ML17164B002
| ML17164B002 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/05/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17164B001 | List: |
| References | |
| 50-387-99-02, 50-387-99-2, 50-388-99-02, 50-388-99-2, GL-96-01, GL-96-1, NUDOCS 9904120212 | |
| Download: ML17164B002 (33) | |
See also: IR 05000387/1999002
Text
U.S. NUCLEAR REGULATORYCOMMISSION
REGION I
Docket Nos:
License Nos:
50-387, 50-388
Report No.
50-387/99-02, 50-388/99-02
Licensee:
Pennsylvania Power and Light Company
2 North Ninth Street
Allentown, Pennsylvania
19101
Facility:
Susquehanna
Steam Electric Station
Location:
P.O. Box 35
Berwick, PA 18603-0035
Dates:
January 19 through January 29, 1999
Inspectors:
R. Fuhrmeister, Senior Reactor Engineer
G. Morris, Reactor Engineer
D. Vito, Senior Allegation Coordinator
Approved by:
Lawrence T. Doerflein, Chief
Engineering Programs Branch
Division of Reactor Safety
9904120212
990405
ADOCK 05000387
8
PDR'
I
EXECUTIVE SUMMARY
Susquehanna
Steam Electric Station (SSES), Units 1 8 2
NRC Inspection Report 50-387/99-02, 50-388/99-02
This inspection included aspects of Pennsylvania Power and Light Company's (PP8L's)
corrective action program and PP8L's response to NRC Generic Letter 96-01.
~En ineeiin
The inspector concluded that PP8L's response to Generic Letter 96-01 adequately
addressed
the issues identified in the GL, and that there was an adequate
basis for the
positions taken in the response.
(Section E1.1)
The inspectors concluded that the surveillance testing procedures reviewed
satisfactorily met the testing requirements outlined in NRC Generic Letter 96-01.
The inspectors also determined that the minor discrepancies
noted in the procedures
and drawings would not have affected the outcome of the testing.
(Section E2.1)
The inspector determined that the procedures for control of modifications to the
facility provided appropriate controls for ensuring that modifications were properly
carried over into the technical specifications and surveillance tests.
(Section E3.1)
The condition report (CR) process
is a high volume, low threshold corrective action
system that is acceptable to meet the requirements of 10 CR 50, Appendix B,
Criterion XVI. Adverse conditions are promptly identified and the process appears
to be widely accepted for use, and is used by a broad cross section of the plant
staff. There are no significant delays in the initial assessment,
investigation, and
completion of initial operability and reportability determinations.
Initial investigations
and root cause assessments
for issues of higher safety significance are thorough.
Investigations and causal analyses for CRs of lesser safety significance are generally
thorough, but there are some instances where extent of conditions and/or generic
implications may not be sufficiently explored.
(Section, E7.1)
The condition report process
is focused on accomplishing initial reviews,
reportability and operability determinations,
cause assessments,
and establishment
of proposed corrective actions to correct the condition and prevent recurrence.
However, process accountability is not readily apparent
in the corrective action
implementation portion of the process.
Corrective action implementation dates do
not correlate with the assigned significance level. Action due dates are controlled
by the responsible manager who may change them during the course of
implementation. The only procedural requirement regarding corrective action
implementation is refueling outage related with the shortest lead time being
approximately 7-'10 months, and the longest lead time being approximately 31-37
months.
The minimal administrative control over the corrective action
implementation portion of the process contributes to a high process backlog.
(Section E7.1)
Executive Summary (cont'd)
~
Internal reviews of the condition report system by Operating Experience Services
(OES) and external reviews by Nuclear Assurance Services (NAS), the Cooperative
Management Audit Program, the Institute of Nuclear Power Operations, and the
Susquehanna
Review Committee are continuing to find incomplete or inadequate
corrective action closure.
A recent process change requiring OES to review
competed'actions
to ensure that the actions satisfies the one prescribed,
is reducing
the amount of items identified. However, problems persist in this area.
(Section
E7.1)
Annual assessments
of the condition report process by NAS were thorough, critical,
and well founded.
NAS assessments
conducted in 1997 and 1998 identified most
of the concerns identified by the inspectors.
(Section E7.1)
.
TABLEOF CONTENTS
EXECUTIVE SUMMARY..
TABLEOF CONTENTS
.IV
III. Engineering
.
E1
Conduct of Engineering
.
E1.1
Response to NRC Generic Letter 96-01
.
E2
Engineering Support of Facilities and Equipment
E2.1
Functional Testing of the Diesel Generator "C" Starting Circuits
E3
Engineering Procedures and Documentation
E3.1
Modification Process Procedures
..
. 3
.3
E7
Quality Assurance in Engineering Activities
E7.1
Corrective Action Program
.
V.
Management Meetings
X1
Exit Meeting Summary
.
16
16
Re ort Details
III. En ineerin
E1
Conduct of Engineering,
E1.1 'es
onse to NRC Generic Letter 96-015
5 251
92
To assess
the adequacy of the Pennsylvania Power and Light (PP8L) response to NRC Generic Letter (GL) 96-01, the inspector reviewed PP8L's response,
dated
April 18, 1996, Licensee Event Reports (LERs) for 1996, 1997, 1998, and Condition
Reports (CRs) for those LERs from the selected period which identified inadequate
testing.
b.
Observations and Findin s
Industry Experience Review Program Action Item (IERP) No. 96007 was opened on
January 19, 1996, to track PP8L actions in response to the. generic letter. The IERP
developed two recommended
actions based on a review of the Surveillance Program
Task Force activities in 1984. The recommendations
were to upgrade the highlighted
drawing system and to revise the surveillance test program implementing procedure to
incorporate the testing philosophy develop'ed by the task force in 1984. The PP8L
'esponse
to GL 96-01 stated that the requirements of'the GL had been met since the
surveillance testing procedures for the Susquehanna
Steam Electric Station (SSES) are
in compliance with the Units'echnical Specifications.
This conclusion was based upon
having performed a review of surveillance testing during the startup testing of Unit 2 in
the 1984 time frame and having an ongoing program to ensure that modifications are
properly tested and incorporated into surveillance tests.
Additional reviews of testing
were not conducted at the time of the response.
9
During the time period 1996 through 1998, seven LERs identified inadequate
surveillance testing of safety related circuits. Two of these were related to response time
testing, and were, therefore, outside the scope of the activities requested
by the generic
letter. CRs generated
by PP8 L to address the issues identified in the other five LERs
were reviewed to determine what corrective actions PP8L had implemented and the
extent of the conditions.
LER 50-387/98-12, which was also applicable to Unit 2, documented that the testing of
the suppression
pool temperature monitoring instrumentation did not meet the
requirements for a logic system functional test of the alarm function of the instruments.
This condition was evaluated under CR 98-1972.
PPBL determined that past tests
actually only performed a channel calibration. Alternative testing to meet the channel
functional test requirement was identified and approved by the plant operations review
committee (PORC).
The CR identified four other CRs which identified occasions where it
was discovered that the surveillance tests did not properly implement surveillance
requirements.
The evaluation of the five deficiencies in total concluded that there was no
significance since-"These errors represent 0.5% (10 of 2000) of all the station's
surveillance procedures."
LER 50-387/98-10, which was also applicable to Unit 2, documented that the main
feedwater turbine and main generator turbine trips on reactor vessel high water level
were not adequately tested.
This condition was documented in CR 98-0697. The testing
was carried out by use of overlapping tests, which it turns out, did not test the final
outputs to the trip function. The final output device was in fact tested, but the test was
not called out for satisfying the surveillance requirement.
This CR developed a time line
for the test methodology, which showed that this problem was first identified during the
surveillance testing review. in 1984. Corrective actions at the time included a procedure
change; however, it was not incorporated into the next procedure revision. The licensee
plans to revise the procedures to correct the discrepancy prior to the next required test.
LER 50-387/98-04 documented a failure to test portions of the residual heat removal
(RHR) pump start logic. This condition was evaluated under CR 98-0427 and 98-0528,
which determined that the power monitor relays for all four pumps in both plants had not
been tested.
Test procedures were revised and the relays were satisfactorily tested.
In
response
to this condition, several other systems were reviewed to determine ifsimilar
conditions existed.
No similar problems were identified in the other systems.
An
additional corrective action was assigned to evaluate the need to revise or supplement
the PP&L response to GL 96-01.
PP&L determined that a supplement was warranted,
and tracking item E20235 was generated to ensure completion of the supplemental
response.
A due date of September
1, 1998, was originally assigned.
At the time of this
inspection, the due date had been previously'extended to January 29, 1999. At the end
of the inspection, the due date was again being extended, with completion expected in
mid- to late February, 1999.
Conclusions
The inspector concluded that PP&L's response to Generic Letter 96-01 adequately
addressed
the issues identified in the GL, and there was an adequate basis for the
positions taken in the response.
Engineering Support of Facilities and Equipment
Functional Testin
of the Diesel Generator "C" Startin
Circuits
Ins ection Sco e (IP Tl 2515/139)
The team reviewed selected portions of control logic drawings, schematic diagrams, and
surveillance test procedures to assess
the completeness
of PP&L's logic system
functional testing.
/
b.
Observations and Findin s
The inspectors reviewed the surveillance tests associated with the Division I Loss of
Offsite Power (LOOP) and Loss of Coolant Accident / Loss of Offsite Power
(LOCNLOOP) initiated starting of the "C" diesel generator and loading to the Unit 1 4.16
kV Bus 20301.
In 1998, PP8L had completely revised the Unit 1 surveillance test
procedures to incorporate the changes
in the surveillance test requirements which
resulted from the approval of the Susquehanna
Steam Electric Station (SSES) Improved
Technical Specifications (ITS), and to incorporate the 24 month operating cycle. The
surveillance procedures which the inspectors reviewed were clear and well written.
PP8L uses multiple overlapping procedures to test the complete circuits covering the
LOOP and LOCNLOOP logic from the initiating instrumentation to the diesel start, bus
load shedding, load sequencing and bypass of the diesel generator non-essential
protective trips for each diesel, each unit, and each division. The procedures
satisfactorily tested the portions of the circuits reviewed by the inspectors.
The
inspectors found only minor errors in several of the procedures and drawings where
associated
schematic drawings were incorrectly referenced (for information only) and
which did not affect the implementation nor the results of the test procedures.
Minor
technical discrepancies
between the surveillance procedures and the Final Safety
Analysis Report (FSAR) had already. been identified by PP8L, had been processed
through their condition report program and a licensing document change notice (LDCN)
had been issued to correct the FSAR.
The team observed that PP8 L did not test the active protective trips of the diesel
generator (generator differential overcurrent, engine overspeed,
and low lubricating oil
pressure) during operation in the emergency mode.
During discussions with the
Improved Technical Specification group in the Office of Nuclear Reactor Regulation
(NRR) the inspectors confirmed that testing these three protective trips is not specifically
required by ITS Surveillance Requirement 3.8.1.13.
b.
Conclusions
The inspectors concluded that the surveillance testing procedures reviewed satisfactorily
met the testing requirements outlined in NRC Generic Letter 96-01. The inspectors also
determined that the minor discrepancies
noted in the procedures and drawings would not
have affected the outcome of the testing.
E3
E3.1
Engineering Procedures and Documentation
Modification Process Procedures
The inspector reviewed procedures which control the design and installation of
modifications at SSES, to determine how modification information is carried over into the
technical specifications and surveillance test procedures.
Observations and Findin s
Nuclear Department Procedure MFI-2309, Rev. 6, "DCP/ECO Preparation Instructions,"
provides "how-to" instructions for preparing items associated
with Design Change
Packages
(DCPs) and Engineering Change Orders (ECOs). The procedure specifies
that appropriate testing strategies, methods, specifications, parameters
and acceptance
criteria be developed during the preliminary design phase.
The modification, its design
and installation are to be developed by a team of personnel, whose specialties address
needs identified in the modification scoping report.
Nuclear Department Instruction MFI-3203, Rev. 1, "Guidelines for Installation Kickoff
Meeting (IKM),"outlines responsibilities for personnel attending the IKM,which must be
held prior to issuance of each plant modification package (PMP). The procedure
provides guidance for developing installation and testing strategies, and assigns the
associated
system engineer the responsibility for determining ifthe modification affects
other systems, nuclear instrumentation, thermal power heat bala'nce inputs, and whether
any procedure changes are required as a result of the modification.
ti
NDAP-QA-1211, Rev. 3, "SSES DCP/ECO Installation Process," ensures that the
installation process for DCPs and ECOs is performed in a consistent, controlled, and
documented manner.
The procedure covers such activities as providing input to the
Engineering Integrated Schedule, preparing the PMP, ordering material, monitoring
installation.and testing, resolving installation problems, and closeout activities. Design
activities are specifically excluded, since they are covered by a separate procedure.
The
procedure also'requires a review of the modification design ifinstallation is delayed more
than thirty months.
This review is intended to ensure that the modification complies with
the current design and licensing basis of the plant. These reviews are conducted on a
case-by-case
basis for delays less than thirty months.
MFP-QA-3904, Rev.1, "Digital Control Scheme Testing," establishes
a standard method
of performing and documenting digital control scheme post modification inspections and
testing for DCPs and ECOs. The procedure requires, among other things, systematic,
thorough, contact-by-contact, conductor-by-conductor continuity checks of the modified
scheme, as shown on the designated drawings prior to energizing the circuit. The
procedure requires the testing to go one point beyond the modified wiring ifpractical, into
the circuitry outside the scope of the DCP/ECO. The requirements for energized testing
of the modification include identifying each logic path configuration, verifying proper
functioning of all contacts (both opening and closing), and testing each path to the end,
including operating initiation devices.
Improved Technical Specifications, license amendment
1?8 for Unit 1 and 151 for Unit 2,
defines a channel functional test as the injection of a simulated or actual signal into the
channel as close as practical to the sensor, and verifying operation of all required
alarm,'nterlock,
display, trip and channel failure trip functions. The definition allows testing to
be performed by sequential, overlapping steps, as well as total channel tests.
I
c.
Conclusions
The inspector determined that the procedures for control of modifications to the facility
provided appropriate controls for ensuring that modifications were properly carried over
into the technical specifications and surveillance tests.
E7
Quality Assurance in Engineering Activities
E7.1
Corrective Action Pro ram (IP 40500)
To evaluate PP8L's program for identifying and correcting conditions adverse to quality,
the inspectors reviewed the site procedures governing identification and correction of
adverse conditions and corrective action system files selected at random, both open files
and files which had been closed within the previous year.
b.
Observations and Findin s
At SSES, the Condition Report (CR) process is the predominant means of identifying,
documenting and correcting conditions adverse to quality. The CR process is a high
volume, low threshold problem reporting system, which in addition to plant safety issues,
may also include site issues that are not specifically related to nuclear safety.
Th'
current CR process has been in place since 1996, when it replaced several separate
processes
that tracked issues in various functional areas (daily operational issues,
quality assurance
issues, radiological controls issues, industry experience issues, etc.).
The current process encompasses,
but is not limited to, the following types of concern:
daily operational and maintenance concerns
quality assurance
audit and surveillance findings
industry experience issues
issues identified by PP8L site engineering
issues identified by PP8L corporate engineering
Institute of Nuclear Power Operators (INPO) issues
NRC generic correspondence
other concerns and requests for information referred to PP8L by NRC
procurement issues (ifnot caught by the receipt inspection program)
technical issues referred from the SSES Employee Concerns Program (ECP)
programmatic and human performance issues related to health physics and
radiological controls
issues identified due to apparent trends from previously identified CRs
issues identified due to incomplete or incorrect corrective actions from previous
CRs
The CR process does not include industrial safety issues (although some significant
industrial safety issues may be included in the CR process).
The inspectors reviewed the current CR procedure (NDAP-QA-0702), and the current
versions of the following procedures which denote interface with the CR process:
NDAP-QA-0703 - Operability Assessments
and Requests for Enforcement Discretion
NDAP-QA-0720 - Station Reporting Matrix and Reportability Evaluation Guidance
NDAP-QA-0721- Human Performance Evaluation System (HPES)
NDAP-QA-0730 - Controlling Changes to Licensing Documents
NDAP-QA-0726 - Safety Evaluations
NDAP-QA-0300 - Conduct of Operations
NDAP-QA-0502 - Work Authorization System
NDAP-00-0109 - Nuclear Safety Concerns of Individuals
OES-AD-001 - Condition Report Processing/Event
Trending
The inspectors obtained and reviewed a summary listing of currently open CRs and a
listing of all CRs closed in calendar year 1998.
From the summary listings, the
inspectors selected and performed a detailed review of 45 CR files (17 closed files and
28 currently open files). The inspectors discussed aspects of the CR process and the
specific CRs reviewed with several members of the SSES plant staff from various
functional departments.
Process Guidance and Trainin
The CR procedure (NDAP-QA-0702) underwent a major revision in September 1998.
The most notable changes include the following:
The procedure clarified the role of Operations in initial operability/reportability
assessments,
to assure that Shift Supervision performs operability/reportability
determinations promptly when appropriate.
The licensee indicated that in the
past, there had occasionally been delays in obtaining initial
operability/reportability assessments
for CRs that were not submitted directly to
Operations.
The procedure added a requirement for Operating Experiences Services (OES),
the department that administratively tracks the CR process, to review completed
actions to ensure that the action taken satisfies the action prescribed.
This
function was instituted due to a persistent process problem which, due to a lack
of contemporaneous
corrective action implementation review, resulted in
numerous issues being revisited because of corrective action being either
inadequate,
incomplete, or not being performed at all.
The procedure added a new requirement for an "effectiveness review" to be
performed for CRs of higher significance, to obtain a more prompt assessment
of
the effectiveness with which more significant concerns are resolved, and to
evaluate whether the actions taken were successful in preventing or significantly
reducing the potential for recurrence of the initial event. At the time of the
inspection, none of these effectiveness reviews had been completed, so none
were available for review.
It should be noted that the effectiveness review will
also affect the overall timeliness for CR closure since it is intended that affected
CR files will not be closed until the effectiveness review is complete.
The CR procedure and other related procedural guidance is relatively clear and easy to
follow. Forms are provided for documentation of the problem, the operability and
reportability assessments,
the root cause or causal analysis, the initial investigation into
the matter, past experience, additional safety assessments,
proposed actions to correct
the problem and prevent recurrences,
Human Performance Evaluation System (HPES)
information, and assignment of corrective actions.
The process is being implemented
as described in the procedure and appears to get wide use.
Regarding the recent full re-write of the CR procedure, each functional group was
provided a training package to brief their staff on the new procedure.
New employees
are provided training on the CR process during their initial indoctrination. OES also
provides periodic update training to engineering personnel at the PP8L corporate office
in Allentown. Also, the OES Supervisor issues periodic electronic mail messages
to the
site staff to discuss changes and enhancements
to the process.
Recent focus is'on the
implementation of the SSES electronic database,
the Nuclear Information Management
System (NIMS), and the services it willmake available to the PP8 L staff regarding CR
activities and status.
Process
Im lementation/ WorkfIow
CR Initiation and Initial Assessment
The CR process applies to the entire site staff and also to personnel in the PP8L
Corporate Office who interface with SSES personnel and activities. Employees are
encouraged to submit Condition Reports through their supervisor, but this is not
specifically required.
This is acceptable since an employee may, in some cases, feel
apprehensive
about having to obtain his/her supervisor's acceptance
before a concern is
submitted.
The inspectors noted that CRs were submitted through the supervisor in the
large majority of cases.
Based on the number of CRs submitted, and the large variety and variability of concerns,
the process appears to be widely accepted for use, and is used by a broad cross section
of the PP8L staff.
Ifan employee wishes to remain confidential, he/she may report their
concern through the Employee Concerns Program (ECP).
Ifappropriate, ECP willfile
the employee's technical issues with the CR process and provide feedback to the
individual with regard to the resolution of each issue.
The inspectors discussed this
aspect of the process with the site ECP manager, who indicated that the number of ECP
issues that he has had to channel through the CR process has been minimal, and that
there has not been a problem with providing feedback to individuals who channel their
technical concerns through ECP.
There is no specific requirement to provide feedback
to those individuals who file CRs through the normal process.
CR issues that may impact plant operations, the plant Technical Specification, or the
SSES Technical Requirements Manual are to be provided to Shift Supervision so that
initial operability/reportability assessment
may be performed.
Ifa CR is submitted to
OES before it is presented to Operations, OES assures that the CR is provided to
Operations promptly so that initial operability/reportability assessment
may be
performed.
Upon receipt, CRs are processed
through OES for initial screening and establishment
in
the tracking database.
An initial screening meeting is held within OES within a day or
two of the receipt of the CR at which a significance level is assigned to each CR.
Significance is assigned as follows:
Level 1 - Significant and Consequential
Level 2 - Significant or Consequential
Level 3 - Low Significance and Low Consequence
Level 4- Precursor or Minor
The OES screening meeting is attended by the OES Supervisor and his evaluation staff,
and often includes representatives
from the Operations and Maintenance departments to
provide additional insight into the issues raised.
The OES Supervisor indicated that his
department has made a concerted effort over the past year to categorize CRs
appropriately, since in the past, significance levels may have been assigned too
conservatively.
In particular, the OES Supervisor indicated that many issues that should
have been categorized as Level 3, were categorized as Level 2, requiring more
evaluation and effort, and contributing to the already high backlog. Current efforts in
assigning significance levels appear to be appropriate and consistent with the procedure.
In addition to the assignment of a significance level, the OES screening meeting
determines the responsible manager, attempts to identify any apparent causes or causal
factors, and whether any interim corrective actions may be appropriate before longer
term corrective actions are proposed by the responsible manager.
The screening
meeting also evaluates whether any specific equipment failure constitutes a
Maintenance Rule functional failure.
The OES screening meeting also attempts to determine whether a given CR describes
an event similar to a previous event or events, or has similar causes to those of previous
events.
Ifso, OES may consider initiating a new CR to demonstrate the existence of an
adverse trend.
OES bases its decision to open a "trend",CR on the knowledge of the
OES staff with regard to previous CRs that have been, received.
"Trend" CRs may also
be initiated as a result of findings made by other assessments
of the corrective action
program (QA audits and surveillances, INPO, etc.). The inspectors noted that there are
no formal criteria to establish what is or is not an adverse trend, and there is no
structured method of performing periodic trending. The OES supervisor acknowledged
that with the'recent inclusion of CR database
information on the Nuclear Information .
Management System (NIMS), the opportunity is available to develop more formalized
trend analysis and reporting capabilities.
The OES Supervisor has recently assigned a
staff member to develop electronic trending capabilities.
Within a day or two of the OES screening meeting, a CR is scheduled for discussion at
the daily Corrective Action Team (CAT) Meeting.
The CAT meeting is attended by
higher level managers from the site functional departments
and OES personnel.
Corporate engineering is telephonically linked. At the CAT meeting, OES personnel
present Level 1 and Level 2 CRs and the results of the initial OES screening of those
CRs.
Comments and changes are incorporated into the CR by OES.
Level 3 and Level
4 CRs are not discussed at the CAT meeting unless questions are asked by meeting
attendees.
Afterthe CAT meeting, Level 1 and Level 2 CR packages,
and Level 3 CR packages for
which an evaluation is required are issued to the responsible manager, who will
determine the course of action to resolve the CR and prevent its recurrence, ifpossible.
At this point in the process, the responsible manager develops the investigative
documentation for each CR as required by NDAP-QA-0702, and puts together the
Evaluation and Action Plan (E8AP). Level 1 CRs require an investigation into the
concern, a safety assessment,
a root cause assessment,
an assessment
of causes and
causal factors, and development and assignment of corrective actions.
Level 2 CRs
require a similar review to Level 1 CRs, but only require an assessment
of apparent
cause versus a root cause assessment.
For Level 3 CRs requiring an evaluation, an
investigation is performed and causes and causal factors are assessed,
in addition to the
development and assignment of corrective actions.
For Level 3 CRs which simply
require that the noted condition be corrected, OES inputs the actions derived at the OES
screening meeting to the CR electronic database,
and assigns them to the responsible
manager.
The responsible managers must complete the initial evaluation of a CR and
determine the corrective actions to be completed according to the following schedule:
20 calendar days from the event date for Level 1 and 2 CRs involving reportable
or potentially reportable events
30 days from the CAT meeting for all other Level 1 and 2 CRs
120 calendar days for Level 3 CRs
CR Corrective Action lm Iementation
While the timeliness of developing the E8AP is correlated with the CR significance level,
the implementation of corrective actions is not. NDAP-QA-0702 describes the
development of the EBAP as the point of resolution of a CR. However, the deficiency is
not actually resolved at this point, as corrective actions have yet to be implemented.
The
completion dates for corrective actions are developed and assigned by the responsible
manager.
The responsible manager may change the due date of a given corrective
action at his/her prerogative during the course of its implementation.
The only
procedural requirement regarding the implementation of corrective actions is refueling
outage related.
Generally, corrective actions for CRs are assigned to the next refueling
outage which starts greater than 6 months from the event date.
For CRs affecting both
units, corrective actions are assigned to the refueling outage with the longest lead time.
In other words, the shortest lead time for implementing a CR corrective action would be 6
months plus the duration of the next refueling outage, (plus an additional month if
completion of a specific CR is not a requirement for restart).
The longest lead time for
10
implementing a CR corrective action could be 30 months (assuming a 24-month
operating cycle) plus the duration of the next two refueling outages (plus an additional
month ifcompletion of a specific CR is not a requirement for restart).
Ifthese
completion times cannot be met, extensions must be approved by the General Manager-
SSES and the General Manager - Nuclear Engineering.
These requirements do not
apply to the completion of modifications. The lack of administrative control over the
corrective action implementation portion of the CR process appears to be a significant
contributor to corrective action timeliness issues and the overall CR backlog. Of the
'690 CRs open at the beginning of the inspection, 2.3 % were greater than 2 years old
(including 3 Level
1 and 30 Level 2 CRs), 23+% were greater than
1 year old, and 53+%
were greater than 6 months old.
Since December 1998, some site managers
have implemented internal departmental
requirements that CR action due date extensions be negotiated and approved by the
responsible manager before the date can be changed in the database.
While this
appears to have imparted some degree of additional accountability into the process, the
cycle end date remains as the only established requirement for completion.
CR Closure
After corrective actions have been implemented, OES reviews the CR package to ensure
'hat implemented corrective actions are commensurate with those prescribed.
Ifa
corrective action is incomplete or inappropriate, OES refers it back to the responsible
manager for reassessment
and resolution. This aspect of the process was initiated in
October 1997, due to a persistent process problem which, due to a lack of
contemporaneous
corrective action implementation review, resulted in numerous issues
being revisited because of corrective action being either inappropriate, incomplete, or not
being performed at all.
When these reviews were initiated, OES indicated that a "reject
rate" of 15+% was encountered.
The efforts and oversight provided by OES for this
function over the past year have reduced the "reject rate" considerably, to less than 2%.
This effort notwithstanding, problems with corrective action implementation continue to
be identified by sources external to OES (QA, INPO, etc.). At the beginning of the
inspection, there were 34 open CRs related to inappropriate or incomplete completion, of
CR corrective actions.
Once the corrective actions are determined to be complete, OES closes out the CR in
the NIMS database.
CR Trendin
Trending capability is an important aspect of a corrective action program with the volume
and backlog of that retained by the SSES CR process.
As noted above, trend
assessment
of individual CR files is performed by the OES evaluators at the initial OES
screening meeting, based on their knowledge of past events.
The licensee
acknowledged that more can be done.
Efforts are being initiated to implement more
structured trending activities (e.g., keyword searching, establishment of indicators, etc.)
11
OES also provides weekly data trending to the technical staff which includes charts
depicting: the backlog of open condition report actions, number of CR actions issued and
completed per month by department, the number of Level 1 and Level 2 E&APs that
have yet to be developed and are overdue, and a "Top Ten List" of corrective actions to
be implemented.
OES also provides a quarterly briefing on adverse trends through
reference to open "trend" CRs and actions being taken to respond to them. There were
approximately.20 open "trend" CRs at the beginning of the inspection.
Audits
The SSES Corrective Action Program is audited annually by PP8L Nuclear Assurance
Services (NAS). The inspectors reviewed results of the most recent NAS assessment
performed in November -December 1998. The NAS assessment
was thorough, critical,
and well founded.
While the NAS assessment
noted improvement in some areas (issue
identification, timeliness for developing Level 1 ESAPs), it did not find measurable
improvement from the 1997 NAS assessment,
and noted that most corrective actions in
response to the prior assessment
were either untimely, incomplete or inadequate.
The,
1998 NAS assessment
summarized its findings into two general areas:
1) that corrective
actions to address adverse conditions are not always effective in preventing repeat
incidents; and 2) corrective actions to address adverse conditions are not always
implemented quickly enough to prevent repeat incidents.
The NAS assessment
highlighted that improvements in timeliness, trending, and oversight are needed in order
to establish control over the corrective action implementation portion of the CR process.
Improvements in these areas willreduce backlog, and reduce repeat occurrences.
The
SSES staff acknowledged that the problems noted in the NAS assessment
have
occurred in the past, and are still occurring, to a lesser degree.
The NAS assessment
findings along with the findings of other external assessments
of
the CR process (e.g., NAS audit and surveillance findings, INPO assessment
findings,
Cooperative Management Audit Program, CMAP, audits, etc.) are documented
in a CR
and entered into the tracking system for further review.
The Susquehanna
Review Committee (SRC), a corporate level safety review committee,
has a CR Review Subcommittee which performs periodic assessments
of the SSES
corrective action program.
The last SRC review, completed in October 1998, looked at
Level 1 and Level 2 reportable CRs that were recommended for approval by the Plant
Operations Review Committee (PORC). A notable concern from this review related to a
June 1998 CR on SRV acoustic monitors, which indicated that a Licensee Event Report
(LER) was written and that no supplement was expected.
The causal analysis for the
CR indicated that a supplement would be written ifthe cause was found to be different
than initiallystated.
The cause was eventually found to be different but no supplemental
LER was written. The inspectors also identified concerns with the corrective action
program's treatment of problems with the SRV acoustic monitors (discussed CR File
Review below).
12
CR File Review
The following items describe the results of the inspectors'eview of information in the CR
database
and specific CR files:
adverse conditions are promptly identified
no significant delays in initial assessment,
initial investigation, and completion of
operability determinations and reportability determinations.
For more recent CRs
(those opened within the past year), these activities are being completed within a
few days, whereas a year or more ago, these activities took a week or more to
complete.
Root cause assessments
for those which require them are thorough.
Investigations and causal analyses for CRs of lesser significance are generally
thorough, but are occasionally too specific to the system, structure or component
in question, and may not sufficiently explore extent of condition and/or generic
implications. This has been noted by OES, NAS, and other external reviews.
Examples noted by the inspectors were several CRs related to measuring and
test equipment (M8TE), in which corrective actions were focused on repair of the
M8TE and not on evaluation of the plant equipment that may have been tested
with the faulty M8TE. Problems with the non-safety-related emergency diesel
generator (EDG) pneumatic control system (discussed below) provide an
example of a system of lesser safety significance ultimately affecting the
operability of a safety-related system.
In response,
OES noted that recent efforts
have been initiated to not make the required actions too prescriptive, so as not to
restrict a responsible manager from developing appropriate closure.
Evaluation and Action Plans (E8APs) are generally established within the times
prescribed in Procedure NDAP-QA-0702. From the specific files reviewed, the
inspectors found seven (7) EBAPs that were established beyond the procedurally
prescribed due date, but extensions were requested
and approved by CAT, as
permitted by the procedure.
corrective action implementation dates do not correlate with the assigned CR
significance level. Action due dates are routinely extended by the assigned
department.
In a large number of cases, process accountability does not appear
evident until the End-of-Cycle due date approaches.
several instances were noted where final CR closure dates were several months
after the completion of the last action. This would appear to be attributable to CR
backlog and OES workload.
OES closeout reviews,-and independent reviews by NAS, INPO and CMAP are
continuing to find instances of incomplete or inadequate action closure.
There
also appears to be a tendency to adjust actions at the end of the process, e.g., by
categorizing issues as no longer necessary,
by changing actions, or by
13
transferring actions to another CR so that an earlier CR can be closed.
While
such adjustments can be expected in certain situations, better control of the
corrective action implementation portion of the CR process would minimize this
effort:
The inspectors found similar instances.
~Exam
las:
CR 96-1504 (9/1 6/96) and CR 96-2012 (10/31/96).
CRs dealt with an unmonitored
release path that was established from the turbine building exhaust system through the
reactor recirculation pump motor-generator (M-G) set ventilation system.
As a result of
an earlier design change, a damper was set to open upon shutdown of the M-G set
ventilation system, creating the release path.
Upon investigation, PP8L determined that
the design change package did not address'the
potential for establishing an unmonitored
release path. Afterthe repeat occurrence (CR 96-2102), a reportability evaluation
determined that, based on surveillance test results, this release path had existed for both
units for approximately 15 total days between July 15, 1996 and October 31, 1996. This
was deemed not to be reportable based on a conclusion that any releases from this
pathway during the times in question would have been within allowable release limits. In
the summary statement for CR 96-2102, PP8L recognized that a principal causal factor
for this repeat occurrence was due to untimely corrective actions for CR 96-1504.
CR 96-1751 (10/3/96) dealt with a half scram due to a nuclear instrumentation
intermediate range monitor (IRM) Channel B upscale trip. The trip was caused by
instrument spiking as a result of the collapsing field of an Agastat relay. Proposed
corrective actions included modifications for both units to filterthe instrument noise
spiking and procedure changes related to IRM ranging.
The Unit 1 instrumentation
modifications were completed in 1998 and the Unit 2 modifications are scheduled for
completion in March 1999. The procedure changes, which were scheduled to be
completed on December 1, 1997, were not implemented.
Instead, a conclusion was
reached on April 1, 1998, 18 months after the occurrence of the half scram, that the
modifications were adequate to reduce the probability of a half scram during startup and
shutdown conditions.
CR 3077 (9/1 7/97) dealt with an audit finding that a prior CR (97-1796) was closed
without performing all required actions, in that work to repair damaged rubber lining
inside a condensate
demineralizer transfer inlet line was not scheduled until after the CR
was closed.
Also, the cause determination was never done to address why the rubber
was coming offthe pipe. The closure mechanism for CR 97-3077 was to change the
classification of the CR from a Level 3 EVAL(requiring evaluation), to a Level 3
CORRECT (only requiring correction), and delete the requirement for engineering to
perform a cause determination review. Since the equipment in question is not safety-
related, the level of response
is at the license's prerogative.
However, the method of
resolution appeared to be a mechanism of convenience to eliminate an assigned action
that had not been accomplished.
14
CR 97-3199 (9/1 7/97) documented the improper closure of two earlier CRs (97-0771
(3/19/97) and 97-0801 (3/22/97)). The earlier CRs identified manganese
deposits on the
inner tube diameters for the core spray pump room coolers and the A and B residual
heat removal (RHR) service water heat exchangers.
The CRs documented an
acceptable cleaning method that was not completed before the CRs were closed.
The
heat exchangers are now scheduled for cleaning in May 1999,'more than two years after
initial identification.
CR 98-2088 (6/1 2/98) dealt with a control rod that was not positioned as required during
a control rod startup sequence.
One of the assigned actions was to form a team to
develop a Reactivity Management Improvement Plan. The CR file indicated that this
action was transferred to another similar CR (98-2267) under which it would be
implemented.
Upon review, CR 98-2267 was not particularly similar to CR 98-2088, i.e.,
it was primarily a maintenance
item with no human performance issues.
The referred
action was not located in CR 98-2267.
The OES Supervisor indicated that, to his
recollection, the Reactivity Management Improvement Plan had been developed and that
he would review it to determine why it was not appropriately captured in the CR process.
CR 87725 (12/30/98) identified four (4) Level 3 CRs that had been closed based on a
fifth CR.
The fifth CRwas closedinfavorofa
"trend" CR. The ESAP ofthe trend CR
did not address the specific issues for the prior CRs. The issues were primarily
administrative items dealing with material traceability, use of procedure forms, and
quality control notification.
Safety relief valve (SRV) acou'stic monitor problems:
Numerous CRs were written in 1998 regarding continuing problems with the SRV
acoustic monitors. While one of these CRs identified that corrective actions were not
preventing failures of the sensor cable connectors, additional problems persisted.
More
recent evaluations, including a containment temperature study, have found that higher
drywell temperatures
near the SRV acoustic monitors have a significant affect on the
expected life of electrical components
in the area.
Allof the SSES Unit 1 SRV acoustic
monitor accelerometers
were replaced in July 1998, and efforts are still on progress to
determine qualified life based on the new containment temperature profiles. While it
appears that this matter may be approaching final resolution, related issues have been
raised in the CR process since the 1994-1995 time frame. Since that time, this matter
has resulted in three Technical Specification required shutdowns and three Notices of
It does not appear that the CR process was effective in
'esolving this issue in a timely manner.
Emergency diesel non-safety related pneumatic control system problems:
NRC Inspection Report 50-3878388/99-01
documented recent problems with the "B"
EDG non-safety-related
pneumatic control system.
Briefly, on January 26, 1999, during
operability testing of the "B" EDG, an automatic cooldown run was not competed due to
a failed diaphragm on a pressure switch. The failure of the diaphragm admitted carbon
into the pneumatic control system, causing a subsequent
problem with the cooldown run
15
due to clogging of a check valve. The following day, the "B" EDG was taken out of
service for additional work on the pneumatic control system.
Upon reconnection of the
"E" EDG to place it back into service, it shutdown prior to competing its cooldown run
and was declared inoperable.
The"E" EDG was returned service, tested and declared
operable late on January 28, 12999.
A review of the CR database
found numerous (8) prior CRs related to problems with the
non-safety-related
pneumatic control system.
On January 29, 1999, a "trend" CR was
initiated. However, the events of January 26, 1999, demonstrate that prior corrective
actions were neither effective or timely.
Recent overcurrent protection failures
On January 2, 1999, due to overcurrent protection failure, a normal feeder breaker
tripped, causing automatic startup of several pumps ('A'tator cooling, 'A'lectro-
hydraulic control system pump, standby and emergency lube oil pumps for the reactor
feed pump turbine and seal oil system).
The loads were stripped and the breaker re-
closed.
Some loads were not restored pending an evaluation by electrical maintenance.
CR 87595 was initiated. This was the fifth similar failure in a one month period.
Before
this failure, a "trend" CR (82734) had been written to document the previous four
failures. More accelerated corrective action would appear to have been prudent in this
instance.
c.
Conclusions
The CR process is a high volume, low threshold corrective action system that is
acceptable to meet the requirements of 10 CR 50, Appendix B, Criterion XVI. Adverse
conditions are promptly identified and the process appears to be widely accepted for
use, and is used by a broad cross section of the plant staff. There are no significant
delays in the initial assessment,
investigation, and completion of initial operability and
reportability determinations.
Initial investigations and root cause assessments
for issues
of higher safety significance are thorough.
Investigations and causal analyses for CRs of
lesser safety significance are generally thorough, but there are some instances where
extent of conditions and/or generic implications may not be sufficiently explored.
The CR process is focused on accomplishing initial reviews, reportability and operability
determinations, cause assessments,
and establishment of proposed corrective actions to
correct the condition and prevent recurrence.
However, process accountability is not
readily apparent in the corrective action implementation portion of the process.
Corrective action implementation dates do not correlate with the assigned significance
level. Action due dates are controlled by the responsible manager who may change
them during the course of implementation. The only procedural requirement regarding
corrective action implementation is refueling outage related with the shortest lead time
being approximately 7-10 months, and the longest lead time being approximately 31-37
months.
The minimal administrative control over the corrective action implementation
portion of the process contributes to a high process backlog.
16
Internal reviews by OES and external reviews by NAS, CMAP, INPO, and SRC are
continuing to find incomplete or inadequate corrective action closure. A recent process
change requiring OES to review competed actions to ensure that the actions satisfies the
one prescribed, is reducing the amount of items identified. However, problems persist in
this area.
Annual assessments
of the CR process by NAS were thorough, critical, and well
founded.
Past assessments
(1997 and 1998) have identified most of the concerns noted
above.
V. Mana ementNleetin
s
X1
Exit Meeting Summary
The inspectors discussed the inspection results with members of PP8 L management during the
inspection on January 22 and 29, 1999. A formal inspection exit was conducted
February 19, 1999.
The inspectors asked PP&L whether any materials examined during the inspection should be
considered proprietary.
No proprietary information was identified.
PARTIALLIST OF PERSONNEL
CONTACTED'enns
lvania Power and Li ht
R. Saunders, Vice President of Site Operations
G. Kuczynski, General Manager, SSES
'. Male, Manager of Nuclear Assurance
T. Iorfida, Manager of Special Projects
A. Dominguez, Employee Concerns
R. Wehry, Supervisory Licensing Engineer
M. Golden, Nuclear Systems Engineering Supervisor
J. Maertz, Nuclear Systems Engineering Supervisor
S. Ellis, Licensing Engineer
J. Akus, System Analyst
U.S. Nuclear Re ulato
Commission
S. Hansell, Senior Resident Inspector
J. Richmond, Resident Inspector
A. Blarney, Resident Inspector
17
INSPECTION PROCEDURES USED
Tl 2515/139
Inspection of Licensee's Implementation of Generic Letter 96-01 Testing of
Safety Related Logic Circuits
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing
Problems
~Oened
None
ITEMS OPENED, CLOSED, AND DISCUSSED
Closed
None
LIST OF ACRONYMS USED
CAT
CMAP
CR
ESAP
ECO
GL
HPES
IERP
IKM
LER
M8TE
M-G
NAS
NDAP
NIMS
NRC
Corrective Action Team
Cooperative Management Audit Program
Condition Report
Design Change Package
Evaluation and Action Plan
Engineering Change Order
Employee Concerns Program
Electro-Hydraulic Control System
Final Safety Analysis Report
Generic Letter
Human Performance Evaluation System
Industry Events Review Program
Installation KickoffMeeting
Institute of Nuclear Power Operators
Improved Technical Specification
Licensing Document Change Notice
Licensee Event Report
Loss-of-Coolant Accident
Loss of Offsite Power
Measuring and Test Equipment
Motor Generator
Nuclear Assurance Services
Nuclear Department Administrative Procedure
Nuclear Information Management System
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
18
PP8L
Operating Experience Services
Plant Modification Package
Plant Operations Review Committee
Pennsylvania Power and Light
Quality Assurance
Susquehanna
Review Committee
Susquehanna
Steam Electric Station
C7
0