ML080770308
| ML080770308 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/17/2008 |
| From: | Mel Gray Division Reactor Projects I |
| To: | Mckinney B Susquehanna |
| Gray M, RI/DRP/TSAB/610-337-5209 | |
| References | |
| IR-08-006 | |
| Download: ML080770308 (29) | |
See also: IR 05000387/2008006
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
475 ALLENDALE ROAD
KING OF PRUSSIA, PA 19406-1415
March 17, 2008
Mr. Britt T. McKinney
Senior Vice President and Chief Nuclear Officer
769 Salem Blvd. - NUCSB3
Berwick, PA 18603-0467
SUBJECT: SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2
PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION
INSPECTION REPORTS NOS. 05000387/2008006; 05000388/2008006
Dear Mr. McKinney:
On February 1, 2008, the US Nuclear Regulatory Commission (NRC) completed a team
inspection at the Susquehanna Steam Electric Station. The enclosed inspection report
documents the inspection results, which were discussed on February 1, 2008, with you and
members of your staff.
This inspection was an examination of activities conducted under your license as they relate to
the identification and resolution of problems, and compliance with the Commission=s rules and
regulations and the conditions of your license. Within these areas, the inspection involved
examination of selected procedures and representative records, observations of activities, and
interviews with personnel.
On the basis of the sample selected for review, the team concluded that the implementation of
the corrective action program (CAP) was adequate in that personnel identified issues at a low
threshold; generally screened and prioritized issues in a timely manner; evaluated the issues
commensurate with their safety significance; and implemented corrective actions in a timely
manner commensurate with the safety significance.
The team identified four findings of very low safety significance (Green). These findings were
determined to involve violations of regulatory requirements. However, because each of the
violations was of very low safety significance (Green) and because they were entered into your
corrective action program, the NRC is treating these as Non-Cited Violations (NCVs), in
accordance with Section VI.A.1 of the NRC=s Enforcement Policy. If you contest any NCV in
this report, you should provide a response within 30 days of the date of this inspection report,
with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington DC, 20555-0001, with copies to the Regional Administrator, Region I;
B. McKinney 2
the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC,
20555-0001; and the NRC Resident Inspector at the Susquehanna facility.
In accordance with 10 CFR 2.390 of the NRC=s ARules of Practice,@ a copy of this letter and its
enclosure, and your response (if any), will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRC=s document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mel Gray, Chief
Technical Support & Assessment Branch
Division of Reactor Projects
Docket Nos. 50-387, 50-388
Enclosure: Inspection Report Nos. 05000387/2008006; 05000388/2008006
w/ Attachment: Supplemental Information
cc w/encl:
C. Gannon, Vice President, Nuclear Operations
R. Paley, General Manager, Plant Support
R. Pagodin, General Manager, Nuclear Engineering
R. Sgarro, Manager, Nuclear Regulatory Affairs
Supervisor, Nuclear Regulatory Affairs
M. Crowthers, Supervising Engineer, Nuclear Regulatory Affairs
R. Peal, Mgr, Training, Susquehanna
Manager, Quality Assurance
J. Scopelliti, Community Relations Manager, Susquehanna
B. Snapp, Esq., Associate General Counsel, PPL Services Corporation
Supervisor - Document Control Services
R. Osborne, Allegheny Electric Cooperative, Inc.
D. Allard, Dir, PA Dept of Environmental Protection
Board of Supervisors, Salem Township
J. Johnsrud, National Energy Committee, Sierra Club
E. Epstein, TMI-Alert (TMIA)
J. Powers, Dir, PA Office of Homeland Security
R. French, Dir, PA Emergency Management Agency
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-387, 50-388
Report No: 05000387/2008006, 05000388/2008006
Licensee: PPL Susquehanna, LLC
Facility: Susquehanna Steam Electric Station, Units 1 and 2
Location: 769 Salem Boulevard - NUCSB3
Berwick, PA 18603-0467
Dates: January 14 - February 1, 2008
Team Leader: B. Norris, Senior Project Engineer, Division of Reactor Projects
Inspectors: F. Arner, Senior Reactor Inspector, Division of Reactor Safety
R. Fuhrmeister, Senior Project Engineer, Division of Reactor Projects
G. Ottenberg, Resident Inspector, Division of Reactor Projects
J. Bream, Reactor Engineer, Division of Reactor Projects
R. McKinley, Operations Examiner, Division of Reactor Safety
Approved by: Mel Gray, Chief
Technical Support & Assessment Branch
Division of Reactor Projects
Enclosure
2
SUMMARY OF FINDINGS
IR 05000387/2008-006, 05000388/2008-006; 01/14/2008 - 02/01/2008; Susquehanna Steam
Electric Station; Biennial Baseline Inspection of the Identification and Resolution of Problems;
Corrective Action Program, Simulator Fidelity, and Procedure Quality.
This team inspection was performed by five NRC regional inspectors and one resident
inspector. Four findings of very low safety significance (Green) were identified during this
inspection and determined to be Non-Cited Violations (NCVs). The significance of most findings
is indicated by their color (Green, White, Yellow, Red) using NRC Inspection Manual Chapter (IMC) 0609, ASignificance Determination Process@ (SDP). The NRC=s program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649, AReactor
Oversight Process,@ Revision 4, dated December 2006.
Identification and Resolution of Problems
The team concluded that the implementation of the corrective action program (CAP) at
Susquehanna was adequate in that personnel identified issues at a low threshold and used a
single entry-point system to document the problems by the initiation of an Action Request (AR).
About 20 percent of the ARs were considered to be conditions adverse to quality (CAQ) and
sub-classified as a Condition Report (CR). However, the team identified several ARs that
should have been classified as CAQs; as a result, CRs were not written and corrective actions
were not timely. The team identified two findings of very low significance related to the AR
process that had current performance cross-cutting aspects in problem identification because
the issues were not categorized commensurate with their safety significance. Notwithstanding
these two findings, the team concluded that in general Susquehanna personnel screened and
prioritized CRs in a timely manner using established criteria.
The team also concluded that Susquehanna personnel properly evaluated the issues
commensurate with their safety significance; and generally implemented corrective actions in a
timely manner, commensurate with the safety significance. The team noted that Susquehanna
reviewed and applied industry operating experience lessons learned. Audits and self-
assessments added value to the corrective action process. On the basis of interviews
conducted during the inspection, workers at the site expressed freedom to enter safety
concerns into the CAP.
Enclosure
3
a. NRC Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
C Green: The NRC identified a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V,
Instructions, Procedures, and Drawings, because, in the 1990s, Susquehanna failed to
adequately evaluate a deviation from the Boiling Water Reactor Owners Group
Emergency Procedure Guidelines / Severe Accident Guidelines (BWROG EPG/SAG),
which resulted in one of the emergency operating procedures (EOPs) being inadequate.
Specifically, Caution #1 in the BWROG EPG/SAG warned the operators that reactor
pressure vessel (RPV) level instrumentation may be unreliable if the drywell
temperatures exceeded RPV saturation temperature. The purpose of the Caution was
to give the operators a chance to evaluate the validity of the RPV level instrumentation
to avoid premature entry into the RPV flooding contingency procedure. Susquehanna
did not adequately evaluate the deviation, and the Susquehanna EOPs did not use a
Caution statement; but instead, changed the caution to a procedural step, which directed
the operators to transition directly to the RPV flooding procedure.
The performance deficiency is more than minor because it is associated with the
Procedure Quality attribute of the Mitigating Systems cornerstone and affects the
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, the EOP could have
directed entry into the RPV flooding procedure unnecessarily which would have
restricted the use of suppression pool cooling and required other actions that would have
complicated the operators response to the event. The finding was determined to be of
very low safety significance because it was not a design deficiency, did not result in an
actual loss of safety function, and did not screen as potentially risk significant due to
external initiating events. (Section 4OA2.a.3 (a))
C Green: The NRC identified a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion
XVI, Corrective Action, for the failure to identify that an inconsistency between the
procedures and the design basis for suppression pool (SP) cooling was a condition
adverse to quality (CAQ), which resulted in corrective actions not being taken in a timely
manner. Specifically, in January 2006, a Condition Report (CR) identified an
inconsistency between an assumption in the Final Safety Analysis Report (FSAR) for the
design basis accident and the emergency operating procedures (EOPs) regarding the
timing for the implementation of SP cooling. At the time of the inspection, the
inconsistency had not been resolved because Susquehanna did not recognize that it
impacted current plant operations. This performance deficiency has a cross-cutting
aspect in the area of Problem Identification and Resolution, Corrective Action Program,
because Susquehanna did not identify that the inconsistency documented in the CR
should have been categorized as a CAQ, commensurate with its safety significance.
The performance deficiency is more than minor because it is associated with the Design
Control attribute of Mitigating Systems and affects the cornerstone objective to ensure
the availability, reliability, and capability of systems that respond to initiating events to
Enclosure
4
prevent undesirable consequences. Specifically, the EOPs provided direction that,
under some accident conditions, would affect the availability and/or capability of the SP
cooling system to perform its safety function. The finding screened out as having very
low safety significance because it was not a design deficiency, did not result in an actual
loss of safety function, and did not screen as potentially risk significant due to external
initiating events. (Section 4OA2.a.3 (b))
C Green: The NRC identified a Non-Cited Violation of 10 CFR 55.46(c)(1), Plant
Referenced Simulators, because the Susquehanna simulator did not accurately model
reactor pressure vessel (RPV) level instrumentation following a design basis accident
loss of coolant accident (DBA LOCA). Specifically, an analysis performed in 1994 to
determine if the observed simulator response during a large break LOCA was consistent
with the expected plant response, was based on an overly conservative assumption that
the drywell would experience superheated conditions, which would cause RPV water
level instrumentation reference leg flashing and a subsequent loss of all RPV level
indication. The expected plant response, as stated in the analysis, was incorrect; in that
a LOCA would not always cause a loss of all RPV level instruments. As a result, the
simulator modeling was incorrect.
The performance deficiency is more than minor because it is associated with the Human
Performance attribute of Mitigating Systems and affects the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. Specifically, the modeling of the
Susquehanna simulator introduced negative operator training that could affect the ability
of the operators (a mitigating system) to take the appropriate actions during an actual
event. The finding was determined to be of very low safety significance because it is not
related to operator performance during requalification, it is related to simulator fidelity,
and it could have a negative impact on operator actions. (Section 4OA2.a.3 (c))
C Green: The NRC identified a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion
XVI, Corrective Action, for the failure to identify that a setpoint error in the operating
procedures for safety-related systems was a condition adverse to quality (CAQ),
resulting in the procedures not being corrected in a timely manner. The setpoint for the
low pressure injection permissive interlock in the RHR and CS systems had been
changed in 1999 as part of a modification. However, the setpoint was not changed in
the system operating procedures and operator aids. When this issue was identified by
Susquehanna staff in 2006, the setpoint error in the procedure was not screened as a
CAQ, which resulted in the procedures not being revised for 17 months after the issue
was identified in an Action Report. This performance deficiency has a cross-cutting
aspect in the area of Problem Identification and Resolution, Corrective Action Program,
because Susquehanna did not identify that a setpoint error in operating procedures for
safety-related systems was a CAQ, commensurate with its safety significance. P.1(a)
The performance deficiency is more than minor because it is associated with the
Procedure Quality attribute of Mitigating Systems and affects the cornerstone objective
to ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. Specifically, the incorrect setpoint
Enclosure
5
reference in the procedure impacted the reliability of operator response to the event in
that it could delay operator actions or result in misoperation of equipment. The finding
screened out as having very low safety significance because it was not a design
deficiency, did not result in an actual loss of safety function, and did not screen as
potentially risk significant due to external initiating events. (Section 4OA2.a.3 (e))
b. Licensee-Identified Violations
None.
Enclosure
6
REPORT DETAILS
4. OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution (PI&R) (Biennial - IP 71152B)
a. Assessment of the Corrective Action Program
1. Inspection Scope
The inspection team reviewed the procedures describing the corrective action program
(CAP) at the Susquehanna Steam Electric Station. Susquehanna used a single-point
entry system and identified problems by the initiation of an Action Request (AR). The
AR would then be sub-classified depending on the information provided; for example, as
WO for a maintenance Work Order, as CPG for assignment to the Central Procedure
Group, or as CR for a Condition Report. ARs were sub-classified as CRs for conditions
adverse to quality (CAQ), such as plant equipment deficiencies, industrial or radiological
safety concerns, or other significant issues. The CRs were subsequently screened for
operability and reportability, categorized by significance (1 to 3), assigned a level of
evaluation, and issued for resolution.
The team reviewed CRs selected across the seven cornerstones of safety in the NRC=s
Reactor Oversight Process (ROP) to determine if problems were being properly
identified, characterized, and entered into the CAP for evaluation and resolution. The
team selected items from the maintenance, operations, engineering, emergency
preparedness, physical security, radiation safety, training, and oversight programs to
ensure that Susquehanna was appropriately considering problems identified in each
functional area. The team used this information to select a risk-informed sample of CRs
that had been issued since the last NRC PI&R inspection, which was conducted in
February 2006.
The team selected ARs from other sub-classifications, to determine if Susquehanna had
appropriately classified these items as not needing to be a CR. The team also reviewed
operator log entries, control room deficiency lists, operator work-around lists, operability
determinations, engineering system health reports, completed surveillance tests, and
current temporary configuration change packages. In addition, the team interviewed
plant staff and management to determine their understanding of and involvement with
the CAP at Susquehanna. The CRs, and other documents reviewed, and the key
personnel contacted, are listed in the Attachment to this report.
The team considered risk insights from the NRC=s and Susquehanna=s risk analyses to
focus the sample selection and plant tours on risk-significant components. The team
determined that the five highest risk-significant systems at Susquehanna were
emergency service water, emergency diesel generators, residual heat removal service
water, station black-out diesel generator, and reactor core isolation cooling. For the
risk-significant systems, the team reviewed a sample of the applicable system health
Enclosure
7
reports, work requests and engineering documents, plant log entries, and results from
surveillance tests and maintenance tasks.
The team reviewed CRs to assess whether Susquehanna adequately evaluated and
prioritized the identified problems. The CRs reviewed encompassed the full range of
Susquehanna=s causal evaluations, including root cause analyses (RCA - to determine
the cause and prevent recurrence), apparent cause evaluations (ACE - to obtain a basic
understanding of the cause), and evaluations (to determine if a problem exists). The
review included the appropriateness of the assigned significance, the scope and depth
of the causal analysis, and the timeliness of the resolutions. For significant conditions
adverse to quality, the team reviewed the effectiveness of the corrective actions to
prevent recurrence. The team observed meetings of the CR Screening Team - in which
Susquehanna personnel reviewed new CRs for prioritization, and evaluated preliminary
corrective action assignments, analyses, and plans. The team also attended meetings
of the Corrective Action Review Board (CARB) - where senior managers reviewed
selected evaluations, effectiveness reviews, and extension requests.
The team reviewed equipment operability determinations, reportability assessments, and
extent-of-condition reviews for selected problems. The team assessed the backlog of
corrective actions in the maintenance, engineering, and operations departments, to
determine, individually and collectively, if there was an increased risk due to delays in
implementation of corrective actions. The team further reviewed equipment
performance results and assessments documented in completed surveillance
procedures, operator log entries, and trend data to determine whether the evaluations
were technically adequate to identify degrading or non-conforming equipment.
The team reviewed the corrective actions associated with selected CRs to determine if
the actions addressed the identified causes of the problems. The team reviewed CRs
for significant repetitive problems to determine if previous corrective actions were
effective. The team also reviewed Susquehanna=s timeliness in implementing corrective
actions. The team reviewed the CRs associated with selected non-cited violations
(NCVs) and findings to determine if Susquehanna properly evaluated and resolved these
issues.
2. Assessment
(a) Identification of Issues
In general, the team considered the identification of equipment deficiencies at
Susquehanna to be adequate. There was a low threshold for the identification of
individual issues, 23,000 ARs were written per year, and about 4,000 of those were
sub-classified as CRs. The housekeeping and cleanliness of the plant was generally
good; the general cleanliness of the plant enhanced the ability of personnel to more
easily identify equipment deficiencies and monitor equipment for worsening conditions.
Notwithstanding, during a tour of the facility, the inspectors observed that high density
concrete shield blocks were stacked on pallets in the vicinity of the Unit 1 recirculation
Enclosure
8
motor generator sets. The blocks were pre-staged for work during the upcoming
refueling outage, and were in a heavily trafficked area of the turbine building. There was
a painted warning on the floor, near the pallets, that the floor loading should not exceed
400 pounds per square foot (psf). When the inspectors asked whether the weight of the
blocks was within the rated floor load limit, it was determined that this condition had not
been identified and documented as acceptable. Initially, Susquehanna personnel
concluded that the blocks exceeded the posted limit and moved the pallets to reduce the
floor loading. Subsequently, Susquehanna weighed the pallets and blocks and
determined that they did not exceed the allowable floor loading. Based on this
evaluation the inspectors concluded the missed identification of this issue was minor.
The issue was documented in CR 954950.
The team also identified that several ARs were not classified as CRs, commensurate
with the safety significance, as required by their procedure (NDAP-QA-0702, Action
Request and Condition Report Process). The result was that the issues did not go to
the Screening Team, did not receive the necessary management attention, and were not
corrected in a timely manner (CR 957319). In addition, ARs are not normally trended to
allow the identification of an adverse change in performance. With the exception of the
first example, the below are considered procedure violations of minor significance due to
no impact on the related equipment. As such, these issues are not subject to
enforcement action, in accordance with the NRC=s Enforcement Policy.
Examples include:
C AR/CPG/OPS 751412, initiated February 2, 2006, identified that the Low Pressure
Injection Permissive setpoint was not changed in the residual heat removal (RHR)
and core spray (CS) operating procedures. The setpoint was changed in 1999, as
part of a modification; the procedures were not changed until July 2007. (See
Section 4OA2.a.3(d) for additional details.)
C AR/OPS/CSHIFT 777335, initiated July 25 2006, identified that an operator started
the suppression pool (SP) filter pump contrary to the procedure. The AR was closed
with no documented corrective actions taken.
The safety significance is that the operator did not operate the safety-related system
in accordance with the licensees written procedures and the Technical
Specifications (TS). The documentation of corrective actions should have included a
determination of the affects of starting of the pump, and counseling of the operator
on the requirement to follow procedures.
C AR/CPG 810513, initiated September 16, 2006, identified that the wrong valve
numbers were listed for the emergency service water (ESW) system valves for the
E EDG. As of the inspection, the procedure had not been changed.
The safety significance is that operators may not have been able to use the
licensees written procedure to align the ESW system in support of the operation of
the swing E EDG in a timely manner.
Enclosure
9
C AR/CPG/I&C 938054, initiated December 10, 2007, identified that a functional testing
and calibration procedure for the RHR service water radiation monitor could not be
performed, as written. As of the inspection, corrective actions had not been taken.
an inconsistency between the procedures and the design basis for SP cooling was a
CAQ, which resulted in corrective actions not being taken for two years to the time of the
inspection. Although the inconsistency was identified in 2006, Susquehanna personnel
did not recognize that the issue impacted current plant operations; as a result, the issue
was not scheduled for resolution in a timely manner. The team noted that, although
Susquehanna had classified the issue as a CR, it was considered to be NAQ - not a
CAQ - and was not scheduled for evaluation until the EPU had been approved. Refer to
Section 4OA2.a.3(b) for a detailed discussion of the finding.
(b) Prioritization and Evaluation of Issues
The team determined that Susquehannas performance in this area was adequate.
Notwithstanding the above discussion of some ARs not being classified as CRs, the
station appropriately reviewed those CRs that went to the Screening team and properly
classified them for significance. The discussions about specific topics at the Screening
meetings were detailed, and there were no classifications or immediate operability
determinations with which the team disagreed. The team considered the contributions of
the CARB to add value to the CAP process. One CARB review was noted to be
particularly insightful with respect to the quality of the causal analysis for CR 773046.
The CR identified problems with the closing of CRs by the nuclear training department
without completing all the required actions. The team did not identify any items in the
operations, engineering, or maintenance backlogs that were risk significant, individually
or collectively. In addition, the quality of the causal analyses reviewed was generally of
adequate technical detail and scope to identify causal factors and develop effective
corrective actions. The team noted that the RCA for the NCV from the last PI&R
inspection related to scaffolding was effective in that there had not been significant
recurrences of inadequate scaffold installations since the evaluation was completed.
With regard to operability evaluations, the team observed that, an operability
determination for the PAM level instruments, conducted in response to an inconsistency
between the FSAR and EOPs, determined that the level instruments would be operable.
(The inconsistency between the FSAR and the EOPs is described in detail in section
4OA2.a.3(b).) During follow-up discussions, the inspectors were told by operations and
engineering personnel that all of the PAM instrumentation together functioned to provide
the needed indications to the operators, and that the RPV level indications were not
needed after the initial entry into the EOPs. This was not consistent with the
requirements for the operability of each individual function of the PAM, as detailed in TS 3.3.3.1. Although subsequent discussions with the Susquehanna staff determined that
the most (if not all) of the PAM RPV level instruments would indicate post-LOCA, the
initial operability determination and statements during the inspection did not consider
that the PAM level instruments are required to be operable post-accident regardless of
whether EOPs have been entered. This issue was related to the performance
Enclosure
10
deficiencies discussed in findings 4OA2.a.3(a), (b) and (c), and is not identified as an
additional finding. The issue was entered into the CAP as AR/CR964836.
(c) Effectiveness of Corrective Actions
No findings of significance were identified in the area of effectiveness of corrective
actions. The team determined that the effectiveness of corrective actions at
Susquehanna was generally good. The control of scaffolds was a significant problem
during the last PI&R inspection; the team noted that oversight of scaffolds has improved,
but station personnel continue to identify examples where the scaffold does not appear
to be built in accordance with the procedure. In addition, the team identified
weaknesses in the scaffold procedure, such as allowing the installer to approve
deviations from the approved construction. During the inspection, the procedure was
revised, and plans were developed for engineering to review all current deviations.
3. Findings
(a) Failure to Adequately Evaluate a Deviation from BWROG EPG/SAG Resulted in an
Inadequate Procedure
Introduction: The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V,
Instructions, Procedures, and Drawings, because Susquehanna failed to adequately
evaluate a deviation from the Boiling Water Reactor Owners Group Emergency
Procedure Guidelines / Severe Accident Guidelines (BWROG EPG/SAG), which
resulted in one of the Emergency Operating Procedures (EOPs) being inadequate.
Description: On January 5, 2006, AR/CR 739371 was initiated to document an
inconsistency between the EOPs and assumptions in the Final Safety Analysis Report
(FSAR) regarding the initiation of suppression pool cooling. Specifically, it was identified
that the assumptions used in evaluating SP temperature response for the most limiting
design basis accident (DBA) loss of coolant accident (LOCA) did not appear to be
consistent with direction provided in the EOPs.
During this inspection, the team noted that the Susquehanna EOPs were not consistent
with the BWROG EPG/SAG. Specifically, BWROG EPG/SAG, Revision 2, Caution #1,
warned the operators that reactor pressure vessel (RPV) level instrumentation may be
unreliable if the temperatures near the instrument sensing lines exceeded RPV
saturation temperature. The EPG Bases stated that the purpose of Caution #1 was to
give the operators a chance to evaluate the validity of the RPV level instrumentation, in
order to avoid premature entry into the RPV flooding contingency procedure before it
was appropriate to do so. Susquehanna did not adequately evaluate the deviation from
the generic guidance in the EPG/SAG with respect to the caution. The Susquehanna
EOPs did not use a Caution statement, which would have allowed the operators the
opportunity to evaluate the level instrumentation; but instead, changed the caution to a
procedural step which directed the operators to transition directly to the RPV Flooding
procedure. Specifically, EO-100-103-1, Primary Containment Cooling, step DWT-3,
Enclosure
11
directed the operators to transition to contingency procedure EO-000-114-1, RPV
Flooding, when drywell temperature exceeded RPV saturation temperature.
The evaluation for the deviation was not completed in accordance with the requirements
of procedure NDAP-QA-0330, Symptom Oriented EOP and EP-DS Program and
Writers Guide. The procedure required that all deviations be evaluated to determine if
the deviation was technically justified and appropriate. Susquehanna documented that
the deviation was a minor difference from the generic guidelines in 50.59 Safety
Evaluation NL-92-019 (October 29, 1998) and 50.59 Screen 5059-01-976 (July 3, 2002).
The evaluation was based on an overly conservative assumption that all RPV level
instrumentation would be lost after a DBA LOCA. The reviews did not evaluate the
potential adverse consequences associated with the deviation, including the potential
impact on the SP cooling safety function. Immediate corrective actions included the
initiation of an informational Night Order to the control room operators explaining the
issue, and the cessation of all simulator scenarios that involve the use of EO-100-103-1
until the issue is resolved.
The performance deficiency is the failure to adequately evaluate a deviation from the
BWROG EPG/SAG, which resulted in one of the EOPs being inadequate for use by the
operators in the event of a DBA LOCA. Specifically, under some accident conditions,
the EOPs would have unnecessarily directed entry into RPV flooding which would have
limited the availability of SP cooling and complicated the operators response to the
event.
Analyses: This performance deficiency is more than minor because it is associated with
the Procedure Quality (EOP) attribute of the Mitigating Systems cornerstone and affects
the objective to ensure the availability, reliability, and capability of systems that respond
to initiating events to prevent undesirable consequences. Specifically, the EOP could
have directed entry into the RPV flooding procedure unnecessarily which would have
restricted the use of suppression pool cooling and required other actions that would have
complicated the operators response to the event. The inspectors performed a review of
the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609,
Significance Determination Process (SDP), Attachment 4, Phase 1 - Initial Screening
and Characterization of Findings, and determined that the finding screened out as
having very low safety significance (Green), because it was not a design deficiency, did
not result in an actual loss of safety function, and did not screen as potentially risk
significant due to external initiating events.
Enforcement: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, states, in part, that activities affecting quality shall be prescribed by
documented procedures appropriate to the circumstances and that the activities shall be
accomplished in accordance with the procedures. Contrary to the above, Emergency
Operating Procedure EO-100-103-1, Primary Containment Cooling, was inadequate, in
that it directed the operators to transition directly to the RPV Flooding procedure when
RPV level instruments may have been available, which resulted in limiting the availability
of SP cooling. However, because the finding was of very low safety significance (Green)
Enclosure
12
and has been entered into the CAP (AR/CR 962881), this violation is being treated as an
NCV, consistent with section VI.A.1 of the NRC Enforcement Policy.
(NCV 05000387/2008006-01; 05000388/2008006-01 - Failure to Adequately Evaluate
a Deviation from BWROG EPG/SAG Resulted in an Inadequate EOP)
(b) Failure to Identify and Correct Inconsistencies Between the FSAR and the EOPs
Introduction: The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, for the failure to identify that an inconsistency between the
emergency operating procedures and the design basis for SP cooling was a CAQ, which
resulted in corrective actions not being taken for two years to the time of the inspection.
Although the inconsistency was identified in 2006, Susquehanna personnel did not
recognize that the issue impacted current plant operations; as a result, the issue was not
scheduled for resolution in a timely manner. The assumption in the FSAR for the DBA
LOCA stated that SP cooling would be implemented ten minutes after entry into the
EOPs. The EOPs would not have allowed initiation of SP cooling for an extended period
of time.
Description: On January 5, 2006, AR/CR 739371 was initiated to document an
inconsistency between the EOPs and design basis assumptions for the SP cooling
response. The problem was identified during Susquehannas review in support of the
extended power uprate (EPU) project. Specifically, Susquehanna Engineering identified
that the assumptions used in evaluating SP temperature response for the most limiting
LOCA did not appear to be consistent with direction provided in the EOPs. The team
noted that, although Susquehanna personnel had classified the issue as a CR, they did
not recognize that the issue impacted current plant operations. Therefore, it was
considered to be NAQ - not a condition adverse to quality - and was not scheduled for
evaluation until the EPU had been approved.
The Susquehanna FSAR, Section 6.2.1.1.3, stated that the maximum SP temperature
would result from a reactor recirculation suction line break. The drywell pressure and
temperature response analyses assumed that RHR heat exchangers were activated
about ten minutes after entry into the EOPs to remove energy from the drywell by
cooling the SP. The CR identified that, in the event of a DBA LOCA, the EOPs would
direct operators to implement the RPV flooding procedure (EO-000-114) to maintain
adequate core cooling, and this required that all available RHR flow be used to flood the
RPV up to the steam lines. The initiators concern was that this would delay establishing
flow through a RHR heat exchanger for SP cooling, because of the unique design of the
RHR system at Susquehanna, and therefore would be inconsistent with the accident
analyses assumptions. In addition, the CR stated that it was assumed in the EOPs that
all RPV water level indications would be unreliable and therefore unavailable for this
scenario. Susquehanna personnel informed the team that they had not evaluated the
issues documented in the CR, at the time it was initiated, because they had assumed
that they were only associated with EPU and not current plant operation. Immediate
corrective actions included the start of an evaluation during the inspection of the
identified inconsistency for SP cooling, and additional guidance to the operators.
Enclosure
13
The performance deficiency is the failure to properly categorize the inconsistency
between the FSAR and the EOPs as a CAQ, which resulted in the deficiency not being
corrected in a timely manner commensurate with its safety significance.
Analyses: The performance deficiency is more than minor because it is associated with
the Design Control attribute of the Mitigating Systems cornerstone and affects the
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, in the event of a
DBA LOCA, SP cooling would not be initiated within the time frame assumed in the
FSAR, which could affect the capability of the system to perform its safety function
consistent with the design basis. The inspectors performed a review of the finding in
accordance with IMC 0609, and determined that the finding screened out as having very
low safety significance (Green) because it was not a design deficiency, did not result in
an actual loss of safety function, and did not screen as potentially risk significant due to
external initiating events.
This performance deficiency has a cross-cutting aspect in the area of Problem
Identification and Resolution (PI&R), Corrective Action Program (CAP), because
Susquehanna did not identify that the inconsistency documented in the CR should have
been categorized as a CAQ, commensurate with its safety significance. P.1(a)
Enforcement: 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, states, in part,
that conditions adverse to quality shall be promptly identified and corrected. Contrary to
the above, Susquehanna failed to identify that the nonconformance identified in AR/CR
739371, January 2006, was a CAQ; this resulted in the condition not being corrected for
over two years. However, because the finding was of very low safety significance
(Green) and has been entered into the corrective action program (AR/CR 959670), this
violation is being treated as an NCV, consistent with section VI.A.1 of the NRC
(NCV 05000387/2008006-02; 05000388/2008006-02 - Failure to Identify and Correct
Inconsistencies Between the FSAR and the EOPs)
(c) Failure to Accurately Model the Simulator for RPV Water Level Instrumentation
Introduction: The NRC identified a Green NCV of 10 CFR 55.46(c)(1), Plant
Referenced Simulators, because the Susquehanna plant-referenced simulator did not
accurately model RPV level instrument response following a DBA LOCA. Specifically,
the RPV level instruments in the simulator were programmed to fail high after a LOCA,
and the expected plant response is that the instruments should indicate properly.
Description: As part of the teams follow-up on the issues in AR/CR 739371, the
inspectors questioned the concern stated in the CR, that the operators would need to
enter the RPV flooding procedure during a DBA LOCA due to a loss of valid RPV level
instrumentation. The inspectors reviewed the Susquehanna specific EOPs and
supporting documents, and determined that the Susquehanna EOP Plant Specific
Enclosure
14
Technical Guideline (PSTG) description of the expected response of the RPV level
instrument response to LOCA events, was based on analysis, EC-SIMU-1001,
Evaluation of Simulator Level Instrument Response to Large LOCA, dated May 4,
1994. The analysis was performed to determine if the observed simulator response
during a large break LOCA (RPV level instrumentation off-scale high) was consistent
with the expected plant response. The analysis assumed that the drywell would
experience superheated conditions, which would cause RPV water level instrumentation
reference leg flashing and a subsequent loss of all RPV level indication. The analysis
concluded that the simulator response reasonably predicted the expected actual plant
response during a large break LOCA event. The expected plant response, as stated in
the analysis, was incorrect; in that a LOCA would not always cause a loss of all RPV
level instruments.
On January 29, 2008, the inspectors observed two scenarios in the simulator to evaluate
the response to a DBA LOCA, with all safety systems available. The inspectors
observed that the RPV level instruments did indicate off-scale high shortly after the
initiation of the event, consistent with the analysis. The inspectors questioned the basis
of the analysis; specifically, why Susquehanna believed that the level instruments would
not be available after a DBA LOCA event. Subsequently, Susquehanna determined that
the RPV level instrument reference legs were not expected to routinely flash during a
DBA LOCA, and that the analysis had been based on an overly conservative assumption
that the drywell would always reach superheated conditions post-LOCA. Immediate
corrective actions included the initiation of an informational Night Order to the control
room operators explaining the issue, and the cessation of all simulator scenarios that
involve the use of EO-100-103-1 until the issue is resolved.
The performance deficiency is that Susquehanna did not ensure that the plant
referenced simulator accurately modeled the expected plant response for RPV level
instrumentation after a DBA LOCA, resulting in negative training of the licensed
operators.
Analyses: This performance deficiency is more than minor because it is associated with
the Human Performance attribute of the Mitigating Systems cornerstone and affects the
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, the incorrect
modeling of the Susquehanna plant referenced simulator introduces negative operator
training that could affect the ability of the operators (a mitigating system) to take the
appropriate actions during an actual event. The simulator training conditioned the
operators to expect the level instruments to be unavailable during events that cause
drywell temperatures to reach or exceed RPV saturation temperature. As a result,
during an actual event, the operators could prematurely transition into the RPV flooding
procedure when the RPV level instruments should be providing valid indication. The
inspectors evaluated the finding in accordance with IMC 0609, Appendix I, Licensed
Operator Requalification Significance Determination Process. The finding was
determined to be of very low safety significance (Green) because it is not related to
operator performance during requalification, it is related to simulator fidelity, and could
have a negative impact on operator actions.
Enclosure
15
Enforcement: 10 CFR 55.46(c)(1), Plant Referenced Simulators, states, in part, that a
plant referenced simulator must demonstrate expected plant response to normal,
transient, and accident conditions. Contrary to the above, as of January 2008, the
Susquehanna plant referenced simulator did not accurately demonstrate the actual
expected plant response of the RPV water level instrumentation following a DBA LOCA,
which could result in negative operator training. However, because the finding was of
very low safety significance (Green) and has been entered into the CAP (AR/CR
962881), this violation is being treated as an NCV, consistent with section VI.A.1 of the
(NCV 05000387/2008006-03; 05000388/2008006-03 - Failure to Accurately Model
the Simulator for RPV Water Level Instrumentation)
(d) Failure to Identify and Correct a Setpoint Error in the RHR and CS Operating
Procedures
Introduction: The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, for the failure to identify that a setpoint error in the operating
procedures for safety-related systems was a CAQ, resulting in the procedures not being
corrected in a timely manner. Specifically, in February 2006, Susquehanna personnel
identified an incorrect setpoint for the low pressure injection permissive interlock in the
RHR and CS systems operating procedures and associated hard cards; however, the
procedures were not revised until July 2007 due to the issue being screened as low
priority and not a condition adverse to quality (CAQ).
Description: On February 11, 2006, an AR was written to identify that the low pressure
injection permissive setpoint in the RHR and CS operating procedures, and the
associated operator hard cards, was incorrect. The correct setpoint is 420 pounds per
square inch gage (psig), but the procedures still had the previous setpoint of 436 psig.
The setpoint had been changed in 1999 as part of a modification. The procedures were
not revised until July 16, 2007, 17 months after the deficiency was identified in an AR. In
addition, the inspectors noted that the setpoint in the procedures (436 psig) was not
within the allowable tolerance (407-433 psig) listed in the Susquehanna TS, Section 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation.
When the AR was initiated, it was sub-classified as AR/CPG/OPS; that is, assigned to
the Central Procedures Group and identified as an Operations procedure. It was not
recognized that deficient operating procedures for safety-related systems may be a CAQ
and that the AR should have been classified as a Condition Report. The affected
section in the procedures was the verification of the response of the systems to an
automatic initiation signal. For example, the Unit 1 RHR procedure OP-149-001, RHR
System, Section 2.2, noted that No operator action is required unless an automatic
action failed to occur ... At 436 psig decreasing Reactor pressure, RHR INJ OB ISO
[injection outboard isolation] HV-151-F015A & B OPEN. If the valves did not open at
the specified pressure in the procedure and hard card, the operator may have diverted
their attention unnecessarily and attempted to open the valve manually, even though the
Enclosure
16
interlock would not have been satisfied (420 psig) and the valve would not open in
accordance with the plant design.
The pressure switches were changed in 1999, as part of a Unit 1 plant modification
(Design Change Package (DCP) 97-9075); Unit 2 switches were changed by DCP
97-9076. The modification replaced the existing pressure switches with Barton pressure
indicating switches, because of improved accuracy. The low pressure injection
permissive interlock prevents the CS and RHR injection valves from opening until
reactor pressure has decreased to the RHR and CS systems design pressure, to
prevent over pressurization of the RHR and CS systems. The DCP identified the
specific RHR and CS operating procedures as needing to be changed. Immediate
corrective actions included the initiation of a new CR to evaluate the other pending
procedure changes to determine if their priority should be revised.
The performance deficiency involved a failure to identify and correct a CAQ, the
incorrect setpoint, in a timely manner commensurate with its safety significance. The
inspectors concluded this action was untimely because the modification process would
have revised these procedures prior to the modification being accepted by operations
personnel.
Analysis: The performance deficiency is more than minor because it is associated with
the Procedure Quality attribute of the Mitigating Systems cornerstone and affects the
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, the incorrect
setpoint reference in the procedure impacted the reliability of operator response to the
event in that it could delay operator actions or result in misoperation of equipment. The
inspectors performed a review of the finding in accordance with NRC Inspection Manual
Chapter (IMC) 0609, Significance Determination Process (SDP), Attachment 4, Phase
1 - Initial Screening and Characterization of Findings. The inspectors determined that
the finding screened out as having very low safety significance (Green), because it was
not a design deficiency, did not result in an actual loss of safety function, and did not
screen as potentially risk significant due to external initiating events
This performance deficiency has a Cross-Cutting aspect in the area of PI&R, CAP,
because Susquehanna did not identify that a setpoint error in operating procedures for
safety-related systems was a CAQ, commensurate with its safety significance. P.1(a)
Enforcement: 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, states, in part,
that conditions adverse to quality shall be promptly identified and corrected. Contrary to
the above, from 1999, when the pressure switches were replaced and the setpoint was
changed, until 2006, when AR 751412751412was written, Susquehanna had failed to identify
that the setpoint was wrong for the low pressure injection permissive interlock in the
operating procedures for RHR and CS. Subsequently, on February 11, 2006, when
Susquehanna personnel initiated and approved AR 751412751412 they failed to identify that
the stated deficiency was a CAQ, which resulted in untimely corrective actions.
Susquehanna considered this to be a procedure change and not a CAQ, and classified
the AR as a CPG versus a CR. As such, the procedures were not changed until July 16,
Enclosure
17
2007, 17 months after the condition was identified and eight years after the setpoint was
changed in the plant. Because this finding is of very low safety significance (Green), and
was entered into the Susquehanna CAP (AR/CR 956917) this violation is being treated
as a Non-Cited Violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement
Policy.
(NCV 05000387/2008006-04; 05000388/2008006-04 - Failure to Identify and Correct
a Setpoint Error in the RHR and CS Operating Procedures)
b. Assessment of the Use of Operating Experience
1. Inspection Scope
The team reviewed a sample of operating experience (OE) issues for applicability to
Susquehanna, and for the associated actions. The documents were reviewed to ensure
that underlying problems associated with the issues were appropriately considered for
resolution. The team also reviewed how Susquehanna considered OE for applicability in
causal evaluations.
Prior to the start of the inspection, the inspectors noted a potential negative trend in the
number of issues associated with reactivity management. In accordance with the
Inspection Procedure, the inspectors increased the scope of the review to determine if
there was an adverse trend in the area of reactivity management over the past five
years. The inspectors reviewed select ARs and CRs associated with the control rod
drive system, control rod problems, human performance issues, and the spent fuel pool;
the inspectors review included how Susquehanna had incorporated applicable OE for
these specific systems and human performance issues into the CAP. The inspectors
interviewed selected licensee staff.
2. Assessment
In general, OE was effectively used at the station. The inspectors noted that OE was
reviewed during the causal evaluation process and incorporated, as appropriate, into the
development of the associated corrective actions. The inspectors noted that OE was
frequently used in work packages and pre-job briefs. The team did not identify any
significant deficiencies within the sample reviewed. The team did not identify a negative
trend nor any significant problems with the control of activities associated with reactivity
management.
3. Findings
No findings of significance were identified in the area of operating experience.
c. Assessment of Self-Assessments and Audits
1. Inspection Scope
Enclosure
18
The team reviewed a sample of departmental self-assessments, CAP trend reports, and
Quality Assurance (QA) audits, including QAs most recent audit of the CAP. The team
also reviewed the latest internal assessment of the safety culture at Susquehanna,
conducted in October 2006. The reviews were performed to determine if problems
identified through these evaluations were entered into the CAP system, and whether the
corrective actions were properly completed to resolve the deficiencies. The
effectiveness of the audits and self-assessments was evaluated by comparing audit and
self-assessment results against self-revealing and NRC-identified findings, and
observations during the inspection.
2. Assessment
The team considered the quality of the audits and self-assessments to be thorough and
critical. ARs were initiated for issues identified by QA and the self-assessments. The
Susquehanna 2006 Comprehensive Cultural Assessment Report consisted of a safety
culture survey and interviews. The cultural assessment report identified some
weaknesses at the station, which were entered into the CAP. The team did not identify
any results that were inconsistent with Susquehannas conclusions.
3. Findings
No findings of significance were identified in the area of audits and self-assessments.
d. Assessment of Safety Conscious Work Environment
1. Inspection Scope
To evaluate the safety conscious work environment (SCWE) at Susquehanna, during
interviews and discussions with station personnel, the team assessed the workers
willingness to enter issues into the CAP and to raise safety issues to their management
and/or to the NRC. The inspectors also interviewed the Employee Concerns Program
(ECP) representative to determine if employees were aware of the program and had
used it to raise concerns. The team reviewed a sample of the ECP files to ensure that
issues were entered into the corrective action program, as appropriate.
2. Assessment
Based on interviews, observations of plant activities, and reviews of the ARs and ECP,
the inspectors determined that the site personnel were willing to raise safety issues and
document them in ARs. Individuals actively utilized the AR system, as evidenced by the
number and significance of issues entered into the program. The inspectors noted that
ARs were written by a variety of personnel, from workers to managers. ECP evaluations
were thorough and appropriate actions were taken to address issues.
3. Findings
No findings of significance were identified related to the SCWE at Susquehanna.
Enclosure
19
4OA6 Meetings, Including Exit:
On February 1, 2008, the team presented the inspection results to Mr. B. McKinney,
Senior Vice President, and to other members of the Susquehanna staff, who
acknowledged the findings. The team confirmed that no proprietary information
reviewed during the inspection was retained.
ATTACHMENT: Supplemental Information
In addition to the documentation that the team reviewed (listed in the Attachment),
copies of information requests given to the licensee are in ADAMS, under accession
number ML080430585.
Enclosure
A-1
ATTACHMENT - SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel:
M. Adelizzi, Risk Engineer
N. DAngelo, Manager, Station Engineering
C. Gannon, Vice President, Nuclear Operations
T. Gorman, Project Manager, Design Engineering
R. Hoffman, Manager, Nuclear Fuels & Analysis
B. McKinney, Chief Nuclear Officer
I. Missien, Project Manager, System Engineering
B. ORourke, Senior Engineer, Nuclear Regulatory Affairs
R. Pagodin, General Manager, Nuclear Engineering
R. Paley, General Manager, Plant Support
A. Price, Supervisor, Corrective Action & Assessment
M. Rochester, Employee Concerns Representative
G. Ruppert, Manager, Maintenance
R. Schechterly, Operating Experience Coordinator
R. Sgarro, Manager, Nuclear Regulatory Affairs
M. Sleigh, Security Manager
B. Stitt, Operations Training
T. Tonkinson, Supervisor, Maintenance Support
D. Weller, Maintenance Foreman
L. West, Supervisor, Central Procedure Group
Nuclear Regulatory Commission:
M. Gray, Branch Chief, Technical Support & Assessment
F. Jaxheimer, Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed:
05000387/2008006-01 NCV Failure to Adequately Evaluate a Deviation from BWROG EPG/SAG
05000388/2008006-01 Resulted in an Inadequate EOP (Section 4OA2.a.3 (a))05000387/2008006-02 NCV Failure to Identify and Correct Inconsistencies in the Licensing Basis05000388/2008006-02 and the EOPs (Section 4OA2.a.3 (b))05000387/2008006-03 NCV Failure to Accurately Model the Simulator for RPV Water Level
05000388/2008006-03 Instrumentation (Section 4OA2.a.3 (c))05000387/2008006-04 NCV Failure to Identify and Correct a Setpoint Error in the RHR and CS
05000388/2008006-04 Operating Procedures (Section 4OA2.a.3 (d))
Attachment
A-2
LIST OF DOCUMENTS REVIEWED
Procedures:
BWROG EGP/SAG and Appendix B Bases, Revision 2
Design Considerations Applicability Sheet Number 42, Emergency Plan, Revision 1
EO-000-102, RPV Control, Revision 2
EO-000-114-1, RPV Flooding, Revision 5
EO-100-103-1, Primary Containment Control, Revision 9
EP-AD-014, Surveillance Testing of Emergency Communications Equipment, Revision 10
EP-AD-015, Review, Revision, and distribution of the SSES Emergency Plan, Revision 11
ME-0RF-161, Control of Fuel Pool Cleanout Activities, Revision 5
ME-0RF-163, Fuel Pool Cleanout - Energy Solutions - Dose Rate Profiling of Irradiated
Hardware and Liners, Revision 4
MFP-QA-1220, Engineering Change Process Handbook, Revision 2
MI-VL-009, Operation of Leak Rate Monitors, Surge Tank Assemblies and 1035 psig Test
Pumps, Revision 3
MT-AD-504, Scaffold Erection, Review and Inspection, Revisions 9 & 10
MT-GM-018, Freeze Sealing of Piping, Revision 15
MT-GM-050, Limitorque Type SMB-000 through SMB-4 Operator Maintenance, Revision 12
NASP-QA-202, Independent Technical Review Program, Revision 2
NASP-QA-401, Internal Audits, Revision 9
NASP-QA-700, Performance Assessment Process, Revision 0
NDAP-00-0109, Employee Concerns Program, Revision 10
NDAP-00-0708, Corrective Action Review Board, Revision 4
NDAP-00-0710, Station Trending Program, Revision 1
NDAP-00-0745, Self-Assessment, Benchmarking and Performance Indicators, Revision 7
NDAP-00-0751, Significant Operating Experience Report (SOER) Review Program, Revision 3
NDAP-00-0752, Cause Analysis, Revisions 3 and 4
NDAP-00-0753, Common Issue Analysis, Revision 0
NDAP-00-0778, Performance Improvement Program, Revision 2
NDAP-QA-0103, Audit Program, Revision 9
NDAP-QA-0330, PSTG and Emergency Procedures, Revision 8
NDAP-QA-0330, Symptom Oriented EOP and EP-DS Program and Writers Guide, Revision 3
NDAP-QA-0412, Leakage Rate Test Program, Revision 10
NDAP-QA-0702, Action Request and Condition Report Process, Revision 20
NDAP-QA-0703, Operability Assessments and Requests for Enforcement Discretion,
Revision 12
NDAP-QA-0720, Station Report Matrix and Reportability Evaluation Guidance, Revision 13
NDAP-QA-0725, Operating Experience Review Program, Revision 11
NDAP-QA-0726, 10CFR50.59 and 10CFR72.48 Implementation, Revision 10
NDAP-QA-1220, Engineering Change Process, Revision 2
NTP-QA-53.1, Susquehanna Fire Brigade Training Program, Revision 15
ODCM-QA-001, ODCM Introduction, Revision 3
ODCM-QA-002, ODCM Review and Revision Control, Revision 4
ODCM-QA-003, Effluent Monitor Setpoints, Revision 3
ODCM-QA-004, Airborne Effluent Dose Calculations, Revision 4
ODCM-QA-005, Waterborne Effluent Dose Calculation, Revision 3
Attachment
A-3
ODCM-QA-006, Total Dose Calculation, Revision 2
ODCM-QA-007, Radioactive Waste Treatment Systems, Revision 2
ODCM-QA-008, Radiological Environmental Monitoring Program, Revision 11
ODCM-QA-009, Dose Assessment Policy Statements, Revision 2
ON-145-004, RPV Water Level Anomaly, Revision 13
OP-024-001, Diesel Generators, Revision 49
OP-024-004, Transfer and Test Mode Operations of Diesel Generator E, Revision 26
OP-149-001, RHR System, Revisions 31 and 32
OP-151-001, Core Spray System, Revisions 27 & 28
SE-124-007, Unit 1 Division 1 Diesel Generator LOCA LOOP Test, Revision 15
SE-259-044, LLRT of RHR Containment Spray Penetration Number X-39A, Revision 11
SOP-054-B03, Quarterly ESW Flow Verification Loop B, Revision 7
SSES-EPG, SSES Plant Specific Technical Guideline, Revision 9
Audits:
666178, Corrective Action, November 2006 - February 2007
667966, QA Internal Audit Report, Fuel Management, Revision 0
691277, QA Internal Audit Report Access Authorization and Fitness for Duty, Revision 0
706249, Operations Training and Qualification Programs, May - June 2007
718607, QA Internal Audit Report, Engineering, Revision 0
744333, Operations, November - December 2007
792034, QA Internal Audit Report, Security, Revision 0
NEIP Audit of Susquehanna Quality Assurance, June 2006
Self-Assessments:
2006 Comprehensive Cultural Assessment, September - October 2006
CA&A Functional Unit Excellence Plan, 1st, 2nd, and 3rd Quarters 2007
CAA-06-01, Site Wide Self-Assessment, December 2006
CAA-06-05, Self-Assessment Program Performance, February 2006
CAA-06-08, Decrease in CR Generation Identified by Trend Report, November 2006
Focused Self Assessment, MOV Program Self-Assessment, October 2007
Maintenance Implementing Procedures Adequacy for Qualified, Inexperienced Employees,
June 2007
Multi-Utility Joint Audit Program Initiative, March - April 2007
NTG Focused Self-Assessment of Operator Training Programs, June 2007
OPS-06-02, Determine the Status of Operator Fundamentals, February 2006
OPS-06-03, Operations Focused Se-f Assessment, July 2006
Pre-PI&R Focused Self-Assessment, September 2007
QA Organization Effectiveness Self-Assessment, October 2006
QA-06-01, Operations QA Audit Preparation Gap Analysis for QC, May - July 2006
SEC-06-01, Analyses of SSES Security Procedures and Physical Security Plan, Revision 0
Attachment
A-4
Action Requests (* denotes an AR/CR generated as a result of this inspection):
478369 724467 741707 759209 779830 810391 843985 873741 896685 941677
524893 724717 741908 759216 780144 810513 845441 873919 897250 941810
542157 726672 741943 759827 780155 811239 849935 874227 898909 947160
545804 728295 742191 760281 780778 811429 851918 875597 899429 954950*
549328 728936 742318 760526 780992 811996 853358 875976 900301 954970*
554362 730852 742342 760526 781644 812948 854681 876021 900720 954972*
554598 730944 742427 762497 782321 813844 855266 876427 901262 954975*
555140 730947 742676 763050 782344 815268 855268 877419 903439 954990*
555263 737236 742966 763128 783655 816097 856997 877727 904689 955072*
555562 738555 743043 763397 784730 816710 858269 877743 908163 955073*
557348 738575 744975 764145 784882 817720 858578 878165 911601 955111*
565795 738634 744979 764738 784890 818082 859082 878326 912213 955130*
575128 738653 745221 764953 785561 818154 859440 879080 912476 955150*
578943 738907 745248 765421 785791 820344 859794 879847 915167 955151*
584400 738999 745462 767566 786149 820380 859839 880331 915620 955761*
591033 739262 745773 767567 786224 820989 860299 880573 916453 955780*
594366 739371 746658 768301 786564 820995 860551 880702 916463 956339*
594887 739371 747077 768502 786735 821006 861162 880806 916873 956344*
595165 739386 747438 768821 786768 821064 861366 881210 917196 956431*
604009 739419 749294 768920 787850 822996 861415 881219 918392 956696*
604296 739579 749341 769304 788616 823908 862474 881225 918549 956914*
610978 739625 749832 769867 788621 824522 864090 881236 919470 956917*
615707 739713 750140 769870 788879 824895 865286 882318 927046 957319*
623914 739737 750232 770453 789971 825107 865423 883987 928515 957484*
623949 740043 751212 771319 791115 825750 865804 886209 929461 957637*
635924 740073 751412 771876 791329 826452 865924 887048 930075 958769*
647827 740303 751433 771961 792158 826870 866930 887067 930571 959670*
655735 740477 751444 773046 793381 827023 867534 888310 931113 961655
666405 740538 752341 773409 794995 827966 867747 889683 932590 962390
668871 740658 752347 774453 795583 828626 867881 889966 936060 962881*
669732 740668 752582 774475 796640 828744 868251 891288 936250 963061*
677145 740723 753392 774509 797517 829065 868259 891733 936370 963065*
687080 740802 753664 774549 799890 829502 868828 891795 936631 963698*
688300 740804 753869 775285 802254 835002 868874 892142 937123 963861*
691108 740825 753990 775718 802539 837153 869819 892152 938054 964512*
693936 740946 755360 776112 802563 837180 869824 892528 938698 964514*
699781 740948 756094 776171 802572 839753 870968 893090 938722 964836*
723483 740955 756415 776769 802697 841169 871013 893157 939516 965167*
723976 740988 756804 776918 805698 841885 872039 893290 939780
724102 741041 757530 777335 806710 842663 872056 895147 941290
724165 741321 757979 777723 809503 842920 873026 896455 941401
724374 741457 758337 778124 809702 843144 873683 896505 941626
Attachment
A-5
Maintenance Work Requests (SPWO):
099065 099364 766396 766413 767284 768234 862569
099115 448229 766401 766416 767490 768618 862578
099120 473889 766406 766496 767506 818282 866262
099259 570758 766411 767283 767532 862503 866284
Non-Cited Violations and Findings Reviewed:
NCV 2005005-01, Inadequate FME Exclusion Procedural Instructions Associated with EDG
Work
FIN 2005009-01, Fire Brigade Drill Program Not Consistent with Regulatory Guidance and
Industry Standards
NCV 2006002-01, Equipment Hatch Plugs are Not Watertight as Indicated in FSAR
FIN 2006002-02, Incomplete Corrective Actions Contribute to CRD Flow Control Failure
NCV 2006003-01, Inadequate Procedures Resulted in Motor Operated Valve Failures
NCV 2006003-02, Failure to Identify Material Degradation which Resulted in the Failure of the
C ESW Pump Breaker
NCV 2006003-03, Inadequate Procedure Results in Elevated Reactor Coolant System Leakage
NCV 2006003-04, Inadequate Design Review of PRDNMS Modification Resulted in a Reactor
NCV 2006003-05, Ineffective Corrective Actions to Assure Training and Qualification of Workers
as Required by 10CFR50, Appendix B, Criterion XVI
NCV 2006004-01, Inadequate Risk Assessment
NCV 2006005-01, Inadequate Work Instructions for the Disassembly and Inspection of Check
Valves
NCV 2006005-02, Inadequate Evaluation of EPA Breaker Failures
NCV 2006006-01, Failure to Identify Scaffolding that Affected the Safety-Related RHR
Discharge Pressure Instrument Tubing Input to ADS
NCV 2006009-01, Safeguards Information
Licensee Identified NCV 2007002, U2 Div II Core Spray Pump Room (a High Radiation Area)
Was Not Posted and Was Open
Licensee Identified NCV 2007002, U1 HPCI Failed a Surveillance Due to the Failure to Perform
Preventive Maintenance
NCV 2007003-01, Failure to Take Timely Corrective Actions for an E EDG Jacket Water Leak
FIN 2007003-02, Failure to Maintain Occupational Radiation Exposure ALARA during Reactor
Water Cleanup Pipe Replacement Activities
FIN 2007003-03, Failure to Maintain Occupational Radiation Exposure ALARA during Outage
ISI of Reactor Pressure Vessel
NCV 2007003-04, Violation of 10CFR71.5 for Inadequately Secured Transport of Condensate
Pump Motors
NCV 2007003-05, Violation of 10CFR71.5 for Inadequately Accounting for Activity in a
Shipment of Irradiated Fuel Channels
Licensee Identified NCV 2007003, U2 Reactor Building HRA Postings and Boundary Moved
without Permission of RP
NCV 2007007-01, Inoperable ESSW Pump-House Ventilation Lineup
NCV 2007007-02, Failure to Use E EDG Procedure
Attachment
A-6
Miscellaneous:
5059-01-2356, 50.59 Screen of Specification C-1056, Long Term Scaffolding, Revision 4
CP067, Corrective Action Program - Evaluation & Resolution, Revision 8
(Lesson Plan & Student Material)
CP068, Managing the Corrective Action Process, Revision 2 (Lesson Plan & Student Material)
Daily CR Screening Team Package
Design Verification Checklist for SCN 6 for Specification C-1056, dated April 27, 2001
EC-059-1024, Design Requirements for and Evaluation of Potential Secondary Containment
Bypass Leakage Pathways, Revision 4
EC-RADN-1029, SSES Design Basis LOCA Dose Consequence Evaluation for Containment
Bypass Leakage Including the Effects of Suppression Pool Scrubbing, Revision 1
EC-SIMU-1001, Evaluation of Simulator Level Instrument Response to Large LOCA, dated
May 4, 1994
Engineering Specification C-1056, Erection of Scaffolding in Safety-Related Areas, Revision 4
EWR #MIS-85-0460, Design Inputs and Considerations Checklist for Specification C-1056,
Revision 2
Hot Box Item 08-01, Reactor Water Instrumentation Response during DBA LOCA, dated
January 31, 2008
IEEE Standard 497-2002, IEEE Standard Criteria for Accident Monitoring Instrumentation, dated
September 30, 2002
Long Term Scaffold Log, dated January 16, 2008
No Degraded Condition Response to OFR 963310, dated January 30, 2008
NRC Information Notice 2007-29, Temporary Scaffolding Affects Operability of Safety-Related
Equipment, dated September 17, 2007
NRC Inspection Procedure 42001, Emergency Operating Procedures, dated June 28, 1991
NRC Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to
Assess Plant and Environs Conditions During and Following an Accident, Revision 2
NRC Regulatory Issue Summary 2005-20, Information to Licensees Regarding Two NRC
Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and
on Operability
NRC Regulatory Issue Summary 2007-21, Adherence to Licensed Power Limits, dated
August 23, 2007
NEDE-24801, Review of BWR Reactor Vessel Water Level Measurement, April 1980
NEDO-24708A, Additional Information Required for NRC Staff Generic Report on Boiling Water
Reactors, Revision 1
Operational Policy Statement (OPS) - 5, Deficiency Control System, Revision 13
Operations Monthly Performance Indicators, December 2007
Operations Quality Assurance Manual, dated December 13, 2007
OPEX Daily Report, January 29, 2008
Plant Modification Package - DCP/ECO #97-9075, Unit 1 Core Spray/RHR/LPCI Pressure
Switch Replacement, Revision 1
PL-NF-02-07, Channel Management Action Plan, Revision 28
Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation, Revision 4
Specification Change Notice #6 for C-1056, Revision 3
Temporary Scaffold Log, dated January 15, 2008
Unit 1 & 2, Control Rod Drive Hydraulics System Health Report, May - August 2007
Unit 1, RHR Residual Heat Removal System Health Report, September - December 2007
Attachment
A-7
LIST OF ACRONYMS
ACE Apparent Cause Evaluation
AR Action Request
BWROG Boiling Water Reactor Owners Group
CAP Corrective Action Program
CAQ Condition Adverse to Quality
CARB Corrective Action Review Board
CFR Code of Federal Regulations
CPG Central Procedure Group
CR Condition Report
DBA Design Basis Accident
DCP Design Change Package
ECCS Emergency Core Cooling System
ECP Employee Concerns Program
EOP Emergency Operating Procedures
EPG/SAG Emergency Procedure Guidelines / Severe Accident Guidelines
EPU Extended Power Uprate
FSAR Final Safety Analysis Report
IMC NRC Inspection Manual Chapter
LOCA Loss of Coolant Accident
NCV Non-Cited Violation
NRC Nuclear Regulatory Commission
OE Operating Experience
PAM Post-Accident Monitoring
PI&R Problem Identification and Resolution
psig pounds per square inch
PSTG Plant Specific Technical Guidelines
QA Quality Assurance
RCA Root Cause Analysis
ROP Reactor Oversight Program
SCWE Safety Conscious Work Environment
SDP Significance Determination Process
TS Technical Specifications
Attachment