ML080770308

IR 05000387-08-006, 05000388-08-006, on 01/14/2008 - 02/01/2008, Susquehanna Steam Electric Station; Biennial Baseline Inspection of He Identification and Resolution of Problems; Corrective Action Program, Simulator Fidelity, and Procedure

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

PPL Susquehanna, LLC

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

License Nos. NPF-14; NPF-22

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

License No: NPF-14, NPF-22

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.

P.1(a)

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

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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

Enforcement Policy.

(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

NRC Enforcement Policy.

(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

Scram

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

CS Core Spray

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

RHR Residual Heat Removal

ROP Reactor Oversight Program

RPV Reactor Pressure Vessel

SCWE Safety Conscious Work Environment

SDP Significance Determination Process

TS Technical Specifications

Attachment