
measures.
The Corrective
Action Program (CAP) has been revised as described
in the cover letter to this Attachment.
L
- * * Document Control Desk LR-N95196 -3 -Attachment 4 -9TH example (cont'd) System Engineering
Department
roles and responsibilities
have been identified
and clearly communicated
to all System Engineering
personnel.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
Additional
corrective
actions, if any, will be identified
after valve disassembly
and inspection, as stated above. System readiness
reviews are currently
underway and include assessment
of the readiness
of plant systems to support unit restart. Self-assessments
of the effectiveness
of the system engineering
organization
to carry out its roles and responsibilities
will be conducted.
These corrective
steps will be completed
prior to restart of either Salem Unit 1 or 2. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED Salem System Engineering
issued their Followup Assessment
of Operability
for this valve condition
on June 15, 1995 . PSE&G will have achieved compliance
with 10CFR50 Appendix B, Criterion
XVI, when the Corrective
Action Program and related processes
have been proven effective
at identifying
and resolving
conditions
adverse to quality in a timely manner. PSE&G will not restart either Salem Unit l*or 2 until performance
in this and other areas has improved .
- * * Document Control Desk LR-N95196 -1 -ATTACHMENT
5 VIOLATION
II. 10 CFR Part 50, Appendix B, Criterion
V, "Instructions, Procedures, and Drawings", requires that activities
affecting
quality shall be prescribed
by documented
instructions, procedures, or drawings of a type appropriate
to the circumstances, and shall be accomplished
in accordance
with these instructions, procedures
and drawings.
Instructions, procedures, or drawings shall include appropriate
quantitative
or qualitative
acceptance
criteria for determining
that important
activities
have been satisfactorily
accomplished.
Contrary to the above, following
a modification
in May 1993, that installed
a drain system for the Salem Unit 2 Pressurizer
code safety loop seals, the Licensee did not ensure that an activity affecting
quality was satisfactorily
accomplished
in that the procedure
that directed the installation
of the modification
to the Pressurizer
code safety loop seals drains did not adequately
ensure that the drain valves were properly positioned
prior to plant startup after the modification.
Specifically, valve 2PR66, a valve in a common drain line for the 2PR3, 2PR4, and 2PR5, Pressurizer
safety valves, was left closed throughout
the operating
cycle between May 1993 and October 1994. (06013) This is a Severity Level III Violation.
I) Civil Penalty -$100,000 RESPONSE ** -DESCRIPTION
OF CIRCUMSTANCES
PSE&G does not dispute the violation.
During the 2R7 outage, a design change package (DCP) was implemented
to *add drain lines and drain valve 2PR66 to the Pressurizer
Overpressure
Protection
system. Valve 2PR66 was installed
to drain the line downstream
of the Pressurizer
Safety Valve ioop seals in order to prevent potential
I : ** * * Document Control Desk LR-N95196 -2 -Attachment
5 (cont'd) Final testing of the newly installed
drain lines was completed
on April 27, 1993. On October 19, 1994, in preparation
for the Salem Unit 2 eighth refueling
outage (2R8), valve 2PR66 was discovered
to be closed. Valve 2PR66 being left in the closed position prevented
drainage of the Power-Operated
Relief Valve (PORV) and Pressurizer
Safety Valve loop seal lines and established
the loop seals, thus defeating
the purpose of the design change. After valve 2PR66 was discovered
closed, the computer-based
Tagging Request and Inquiry System (TRIS) was checked to confirm the expected valve position.
The normal position for this valve is "open" in accordance
with TRIS. The exact time when valve 2PR66 was closed and why this occurred is indeterminate.
The most probable period when valve 2PR66 was manipulated
and left closed was determined
to be after flushing activities
were performed
as part of DCP testing. The DCP .included
verification
of the valve positions
during and at the end of the testing portion of the modification.
Valve 2PR66 was documented
to be open after the testing. Subsequent
to the testing, there was a final acceptance
walkdown of the system pr1or to turnover to Operations.
The DCP did not require a written component
list which documented
valve positions
during the walkdown.
As a result, valve 2PR66 was not verified to be open after the DCP, when the system was turned over to Operat.ions.
The Operations
DCP Coordinator
understood
that, in order to approve the design package turnover to Operations
for TRIS revision, it was only necessary
to verify that the component
change had been made in the computer database.
The Operations
DCP Coordinator
signed off the "Change Package Turnover to Operations" checklist
for DCP 2EC-3190, without ensuring that a temporary
valve position lineup (referred
to as an "auxiliary
lineup") had been or* would be performed
prior to returning
the
to power operation.
The Operations
DCP Coordinator
had not received any training related to expected roles and responsibilities
for providing
or receiving
the final component
configurations
after modification .
- Document Control Desk LR-N95196 -3 -Attachment
5 (cont'd) At the time of this event, there existed an excessive
TRIS backlog of 6000 changes waiting to be processed.
The Operations
Staff supervision
failed to take prompt action when the TRIS backlog became unmanageable.
TRIS database maintenance
received an inappropriately
low priority.
This was compounded
by the fact that the TRIS Coordinator
was assigned other collateral
duties. The TRIS Coordinator
did not create an auxiliary
lineup in accordance
with SC.OP-DD.ZZ-OD16, "TRIS Operations." Procedure
SC.OP-DD.ZZ-OD16
does not specify a time limit for performing
an auxiliary
lineup. However, an auxiliary
lineup was expected to have been performed
prior to declaring TRIS database complete. "RC-MECH-001" is a standard valve lineup used to restore affected systems to a ready condition
in preparation
for plant startup. The revision of SC.OP-DD.ZZ-OD16, in effect* at the time valve 2PR66 was added to the database, specified
that the auxiliary
lineup pe completed
and confirmed
in TRIS before the component
is added to its applicable
standard lineups. The auxiliary
lineup for valve 2PR66 was delayed and eventually
never performed.
As a result, valve 2PR66 was not added to the RC-MECH-001
lineup in a timely manner . ROOT CAUSE ASSESSMENT PSE&G has determined
that the root cause of this event was inadequate
commitment
to the DCP turnover process and TRIS maintenance
program by Operations
Management
as demonstrated
by the fallowing: . * 1. Operations
had less-than-adequate
turnover acceptance
of DCP's. Roles and responsibilities
were not clearly defined. Supervision
failed to communicate
expectations
effectively.
2. The Operations
department
allowed the TRIS database to become unmanageable.
The backlog was accepted.
The safety significance
of the backlog on system design and operability
was not adequately
evaluated .
- * Document Control Desk LR-N95196 -4 -CORRECTIVE
STEPS THAT HAVE BEEN TAKEN Attachment
5 (cont'd) The Operations
Department
reviewed TRIS database change requests initiated
from DCP 1 s completed
during the time period from the beginning
of 2R7 to the present. This review encompassed
485 DCP's which have gone to "Part A" closure since the beginning
of 2R7. Part A closure signifies
that the activity has been installed
and the DCP has been turned over to Operations.
PSE&G has evaluated
the elements of the DCP process which ensure that Operations
procedures
and the TRIS database are updated. This evaluation
has determined
that the process is adequate.
The TRIS backlog was reduced to zero in May of 1995. The backlog is being maintained
at zero. Operations
has assigned additional
personnel
as TRIS coordinators.
The Coordinators
are responsible
for all TRIS interfaces
including
procedure
SC.OP-DD.ZZ-OD16.
Operations
Senior Reactor Operators (SRO's) have been assigned ownership
of plant systems. The SRO interacts
with the project managers and System Managers associated
with the DCP from conception.
The SRO accepts responsibility
for system turnover *to Operations.
CORRECTIVE
STEPS TO BE TAKEN TO PREVENT RECURRENCE
Operations
procedure
SC.OP-DD.ZZ-OD16
is being revised. The revision will emphasize
Operations
management's
expectations
and incorporate
the auditing process for TRIS revision requests.
This will be completed
by December 1, 1995. DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED Valve 2PR66 was correctly
positioned
for existing plant conditions.
PSE&G has evaluated
the DCP process relative to accomplishing
appropriate
valve positioning
after modification
activities
are complete.
This evaluation
indicates
that the process is adequate.
- l , l . [