ML050310062

IR 05000387-04-005, 05000388-04-005; 10/01/2004 - 12/31/2004; Susquehanna Steam Electric Station, Units 1 and 2; Equipment Alignments, Operability Evaluations, Access Control to Radiologically Significant Areas, and Radioactive Material Pro

ML050310062
Person / Time
Site:	Susquehanna
Issue date:	01/28/2005
From:	Shanbaky M Division Reactor Projects I
To:	Shriver B PPL Generation
References
IR-04-005
Download: ML050310062 (40)

See also: IR 05000388/2004005

Text

January 28, 2005

Mr. Bryce L. Shriver

President, PPL Generation, LLC and

Chief Nuclear Officer

PPL Generation, LLC

2 North Ninth Street

Allentown, PA 18101

SUBJECT: SUSQUEHANNA STEAM ELECTRIC STATION - NRC INTEGRATED

INSPECTION REPORT 05000387/2004005 AND 05000388/2004005

Dear Mr. Shriver:

On December 31, 2004, the US Nuclear Regulatory Commission (NRC) completed an

inspection at your Susquehanna Steam Electric Station Units 1 and 2. The enclosed integrated

inspection report and Notice of Violation presents the results of that inspection, which was

discussed with Mr. R. Saccone, Vice President - Nuclear Operations and other members of

your staff on January 13, 2005.

This inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

Based on the results of this inspection, the NRC has determined that a Severity Level IV

violation of NRC requirements occurred. The violation was evaluated in accordance with the

"General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement

Policy), NUREG-1600. The current Enforcement Policy is included on the NRCs Web site at

www.nrc.gov; select What We Do, Enforcement, then Enforcement Policy. The violation is

cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding it are

described in detail in the subject inspection report. The violation is being cited in the Notice

because PPL did not restore compliance within a reasonable time by performing a 10 CFR 50.59 evaluation or controlling the Unit 1 railroad bay as part of secondary containment during

subsequent receipt of equipment. Thus, the violation does not qualify for issuance of an NCV

under Section VI the NRC Enforcement Policy.

You are required to respond to this letter and should follow the instructions specified in the

enclosed Notice when preparing your response. The NRC will use your response, in part, to

determine whether further enforcement action is necessary to ensure compliance with

regulatory requirements.

This report also documents three findings of very low safety significance (Green). All three of

the findings were determined to involve violations of NRC requirements. However, because of

the very low safety significance and because the issues were entered into your corrective action

program, the NRC is treating these findings as non-cited violations (NCVs), consistent with

Mr. Bryce L. Shriver 2

Section VI.A of the NRC Enforcement Policy. Additionally, one licensee-identified violation,

which was determined to be of very low safety significance, is listed in this report. If you contest

the NCVs 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 Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional

Administrator Region I; the Director, Office of Enforcement, United States Nuclear Regulatory

Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the

Susquehanna Steam Electric Station.

In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publically Available Records (PARS) component of the NRCs 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).

If you have any questions please contact me at 610-337-5209.

Sincerely,

/RA/

Mohamed Shanbaky, Chief

Projects Branch 4

Division of Reactor Projects

Docket Nos. 50-387; 50-388

License Nos. NPF-14, NPF-22

Enclosures:

1. Notice of Violation

2. Inspection Report 05000387/2004005 and 05000388/2004005

w/Attachment: Supplemental Information

cc w/encls:

J. H. Miller, Executive Vice-President and COO - PPL Services

B. T. McKinney, Vice President - Nuclear Site Operations

R. A. Saccone, Vice President - Nuclear Operations for PPL Susquehanna LLC

A. J. Wrape, III, General Manager- Performance Improvement and Oversight

T. L. Harpster, General Manager - Plant Support

K. Roush, Manager - Nuclear Training

G. F. Ruppert, General Manager - Nuclear Engineering

J. M. Helsel, Manager - Nuclear Operations

R. D. Pagodin, Manager - Station Engineering

J. E. Krais, Manager - Nuclear Design Engineering

T. Mueller, Manager - Nuclear Maintenance

R. Paley, Manager - Work Management

V. L. Schuman, Radiation Protection Manager

J. N. Grisewood, Manager - Corrective Action

R. E. Smith, Manager - Nuclear Site Preparedness and Response

D. F. Roth, Manager - Quality Assurance

R. R. Sgarro, Manager - Nuclear Regulatory Affairs

Mr. Bryce L. Shriver 3

M. Sleigh, Manager - Nuclear Security

W. E. Morrissey, Supervisor - Nuclear Regulatory Affairs

M. H. Crowthers, Supervising Engineer

L. A. Ramos, Community Relations Manager, Susquehanna

B. A. Snapp, Esquire, Associate General Counsel, PPL Services Corporation

R. W. Osborne, Allegheny Electric Cooperative, Inc.

Board of Supervisors, Salem Township

J. Johnsrud, National Energy Committee

Supervisor - Document Control Services

D. Allard, Director, Pennsylvania Bureau of Radiation Protection

Commonwealth of Pennsylvania (c/o R. Janati, Chief, Division of Nuclear Safety,

Pennsylvania Bureau of Radiation Protection)

Mr. Bryce L. Shriver 4

Distribution w/encls: (via E-mail)

S. Collins, RA

J. Wiggins, DRA

M. Shanbaky, DRP

A. Blamey, DRP - SRI Susquehanna

F. Jaxheimer, DRP - RI Susquehanna

S. Farrell, DRP - Susquehanna OA

S. Lee, RI OEDO

R. Laufer, NRR

R. Guzman, NRR

R. Clark, PM, NRR (Backup)

Region I Docket Room (with concurrences)

DOCUMENT NAME: E:\Filenet\ML050310062.wpd

SISP Review Complete: ALB (Reviewers Initials)

After declaring this document An Official Agency Record it will/will not be released to the Public.

To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy

OFFICE RI:DRP RI:DRP RI:ORA RI:DRP

NAME Blamey Burritt Holody Shanbaky

DATE 01/28/05 01/28/05 01/28/05 01/28/05

OFFICIAL RECORD COPY

NOTICE OF VIOLATION

PPL Susquehanna, LLC Docket No. : 50-387

Susquehanna Steam Electric Station License No. : NPF-14

During an NRC inspection conducted between October 1 and December 31, 2004, for which an

exit meeting was held on January 13, 2005, a violation of NRC requirements was identified. In

accordance with the "General Statement of Policy and Procedure for NRC Enforcement

Actions," NUREG-1600, the violation is listed below:

Paragraph (c)(1) of 10 CFR 50.59 states, in part, that a licensee may make changes in

the facility and procedures as described in the Final Safety Analysis Report (FSAR) and

conduct tests or experiments not described in the FSAR without obtaining a license

amendment only if the change, test or experiment does not meet any of the criteria in

paragraph (c)(2) of this section.

Paragraph (d)(1) of 10 CFR 50.59 states, in part, that the licensee shall maintain

records of changes to the facility, procedures, conduct of tests and experiments made

pursuant to paragraph (c) of this section. These records must include a written

evaluation which provides the bases for determination that the change does not require

a license amendment pursuant to paragraph (c)(2) of this section.

Contrary to the above, PPL made a change to the facility, ie the method for performing

or controlling a function, different from that described in the FSAR and did not perform

and maintain records of a written evaluation which provided the basis for determination

that the change does not require a license amendment. Specifically, on December 16,

20, 23, 2004, and on January 4, 2005, PPL changed the ventilation of the Unit 1 railroad

bay from an area within the secondary containment, as described in the FSAR, to an

area outside the secondary containment without a written evaluation pursuant to 10 CFR

50.59.

This is a Severity Level IV violation.

Pursuant to the provisions of 10 CFR 2.201, PPL is hereby required to submit a written

statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document

Control Desk, Washington, DC 20555 with a copy to the Regional Administrator, Region I, and

a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30

days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be

clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1)

the reason for the violation, or, if contested, the basis for disputing the violation or severity level,

(2) the corrective steps that have been taken and the results achieved, (3) the corrective steps

that will be taken to avoid further violations, and (4) the date when full compliance will be

achieved. Your response may reference or include previous docketed correspondence, if the

correspondence adequately addresses the required response. If an adequate reply is not

received within the time specified in this Notice, an order or a Demand for Information may be

issued as to why the license should not be modified, suspended, or revoked, or why such other

Enclosure 1

Notice of Violation 2

action as may be proper should not be taken. Where good cause is shown, consideration will

be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC

Public Document Room or from the NRCs document system (ADAMS), accessible from the

NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should

not include any personal privacy, proprietary, or safeguards information so that it can be made

available to the public without redaction. If personal privacy or proprietary information is

necessary to provide an acceptable response, then please provide a bracketed copy of your

response that identifies the information that should be protected and a redacted copy of your

response that deletes such information. If you request withholding of such material, you must

specifically identify the portions of your response that you seek to have withheld and provide in

detail the bases for your claim of withholding (e.g., explain why the disclosure of information will

create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial

information). If safeguards information is necessary to provide an acceptable response, please

provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working

days.

Dated this 28th day of January 2005

Enclosure 1

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket Nos.: 50-387, 50-388

License Nos.: NPF-14, NPF-22

Report No.: 05000387/2004005, 05000388/2004005

Licensee: PPL Susquehanna, LLC

Facility: Susquehanna Steam Electric Station

Location: 769 Salem Boulevard

Berwick, PA 18603

Dates: October 1, 2004 through December 31, 2004

Inspectors: A. Blamey, Senior Resident Inspector

F. Jaxheimer, Resident Inspector

J. Furia, Sr. Health Physicist

D. Silk, Sr. Emergency Preparedness Inspector

J. Lilliendahl, Reactor Engineer

N. McNamara, Emergency Preparedness Inspector

S. Iyer, Reactor Engineer

G. Meyer, Senior Reactor Inspector

Approved by: Mohamed M. Shanbaky, Chief

Projects Branch 4

Division of Reactor Projects

i Enclosure 2

CONTENTS

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

1. REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1R01 Adverse Weather Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1R04 Equipment Alignments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

1R11 Licensed Operator Requalification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

1R13 Maintenance Risk Assessments and Emergent Work Evaluation . . . . . . . . . . . 5

1R14 Personnel Performance During Non-Routine Plant Evolutions . . . . . . . . . . . . . 6

1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

1R16 Operator Work-Around . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

1R19 Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

1R23 Temporary Plant Modification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

1EP4 Emergency Action Level and Emergency Plan Changes . . . . . . . . . . . . . . . . . 11

2. RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 12

2OS2 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

2OS3 Radiation Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

2PS2 Radioactive Materials Processing and Shipping . . . . . . . . . . . . . . . . . . . . . . . 15

4. OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

4OA3 Event Follow-up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4OA4 Cross Cutting Aspects of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4OA7 Licensee-identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

KEY POINT OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

LIST OF BASELINE INSPECTIONS PERFORMED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2

LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2

LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-5

ii Enclosure 2

SUMMARY OF FINDINGS

IR 05000387/2004005, 05000388/2004005; 10/01/2004 - 12/31/2004; Susquehanna Steam

Electric Station, Units 1 and 2; Equipment Alignments, Operability Evaluations, Access Control

to Radiologically Significant Areas, and Radioactive Material Processing and Shipping.

The report covered a 3-month period of inspection by resident inspectors and announced

inspections by a regional senior health physicist, a senior reactor inspector and two reactor

inspectors. One Severity Level IV Violation and three, Green, non-cited violations (NCVs) of

very low safety significance were identified. The significance of most findings are indicated by

their color (Green, White, Yellow, Red) using Manual Chapter 0609 "Significance

Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be

assigned a severity level after NRC management review. The NRCs program for overseeing

the safe operation of commercial nuclear power reactors is described in NUREG-1649,

"Reactor Oversight Process," Revision 3, dated July 2000.

A. NRC Identified Findings

Cornerstone: Barrier Integrity

C Severity Level VI Violation. The inspectors identified a Severity Level IV violation

of 10 CFR 50.59 requirements for the failure to evaluate a change in plant

system configuration that was known to be inconsistent with accident analysis

and the final safety analysis report (FSAR) description. On December 16, 20, 23

2004, and on January 4, 2005, PPL aligned the ventilation of the Unit 1 Reactor

Building railroad bay to be outside of secondary containment which was

inconsistent with the assumptions of a previously analyzed accident described in

FSAR Chapter 15.6.2. PPL did not perform an evaluation in accordance with the

requirements of 10 CFR 50.59 to determine if the change required a license

amendment prior to implementation of this change in plant configuration.

This finding was addressed using traditional enforcement since it potentially

impacts or impedes the regulatory process in that a required 10 CFR 50.59

evaluation was not performed and documented. A SDP Phase-1 screening was

performed and determined that the condition resulting from the violation of

10CFR 50.59 was of very low safety significance because the finding only

represents a degradation of the radiological barrier function provided by

secondary containment and the standby gas treatment system. This is a

Severity Level IV Violation of NRC requirements in accordance with Section VI.A

of the NRC Enforcement Policy (Supplement I - Reactor Operations; Example

D.5). This violation is being cited in a Notice of Violation under Section VI of the

NRC Enforcement Policy since PPL did not restore compliance within a

reasonable time after the violation was identified nor did they enter the violation

into a corrective action program to address recurrence. (Section 1R15)

C Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion III, Design control, because PPL did not have adequate measures

established to control the alignment of the central railroad bay ventilation to the

secondary containment as described in the accident analysis in the FSAR. This

resulted in several reactor recirculation system and residual heat removal system

iii Enclosure 2

Summary of Findings (contd)

instrument lines being outside of secondary containment. Upon discovery PPL

aligned the central railroad bay ventilation to secondary containment.

This finding was greater than minor because it adversely impacted the Barrier

Integrity cornerstone objective to ensure the capability of containment in that

inadequate design control allowed the instrument lines in the central railroad bay

to be outside of secondary containment. Allowing the instrument lines to be

outside of secondary containment resulted in the plant being outside of the

FSAR assumptions and analysis. This finding was considered to have very low

safety significance (Green), using Phase-1 of the significance determination

process. This finding was Green because the finding only represents a

degradation of the radiological barrier function provided by secondary

containment and the standby gas treatment system. (Section 1R04)

Cornerstone: Occupational Radiation Safety

C Green. A self-revealing non-cited violation of 10 CFR20.1501(a)(1) was

identified for not conducting an adequate radiation survey to ensure compliance

with the High Radiation Area (HRA) posting requirements of 10 CFR 20.1902(b)

during the removal of spent fuel module shield walls. PPL posted and shielded

the location and conducted occupational dose assessments for individuals

working in the unposted high radiation area.

This finding is a greater than minor because PPL did not conduct adequate

radiation surveys to ensure proper posting and control of the area. This finding

was evaluated against the criteria in NRC Manual Chapter 609, Appendix C, and

found to be of very low safety significance (Green) because it was not an ALARA

finding, it did not involve an overexposure or substantial potential for an

overexposure, and the ability to assess dose was not compromised.

The cause of this non-cited violation is related to the Human Performance cross-

cutting area because PPL did not complete an adequate survey to identify a high

radiation area. (Section 2OS1)

Cornerstone: Public Radiation Safety

C Green. A self-revealing non-cited violation of 10 CFR 20.2001 was identified.

PPLs transfer of waste resin to Barnwell Low-Level Waste Disposal facility did

not meet Barnwells license requirements as required by 10 CFR 30.41. On

October 25, 2004, Barnwell identified loose spent resin within the annular space

between the waste container and transport cask. PPL suspended resin

shipments until the cause of the October 25, 2004, event was identified and

corrected.

This finding is a greater than minor performance deficiency because PPL failed

to meet a waste disposal facility license requirement. This radioactive material

control transportation finding was evaluated against criteria specified in NRC

Manual Chapter 0609, Appendix D, and determined to be of very low safety

significance (Green) because no radiation limits were exceeded, no package

breach was involved, no certificate of compliance finding was involved, and

iv Enclosure 2

Summary of Findings (contd)

although a low-level burial ground non-conformance was involved, burial ground

access was not denied and no 10 CFR 61.55 waste classification issue was

involved. (Section 2PS2)

B. Licensee Identified Violation

A violation of very low safety significance, which was identified by PPL, has been

reviewed by the inspectors. Corrective actions taken or planned by PPL have been

entered into PPLs corrective action program. This violation and corrective actions are

listed in Section 4OA7 of this report.

v Enclosure 2

Report Details

Summary of Plant Status

Susquehanna Steam Electric Station (SSES) Unit 1 began the inspection period at full power.

On November 6, 2004, reactor power was reduced to 75% power to perform a condensate

pump motor replacement. On November 20, 2004, reactor power was reduced to 17% and the

main generator was taken off line to repair a main generator hydrogen leak. Unit 1 returned to

full power on November 26, 2004, and continued to operate at full power for the remainder of

the inspection period other than for rod sequence exchanges or rod pattern adjustments.

Unit 2 was operating at or near full power at the beginning of the inspection period. On October

29, 2004, reactor power was reduced to 68% for several hours to repair pipe supports on

feedwater heater piping. Reactor power was reduced to 73% on November 29, 2004, due to an

unexpected rapid increase in cooling tower screen debris. Unit 2 continued to operate at full

power for the remainder of the inspection period, other than for rod pattern adjustments and

planned rod sequence exchanges.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection (71111.01- 1 Sample)

a. Inspection Scope

Adverse Weather Readiness. During the week of December 13, 2004, the inspectors

reviewed PPLs preparations for cold weather. This included a review of open work on

heat trace and other freeze protection measures. Plant walkdowns for selected

structures, systems and components were performed to determine the adequacy of

PPLs weather protection activities. The inspectors also reviewed and evaluated plant

conditions related to severe cold weather and reviewed considerations in PPLs

Maintenance Rule station risk assessment. This inspection activity represented

one sample. The following documents were reviewed:

C OP-185-001, Freeze Protection System

C SO-100-006, Shiftly Surveillance Operating Log

C NDAP-00-0024, Winter Operation Preparations

C CR 631468, Condensate Storage Tank Heat Trace Trouble Alarm

C CR 632090, Temperature Damper TD-27326A Fails to Operate

C CR 630656, T-20 Startup Transformer Fans 7 & 9 Frozen in Place

b. Findings

No findings of significance were identified.

Enclosure 2

2

1R04 Equipment Alignments (71111.04Q - 2 Samples, 71111.04S - 2 Samples)

1. Partial System Walkdowns (71111.04Q - 2 Samples)

a. Inspection Scope

The inspectors performed partial system walkdowns to verify system and component

alignment and to note any discrepancies that would impact system operability. The

inspectors verified selected portions of redundant or backup systems or trains were

available while certain system components were out of service. The inspectors

reviewed selected valve positions, electrical power availability, and the general condition

of major system components. This inspection activity represented two samples. The

walkdowns included the following systems:

C Control Structure Ventilation - Emergency Mode Operation. (control room

emergency outside air supply and floor cooling units)

C Unit 1 Reactor Building - Secondary Containment Ventilation Zones.

b. Findings

Introduction: The inspectors identified a Green non-cited violation (NCV) for inadequate

configuration control of secondary containment as required in 10 CFR 50, Appendix B,

Criterion III, Design control. Inadequate configuration control resulted in reactor

recirculation system and residual heat removal system instrument lines, in the central

railroad bay, to be outside of secondary containment.

Description: PPL did not correctly control the central railroad bay ventilation in

accordance with the Final Safety Analysis Report (FSAR) assumptions and analysis.

This area contains residual heat removal (RHR) and reactor recirculation (RR)

instrument lines that are intended to be inside secondary containment as described in

the FSAR. 10 CFR 50, Appendix B, Criterion III, Design control, requires that the

design basis be correctly translated into procedures. Station Procedure OP-134-002,

Reactor Building HVAC Zones 1 and 3, controls the configuration of secondary

containment and section 2.11, Normal Alignment of the Central Railroad Bay, allowed

this area to be maintained outside of secondary containment.

The RHR system instrument lines for FI-15105A, RHR Loop A Flow Indicator, FT-

15105A, RHR Loop A Flow Transmitter, FT-E11-1N013, Reactor Vessel Head Spray

Flow Transmitter, and PSH-E11-1N022A, RHR Loop A Discharge Pressure, are

routed through the central railroad bay. These instrument lines form part of the ASME

pressure boundary and closed system containment boundary for the RHR system and

represent an extension of primary containment. The Final Safety Analysis Report

(FSAR) section 6.2.3.2.3, Secondary Containment Bypass Leakage, states, in part,

that the secondary containment structure completely encloses the primary containment

structure . . . so that leakage can be collected and filtered prior to release to the

environment.

The RR system instrument lines for flow transmitters FT-B31-1N024A, RR Loop A

Flow, and FT-B31-1N024B, RR Loop B Flow, are also in the central railroad bay.

These instrument lines are connected to the reactor recirculation piping and contain

Enclosure 2

3

reactor coolant. The FSAR, Section 15.6.2, Decrease in Reactor Coolant Inventory,

assumed that for an instrument line break all the reactor coolant from the break would

be contained within secondary containment. Failure of these instrument lines, when the

railroad bay ventilation was aligned to be outside secondary containment, would have

resulted in a potential for unfiltered and unmonitored radioactive material release

bypassing the secondary containment.

Analysis: This finding was a performance deficiency because station procedure OP-

134-002, Reactor Building HVAC Zones 1 and 3, did not correctly control the central

railroad bay to maintain the RR and RHR instrument lines inside of secondary

containment as described in the FSAR assumptions and analysis. Traditional

enforcement does not apply because the issue did not have any actual safety

consequences or potential for impacting the NRCs regulatory function and was not the

result of any willful violation of NRC requirements or PPL procedures. This finding was

more than minor because the lack of adequate design control affected the Barrier

Integrity cornerstone objective to ensure the capability of containment and was

associated with the cornerstone attribute of configuration control to preserve the

containment boundary.

This finding was found to have very low safety significance (Green) using the NRC

Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection

Findings for At-Power Situations. This finding was Green because the finding only

represents a degradation of the radiological barrier function provided by secondary

containment and the standby gas treatment system.

Enforcement: 10 CFR 50, Appendix B, Criterion III, Design control, requires, in part

that, that measures shall be established to assure that applicable regulatory

requirements and the design basis (FSAR) for those structures, systems, and

components to which Appendix B applies are correctly translated into specifications,

drawings, procedures, and instructions. Contrary to the above, the design basis for the

Unit 1 Reactor Building railroad bay ventilation was not adequately translated into

procedures. Specifically, procedure OP-134-002, Reactor Building ventilation zones 1

and 3, did not have appropriate controls to ensure that the central railroad bay

ventilation was maintained within secondary containment to ensure that the RHR system

and RR system instrument lines were inside secondary containment as described in the

FSAR. Because this violation is of very low safety significance and PPL entered this

finding into their corrective action program (CR 621353), this violation is being treated

as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement

Policy. (NCV 50-387/04-05-01, Reactor Recirculation and Residual Heat Removal

System Instrument Lines Outside of Secondary Containment)

2. Complete System Walkdowns (71111.04S - 2 Samples)

a. Inspection Scope

The inspectors performed a complete system walkdown on the Unit 1 reactor core

isolation cooling (RCIC) system to verify that the equipment was properly aligned. The

inspectors reviewed system checkoff lists, system operating procedures, system

emergency support procedure, the system piping and instrumentation diagram and the

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FSAR. The inspectors evaluated outstanding maintenance activities and condition

reports associated with the RCIC system to determine if they would adversely affect

system operability. The inspectors also interviewed the system engineer to identify any

outstanding design issues, temporary modifications and operator workarounds affecting

RCIC system operation. The inspectors verified in the control room and in the RCIC

system room that the valves, including locked valves, were correctly positioned and did

not exhibit leakage that would impact the function of the valve. The inspectors also

verified that all the major components were labeled, hangers and supports were

functional and essential support system were operational.

The inspectors conducted a detailed review of the alignment and condition of the Unit 2

125V DC System including the batteries, battery chargers, and the station trailer

mounted diesel generator (Blue Max). The inspectors also verified that the system

design basis was maintained in the present system configuration and the battery room

ventilation was adequate to prevent excessive hydrogen buildup. Corrective actions

were reviewed for previous 125V DC issues. Weekly, quarterly, and biannual

surveillances were reviewed for completeness and conformance to FSAR and Technical

Specification requirements. These inspection activities represented two samples. The

documents included in the reviews are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - 12 Samples)

a. Inspection Scope

The inspectors reviewed PPL's fire protection program to determine the required fire

protection design features, fire area boundaries, and combustible loading requirements

for selected areas. The inspectors walked down those areas to assess PPLs control of

transient combustible material and ignition sources, fire detection and suppression

capabilities, fire barriers, and any related compensatory measures to assess PPL's fire

protection program in those areas. The inspectors reviewed the respective pre-fire

action plan procedures for the inspected areas. This inspection activity represented

twelve samples. The inspected areas included:

C Unit 1 lower switchgear room, procedure FP-113-222

C Unit 1 core spray pump rooms 645', fire zones 1-1A, 1-1B

C Unit 1 high pressure coolant injection pump room 645', fire zone 1-1C

C Unit 1 upper cable spreading room, procedure FP-013-163

C Unit 1 reactor building 749' and motor generator set, fire zone 1-SA-S

C Unit 2 main turbine lube oil reservoir, procedure FP-213-283

C Unit 2 residual heat removal pump rooms 645', fire zones 2-1E, 2-1F

C Unit 2 reactor building 670', fire zones 2-2A, 2-2B

C Unit 2 upper cable spreading room, procedure FP-013-162

C Unit 2 upper relay room, procedure FP-013-161

C Condensate pump rooms, recombiner room, procedure FP-213-270

C E diesel generator building, procedure FP-013-236

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b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11B, 71111.11Q - 1 Sample)

a. Inspection Scope

Routine Licensed Operator Requalification Exam Results (71111.11B)

On December 6, 2004, the inspector conducted an in-office review of PPLs annual

operating test and biannual written exam results for 2004. The inspection assessed

whether pass rates were consistent with the guidance of NRC Manual Chapter 0609,

Appendix I, Operator Requalification Human Performance Significance Determination

Process (SDP). The inspectors verified that:

• Crew failure rate was less than 20%. (Crew failure rate was 5%.)

• Individual failure rate on the dynamic simulator test was less than or equal to

20%. (Individual failure rate was 3%.)

• Individual failure rate on the walk-through test was less than or equal to 20%.

(Individual failure rate was 1.5%.)

• Individual failure rate on the comprehensive biennial written exam was less than

or equal to 20%. (Individual failure rate was 3%.)

• Overall pass rate among individuals for all portions of the exam was greater than

or equal to 75%. (Overall pass rate was 92.7%.)

Simulator Evaluation (71111.11Q - 1 Sample)

On December 14, 2004, the inspectors observed licensed operator performance in the

simulator during operator requalification training. The inspectors compared their

observations to Technical Specifications, emergency plan implementation, and the use

of emergency operating procedures. The inspectors also evaluated PPLs critique of the

operators' performance to identify discrepancies and deficiencies in operator training.

This inspection activity represented one sample. The following training scenario was

observed:

C Licensed Operator Requalification simulator training scenario OP002-05-02-02,

Loss of Instrument Bus / Shutdown Outside Control Room

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13 - 10

Samples)

a. Inspection Scope

The inspectors reviewed the assessment and management of selected maintenance

activities to evaluate the effectiveness of PPL's risk management for planned and

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emergent work. The inspectors compared the risk assessments and risk management

actions to the requirements of 10 CFR 50.65(a)(4) and the recommendations of

NUMARC 93-01 Section 11, "Assessment of Risk Resulting from Performance of

Maintenance Activities." The inspectors evaluated the selected activities to determine

whether risk assessments were performed when required and appropriate risk

management actions were identified.

The inspectors reviewed scheduled and emergent work activities with licensed operators

and work-coordination personnel to verify whether risk management action threshold

levels were correctly identified. In addition, the inspectors compared the assessed risk

configuration to the actual plant conditions and any in-progress evolutions or external

events to evaluate whether the assessment was accurate, complete, and appropriate for

the emergent work activities. The inspectors performed control room and field

walkdowns to verify whether the compensatory measures identified by the risk

assessments were appropriately performed. This inspection activity represented ten

samples. The selected maintenance activities included:

C Unit 1 main generator H2 leakage, November 20 - 24, 2004

C Unit 1 C condensate pump partial discharge readings increased, CR 610556

• Unit 2 stator water coolant heat exchanger system leakage, CR606722

C Unit 2 instrument air valve 225066 replacement, PCWO 359399

C Unit 2 reactor protection system breakers 2-CB-S003B-B & 2-CB-S003B-D

replacement, WO 610916

C Unit 2 B loop core spray out of service / T-20 work, October 21, 2004

C Unit 2 A loop residual heat removal flow oscillations, AR 617546617546 PCWO

617853

C Unit 2 high pressure coolant injection system outage window, PCWO 506345

C A standby gas treatment system fan trip / damper controller replacement, CR

609389

C Wescosville 2S 500 KV circuit breaker overhaul, WR 156955

b. Findings

No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions (71111.14 - 1 Sample)

a. Inspection Scope

Unit 1 Reduction to Seventeen Percent Power to Correct Main Generator Hydrogen

Leak

On November 20, 2004, Unit 1 was reduced to 17% power to correct a main generator

hydrogen leak. The Inspectors assessed personnel performance during the plant power

changes including removal of the generator from service and the return to full reactor

power. Inspectors evaluated operator actions and verified operator response was

appropriate and in accordance with procedures and training. This inspection activity

represented one sample.

Enclosure 2

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b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - 5 Samples)

a. Inspection Scope

The inspectors reviewed operability determinations that were selected based on risk

insights, to assess the adequacy of the evaluations, the use and control of

compensatory measures, and compliance with the Technical Specifications. In addition,

the inspectors reviewed the selected operability determinations to verify whether the

determinations were performed in accordance with NDAP-QA-0703, "Operability

Assessments." The inspectors used the Technical Specifications, Technical

Requirements Manual, FSAR, and associated Design Basis Documents as references

during these reviews. This inspection activity represented five samples. The issues

reviewed included:

C Unit 1 Reactor coolant instrument lines in Unit 1 railroad bay, CR 621353

C Terminations for core spray & residual heat removal pump motors, CR 609668

C GE Part 21 reactor vessel level instrumentation, CR 606222

C C Emergency diesel generator did not increase load, CR 616488, WO 616497

C Testing of control structure envelope unfiltered in-leakage, CR 535347 and EWR

622198, Generic Letter 2003-001

b. Findings

Introduction: The inspectors identified a Severity Level IV violation of 10 CFR 50.59

requirements for not evaluating a change in plant system configuration that was known

to be inconsistent with the FSAR Chapter 15 accident analysis. Specifically, the railroad

bay ventilation was aligned to be outside of secondary containment on December 16,

20, 23, 2004 and on January 4, 2005.

Description: On November 23, 2004, the inspectors identified reactor recirculation

system instrumentation lines, that contain primary coolant, were located in the Unit 1

reactor building central railroad bay. The railroad bay ventilation was aligned as an area

outside of secondary containment. The accident analysis described in the FSAR

assumed that these instrument lines were within secondary containment. As part of

initial response to this non-conforming configuration, PPL re-aligned the railroad bay to

be part of the secondary containment, evaluated the operabilty of the secondary

containment function, and initiated condition report to address the problem. These

actions were consistent with the NRC process for addressing non-conforming conditions

described in Generic Letter 91-18. (details in Section 1R04)

On December 16, 20, 23, 2004, and on January 4, 2005, prior to the final resolution of

the non-conforming condition, PPL used an established procedure to realign the railroad

bay ventilation and place the railroad bay outside of secondary containment. The

ventilation realignment was done to allow opening of the outer door to the railroad bay

to bring new fuel to the refuel floor. The change in plant system configuration that

placed primary coolant instrument lines outside of secondary containment resulted in

Enclosure 2

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plant operation outside of the documented assumptions in the FSAR Chapter 15

accident analysis. The accident analysis assumed, that for a break of primary coolant

instrument lines, the reactor coolant would be contained within the secondary

containment.

PPL had performed an operability evaluation associated with the non-conforming

configuration of primary coolant instrument lines being outside of secondary

containment before realignment of the railroad bay ventilation to be outside of

secondary containment. The inspectors reviewed PPLs operability evaluation, previous

10 CFR 50.59 evaluations, and the Susquehanna Safety Evaluation Report, NUREG 0776, which states in part, that a circumferential rupture of an instrument line which is

connected to the primary coolant system is postulated to occur inside the secondary

containment. The inspectors did not find an adequate operability or 10 CFR 50.59

evaluation that provided the basis for why realignment of the railroad bay ventilation

outside of secondary containment would not increase or create any of the conditions

described in 10 CFR 50.59 (c)(2) i through viii.

On December 16, 2004, the inspectors discussed with PPL, the inspector position that

the proceduralized activity for realigning the railroad bay ventilation outside of secondary

containment is an activity that was inconsistent with the assumptions of the previously

analyzed Chapter 15.6.2 accident and required the performance of a 10 CFR 50.59

analysis. The inspector noted that prior evaluations (mid-1990s) conducted per 10 CFR 50.59 to change ventilation alignment of the railroad bay to outside secondary

containment were not adequate since they did not consider the instrumentation lines

within the railroad bay. PPL maintained that their operability evaluation for the non-

conforming condition provided a sufficient basis to allow the railroad bay to be outside

secondary containment since the dose consequences from an instrument line break

were still bounded by the worst case analyzed accident. The inspectors noted that the

operability evaluation did not document an assessment of items i through viii in 10 CFR 50.59 (c)(2). Further, the inspectors concluded that the evaluation was not sufficient to

establish operability of the secondary containment with the instrument lines outside of

secondary containment since the assumptions of the instrument line break described in

Chapter 15.6.2 were not maintained. For example, the inspectors noted that

Susquehanna Safety Evaluation Report, NUREG 0776, considers a circumferential

rupture of an instrument line which is connected to a reactor coolant system, but instead

PPLs operability determination assumed a pipe crack. PPL did not take action to

restore compliance with 10 CFR 50.59 during the inspection period. PPL continued to

align the railroad bay ventilation outside of secondary containment. On January 15,

2005, PPL restored compliance by controlling and limiting the time that the railroad bay

ventilation was aligned outside of secondary containment consistent with the Technical

Specification (3.6.4.1) requirements for an inoperable secondary containment.

Analysis: This finding was addressed using traditional enforcement since it potentially

impacts or impedes the regulatory process in that a required 10 CFR 50.59 evaluation

was not performed and documented. This is contrary to the regulatory process that

allows licensees to make changes without a license amendment provided that licensees

will comply with 10 CFR 50.59 process. This finding is more than minor because, the

finding is associated with the configuration control attribute of the containment function

and negatively affects the Barrier Integrity cornerstone objective to provide reasonable

assurance that physical design barriers protect the public from radionuclide releases

Enclosure 2

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caused by accidents or events. Although the significance determination process (SDP)

is not designed to assess the significance of violations that potentially impact or impede

the regulatory process, the result of a 10 CFR 50.59 violation can be assessed by SDP.

An SDP Phase 1 screening was performed and determined that the condition resulting

from the violation of 10 CFR 50.59 was of very low safety significance (Green) because

the finding only represents a degradation of the radiological barrier function provided by

secondary containment and the standby gas treatment system.

Enforcement: Paragraph (c)(1) of 10 CFR 50.59 states that a licensee may make

changes in the facility as described in the FSAR and conduct tests or experiments not

described in the FSAR without obtaining a license amendment only if the change, test or

experiment does not meet any of the criteria in paragraph (c)(2) of this section.

Paragraph (d)(1) states that the licensee shall maintain records of changes to the facility

made pursuant to paragraph (c) of this section. These records must include a written

evaluation which provides the bases for determination that the change does not require

a license amendment. Contrary to the above, on December 16, 20, 23, 2004 and

January 4, 2005 the licensee made a change to the facility as described in the FSAR

and without obtaining a license amendment and did not verify that the change does not

meet any of the criteria in paragraph (c)(2). Additionally, the licensee did not maintain a

record of change to the facility including a written evaluation of the bases for

determination that the change does not require a license amendment. Specifically,

while moving new fuel to the refuel floor, PPL did not maintain instrumentation lines

containing reactor coolant inside of secondary containment as evaluated and described

in the FSAR. This change was implemented without an evaluation to determine if it

resulted in a more than minimal increase in the frequency or consequences of the

accident previously evaluated. This is a Severity Level IV Violation of NRC

requirements in accordance with Section VI.A of the NRC Enforcement Policy

(Supplement I - Reactor Operations; Example D.5). This violation is being cited in a

Notice of Violation under Section VI of the NRC Enforcement Policy since PPL did not

restore compliance within a reasonable time after the violation was identified nor did

they enter the violation into a corrective action program to address recurrence. (NOV

05000387/2004005-02, Failure to Complete 10 CFR 50.59 Analysis)

1R16 Operator Work-Around (71111.16 - 2 Samples)

a. Inspection Scope

The inspectors reviewed the D emergency diesel generator motor operated

potentiometer failure to increase load (CR625636) to determine how the affected system

would impact the operators ability to operate the diesel under emergency conditions.

The inspectors also reviewed the aggregate impact of Unit 1 and Unit 2 documented

operator workarounds and challenges, equipment deficiencies, and open operability

evaluations. The inspectors evaluated the cumulative effects of these items on the

ability of operators to respond in a correct and timely manner. The inspectors also

reviewed these deficiencies to determine if there were any items that complicated the

operators ability to implement emergency operating procedures, but were not identified

as operator workarounds. This inspection activity represented one individual sample

and one cumulative effects sample of operator workarounds.

Enclosure 2

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b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing (71111.19 - 8 Samples)

a. Inspection Scope

The inspectors observed portions of post maintenance testing activities in the field to

determine whether the tests were performed in accordance with the approved

procedures. The inspectors assessed the tests adequacy by comparing the test

methodology to the scope of maintenance work performed. In addition, the inspectors

evaluated the test acceptance criteria to verify whether the test demonstrated that the

tested components satisfied the applicable design and licensing bases and the

Technical Specification requirements. The inspectors reviewed the recorded test data

to determine whether the acceptance criteria were satisfied. This inspection activity

represented eight samples. The post maintenance testing activities reviewed included:

C October 1, 2004, C emergency diesel generator start time testing following air

shuttle valve replacement, CR 597661

C SM-258-003, reactor protection system B electrical protection assembly 24

month calibration and functional test after breaker replacement, CR 610916

C October 10, 2004, SE-259-400, residual heat removal / core spray / high

pressure coolant injection / reactor core isolation cooling component post

maintenance closed system test, PCWO 612562

C October 28, 2004, SE-250-002 logic system functional, and SO-250-002,

RCIC flow verification, following RCIC maintenance.

C Valve time testing following motor replacement on HV-251-FO17B

C November 14, 2004, D emergency diesel generator testing following work in

high voltage cabinet

C Standby gas treatment testing following maintenance, SO-070-001 and PCWO

609397

C December 4, 2004, valve dynamic tests, high pressure coolant injection flow

vibration logic system functional, following Unit 2 high pressure coolant injection

system outage window, SO-252-002, SE-252-002

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 8 Samples)

a. Inspection Scope

The inspectors observed portions of selected surveillance test activities in the control

room and in the field and reviewed the test data results. The inspectors compared the

test result to the established acceptance criteria and the applicable Technical

Specification or Technical Requirements Manual operability and surveillance

requirements to evaluate whether the systems were capable of performing their

Enclosure 2

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intended safety functions. This inspection activity represented eight samples. The

observed or reviewed surveillance tests included:

C SO-024-001D, D Emergency Diesel Generator Surveillance Run,

C SO-258-003, Semi-annual Division I Reactor Protection System Electrical

Protection Assembly Functional Test,

C SO-251-805, B Core Spray Comprehensive Flow Verification,

C SO-150-006, Reactor Core Isolation Cooling Comprehensive Flow Verification,

C SO-024-0016, C Emergency Diesel Generator Monthly Operability Test,

C SR-155-004, Control Rod Drive Scram Time Testing & RE-OTP-103, Stroke

Time Testing, on four rippled control rods,

C SO-070-001, Standby Gas Treatment System Monthly Test,

C SE-159-021, Local Leak Rate Test of Main Steam Line Isolation Valve

Penetration X-7A

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modification (71111.23 - 2 Samples)

a. Inspection Scope

The inspectors reviewed temporary plant modifications to determine whether the

temporary changes adversely affected system or support system availability, or

adversely affected a function important to plant safety. The inspectors reviewed the

associated system design bases, including the FSAR, Technical Specifications, and

assessed the adequacy of the safety determination screenings and evaluations. The

inspectors also assessed configuration control of the temporary changes by reviewing

selected drawings and procedures to verify whether appropriate updates had been

made. The inspectors compared the actual installations to the temporary modification

documents to determine whether the implemented changes were consistent with the

approved documents. The inspectors reviewed selected post installation test results to

verify whether the actual impact of the temporary changes had been adequately

demonstrated by the test. This inspection activity represented two samples. The

following temporary modifications and documents were included in the review:

C T mod 584563 Rev 1, Unit 2 turbine trips bypassed

C T mod 623417, Unit 1 main generator hydrogen makeup flow alarm setpoint

b. Findings

No findings of significance were identified.

Enclosure 2

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Cornerstone: Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope (IP 71114.04 - 1 Sample)

A regional in-office review was conducted of licensee-submitted revisions to the

emergency plan, implementing procedures and emergency action levels (EAL) which

were received by the NRC during the period of October - December 2004. A thorough

review was conducted of plan aspects related to the risk significant planning standards

(RSPS), such as classifications, notifications and protective action recommendations. A

cursory review was conducted for non-RSPS portions. These changes were reviewed

against 10 CFR 50.47(b) and the requirements of Appendix E and they are subject to

future inspections to ensure that the combination of these changes continue to meet

NRC regulations. The inspection was conducted in accordance with NRC Inspection

Procedure 71114, Attachment 4, and the applicable requirements in 10 CFR 50.54(q)

were used as reference criteria. This inspection activity represents one sample.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstones: Occupational Radiation Safety and Public Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01 - 9 Samples)

a. Inspection Scope

The inspector reviewed and assessed the adequacy of PPLs internal dose assessment

for any actual internal exposure greater than 50 mrem committed effective dose

equivalent (CEDE). The inspector examined PPLs physical and programmatic controls

for highly activated or contaminated materials (non-fuel) stored within spent fuel and

other storage pools. The inspector also reviewed self-assessments, audits, licensee

event reports, and special reports related to the access control program since the last

inspection. The inspector determined that identified problems were entered into the

corrective action program for resolution. For repetitive deficiencies or significant

individual deficiencies in problem identification and resolution previously identified, the

inspector determined that PPLs self-assessment activities were also identifying and

addressing these deficiencies.

The inspector reviewed PPL documentation packages for all performance indicator (PI)

events occurring since the last inspection.

The inspector selected jobs being performed in radiation areas, airborne radioactivity

areas, or high radiation areas (less than 1 R/hr) for observation. The inspector reviewed

all radiological job requirements and observed job performance with respect to these

requirements. The inspector determined that radiological conditions in the work area

were adequately communicated to workers through briefings and postings. The jobs

Enclosure 2

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reviewed and observed included the removal and replacement of the filter elements in

the 2B condensate filtration system filter.

The inspector discussed with first-line health physics (HP) supervisors the controls in

place for special areas that have the potential to become very high radiation areas

(VHRA) during certain plant operations. The inspector determined that these plant

operations required communication beforehand with the HP group, so as to allow

corresponding timely actions to properly post and control the radiation hazards.

These inspection activities represented nine samples. The documents reviewed are

provided in the Attachment.

In addition the inspector reviewed Licensee Event Reports, Special Reports, audits,

State agency reports, and self-assessments related to the radioactive material and

transportation programs performed since the last inspection to determined that identified

problems were entered into the corrective action program for resolution. The inspector

also reviewed corrective action reports written against the radioactive material and

shipping programs since the previous inspection. The inspector reviewed PPLs

evaluation of the detection of an unposted High Radiation Area during preparation of a

spent fuel storage horizontal module (B-5) on September 16, 2003 (CR 509273).

These reviews were conducted using the requirements contained in 10 CFR 20.

b. Findings

Introduction: A green self-revealing non-cited violation of 10 CFR20.1501(a)(1) was

identified for not conducting an adequate radiation surveys to ensure compliance with

the High Radiation Area posting requirements of 10 CFR 20.1902(b) during the removal

of spent fuel storage module shield walls.

Description: On August 20 and 21, 2003, PPL workers removed the shield walls from

two empty horizontal spent fuel storage modules (HSMs)(B-4, C-4) in preparation for

installing six additional HSMs. Radiation protection personnel performed radiation

surveys to support removal of shielding from the modules due to potential radiation

streaming from previously filled HSMs. The radiation protection personnel briefed

workers on the apparent radiation dose rates during installation and preparation of the

new modules during the period August 21, 2003 - September 16, 2003. During work on

September 16, 2003, on module B-5 one workers integrating alarming dosimeter

alarmed. The worker left the area, informed radiation protection, and an investigation

was initiated. The workers dosimeter alarmed due to the dosimeter exceeding its alarm

set point. Radiation protection personnel conducted detailed radiation surveys to identify

the apparent cause of the alarm and identified, a previously undetected High Radiation

Area that was accessible to personnel. The area exhibited radiation dose rates of 170

mr/hr at 30 centimeters from the wall in the B-5 module. Subsequent PPL review

identified that the High Radiation Area was associated with radiation streaming through

an overhead air vent from an adjacent HSM B-4, where the shielding had been

removed. The High Radiation Area had not been identified after removal of shielding on

August 21, 2003.

PPL suspended work, posted the area, conducted occupational radiation dose

assessments, installed shielding as appropriate, and placed the issue in its corrective

Enclosure 2

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action program. Although the area was accessible, the workers dose alarm was

believed not to be attributable to the undetected High Radiation Area. Notwithstanding,

PPL conducted occupational dose assessments to assess possible additional dose from

the undetected High Radiation Area. PPL identified several individuals who sustained

additional dose but none of the individuals were estimated to receive greater than 100

millirem.

Analysis: This finding is a performance deficiency because PPL did not detect and post

a High Radiation Area, exhibiting accessible radiation dose rates of 170mr/hr at 30

centimeters. The finding is not subject to traditional enforcement in that the finding did

not have any actual safety consequence, did not have the potential for impacting the

NRCs ability to perform its regulatory function, and there were no willful aspects. In

addition, this finding specifically involved the stations basic radiological controls

program.

The finding was greater than minor in that it is associated with the program and process

attribute (exposure control and monitoring) of the Occupational Radiation Safety

Cornerstone and did affect the cornerstone. Specifically, PPLs programs and processes

did not detect an accessible High Radiation Area and ensure appropriate postings and

controls were in-place to preclude workers unknowingly entering and working in the

area. The finding was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix C, and determined to be of very low safety significance (Green), in that

it was not an As Low As Is Reasonable Achievable (ALARA) finding, no overexposure

occurred, there was no substantial potential for an overexposure, and the ability to

assess dose was not compromised. (CR 509273).

The cause of this non-cited violation is related to the Human Performance cross-cutting

area because PPL did not complete an adequate survey to identify a high radiation

area. This resulted in an unposted high radiation area at the HSM B-5.

Enforcement: 10 CFR 20.1501 requires that necessary and reasonable radiological

surveys be conducted to evaluate potential radiological hazards including High Radiation

Areas as required by 10 CFR 20.1902(b). Contrary to this requirement, due to

inadequate radiation surveys, PPL did not detect a High Radiation Area in storage

module B-5 following shield removal in August 2003. This is a violation of 10 CFR

20.1501. Because this finding was of very low safety significance (Green), and PPL

entered this finding into its corrective action program, this violation is being treated as a

Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.

NUREG-1600. (NCV 05000387/2004005-03, Failure to Post Horizontal Spent Fuel

Storage Module B-5 as a High Radiation Area)

2OS2 ALARA Planning and Controls (71121.02 - 2 Samples)

a. Inspection Scope

The inspector reviewed PPLs self-assessments, audits, and special reports related to

the ALARA program since the last inspection. The inspector determined that PPLs

overall audit programs scope and frequency (for all applicable areas under the

Occupational Cornerstone) meet the requirements of 10 CFR 20.1101(c).

Enclosure 2

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The inspector determined that identified problems are entered into the corrective action

program for resolution. The inspector reviewed dose significant post-job (work activity)

reviews and post-outage ALARA report critiques of exposure performance, and

determined that identified problems are properly characterized, prioritized, and resolved

in an expeditious manner. This inspection activity represented two samples. The

documents reviewed are provided in the Attachment.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation (71121.03 - 2 Samples)

a. Inspection Scope

The inspector reviewed PPLs self-assessments, audits, and Licensee Event Reports

and focused on radiological incidents that involved personnel contamination monitor

alarms due to personnel internal exposures. For repetitive deficiencies or significant

individual deficiencies in problem identification and resolution, the inspector determined

that PPLs self-assessment activities are also identifying and addressing these

deficiencies.

The inspector reviewed documents related to PPLs processing of thermoluminescent

dosimeters (TLDs) to measure personnel dose of record. Documents reviewed included

the most recent laboratory testing (Personnel Dosimetry Performance Testing Report

dated January 9, 2004) and laboratory audit (On-Site Assessment 100554-0, February

2003) of PPLs program and facility by the National Voluntary Laboratory Accreditation

Program (NVLAP). This inspection activity represented two samples. The documents

reviewed are provided in the Attachment.

b. Findings

No findings of significance were identified.

2PS2 Radioactive Materials Processing and Shipping (7112202 - 6 Samples)

a. Inspection Scope

The inspector reviewed the solid radioactive waste system description presented in the

FSAR and the recent radiological effluent release report for information on the types and

amounts of radioactive waste disposed, and also reviewed the scope of PPLs audit

program to verify that it met the requirements of 10 CFR 20.1101.

The inspector walked-down and visually inspected the liquid and solid radioactive waste

processing systems to verify that the current system configuration and operation was

consistent with the descriptions provided in the FSAR and the Process Control Program.

The inspector reviewed the status of radioactive waste process equipment that was not

operational or abandoned in place and verified that applicable changes were reviewed

and documented in accordance with 10 CFR 50.59, as appropriate. In addition, the

inspector reviewed current processes for transferring radioactive waste resin and sludge

Enclosure 2

16

discharges into shipping/disposal containers to determine if appropriate waste stream

mixing and/or sampling procedures, and methodology for waste concentration

averaging, provided for representative samples of the waste product for the purposes of

10 CFR 61.55 waste classification.

The inspector reviewed the radiochemical sample analysis results for each of the

stations radioactive waste streams; reviewed the PPLs use of waste scaling factors and

calculations used to account for difficult-to-measure radionuclides; verified that the

program assured compliance with 10 CFR 61.55 and 10 CFR 61.56, as required by

Appendix G of 10 CFR Part 20; and, reviewed the program to ensure that the waste

stream composition data accounted for changing operational parameters and remained

valid between the annual or biennial sample analysis updates.

The inspector observed shipment packaging, surveying, labeling, marking, placarding,

vehicle checks, emergency instructions, disposal manifest, shipping papers provided to

the driver, and PPL verification of shipment readiness; verified that the requirements of

any applicable transport cask Certificate of Compliance had been met; verified that the

receiving licensee was authorized to receive the shipment packages; and, observed

radiation workers during the conduct of radioactive waste processing and radioactive

material shipment preparation activities. The inspector determined that shippers were

knowledgeable of the shipping regulations and that shipping personnel demonstrated

adequate skills to accomplish the package preparation requirements for public transport

with respect to NRC Bulletin 79-19 and 49 CFR Part 172 Subpart H; and verified that

PPLs training program provided training to personnel responsible for the conduct of

radioactive waste processing and radioactive material shipment preparation activities.

The inspector sampled non-excepted package shipment records and reviewed these

records for conformance with applicable NRC and DOT requirements.

b. Findings

Introduction: A green self-revealing non-cited violation of 10 CFR 20.2001 was

identified. PPLs transfer of waste resin to Barnwell Low-Level Waste Disposal facility

did not meet Barnwells license requirements as required by 10 CFR 30.41. On October

25, 2004, Barnwell identified loose spent resin within the annular space between the

waste container and transport cask which is prohibited by Barnwells license (License

No. 097, Condition 61).

Description: On October 25, 2004, personnel from the South Carolina Department of

Health and Environmental Control, conducted an inspection of a shipment of radioactive

waste (04-155) from SSES. Shipment 04-155 was a polyethylene waste container filled

with a mixture of filter sludge and spent bead resin, placed inside an NRC-licensed Type

B shipping packaging (10-142B cask [USA/9208/B]). During off-loading and removal of

the container from the cask at Barnwell, radioactive resin was observed on the bottom of

the shipping cask. The resin was collected, surveyed, and found to exhibit low radiation

levels. PPL was subsequently notified by the Barnwell Low-Level Waste Disposal

Facility that shipment 04-155, shipped from the SSES, had radioactive resin outside the

waste disposal container, in violation of the waste disposal facilitys site operating

license (License No. 097, Condition 61), in that PPL did not package the shipment in a

manner that would prevent the release of radioactive waste into the shipping container.

Enclosure 2

17

The inspectors review identified that following loading of the waste container into the

cask at SSES, a quantity of spent resin was found on the upper surface of the waste

container. PPL vacuumed off this material prior to closing the cask, however, some

material remained in the annular space between the shipping container (cask) and

waste container, unknown to the licensee.

Analysis: This finding is a performance deficiency because PPL did not meet the

disposal license condition which was reasonably within PPLs ability to foresee and

correct, and which should have been prevented. The finding is not subject to traditional

enforcement in that the finding did not have any actual safety consequence, did not

have the potential for impacting the NRCs ability to perform its regulatory function, and

there were no willful aspects.

The finding was greater than minor in that it is associated with the program and process

attribute (radioactive material control/transportation) of the Public Radiation Safety

cornerstone and did affect the cornerstone. Specifically, PPL did not meet the

requirements of Barnwell disposal facilitys operating license to provide for proper

packaging of waste for shipment to prevent release of radioactive waste into the

shipping container. The finding was evaluated against criteria specified in NRC Manual

Chapter 0609, Appendix D, and determined to be of very low safety significance

(Green), because no radiation limits were exceeded, no package breach was involved,

no certificate of compliance finding was involved, and although a low-level burial ground

non-conformance was involved, burial ground access was not denied and no 10 CFR 61.55 waste classification issue was involved. The small quantity of loose resin was

contained within the confines of the shipping cask. PPL suspended resin shipments

when notified and placed the issue in its corrective action program (CR 613944).

Enforcement: 10 CFR 2001 and 10 CFR 30.41 require that the licensee may only

transfer licensed materials to a person authorized to receive such material under terms

of a specific license issued by an Agreement State. Condition 61, of License 097

(Amendment 48) issued for the operation of the Barnwell Waste Management Facility by

the State of South Carolina (an Agreement State), prohibits packaging of shipments in a

manner that would result in release of radioactive waste into the shipping container.

Contrary to this requirement, loose waste resin was found within the annulus space

between the resin container and the shipping container (cask) for SSES shipment No.04-155 on October 25, 2004. This is a violation of 10 CFR 20.2001. Because this

finding was of very low safety significance (Green), and PPL entered this finding into its

corrective action program, this violation is being treated as a Non-Cited Violation (NCV)

consistent with Section VI.A of the NRC Enforcement Policy. NUREG-1600. (NCV 05000387/2004005-04, Failure to correctly Package Waste Resin for Shipment)

Enclosure 2

18

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151 - 16 Samples)

Cornerstone: Reactor Safety

a. Inspection Scope

The inspectors reviewed PPLs performance indicator (PI) data, for the period of

November 2003 through November 2004, to verify whether the PI data was accurate

and complete. The inspectors examined selected samples of PI data, PI data summary

reports, and plant records. The inspectors compared the PI data against the guidance

contained in Nuclear Energy Institute (NEI) 99-02, revision 2, "Regulatory Assessment

Performance Indicator Guideline." The inspectors also observed a chemistry technician

obtain a reactor water sample on December 23, 2004. This inspection activity

represented 14 samples. The following indicators and PPL documents were included in

this review:

Initiating Event Performance Indicators

• Units 1 & 2 Unplanned Scrams per 7000 Critical Hours

• Units 1 & 2 Scrams With Loss of Normal Heat Removal

• Units 1 & 2 Unplanned Power Changes per 7000 Critical Hours

Mitigating Systems Performance Indicators

• Units 1 & 2 Emergency AC Power System Unavailability

• Units 1 & 2 Residual Heat Removal System Unavailability

Barrier Integrity Performance Indicators

• Units 1 & 2 Reactor Coolant System (RCS) dose equivalent iodine specific

activity

• Units 1 & 2 RCS Identified leak rate measured by the drywell leakage calculation

PPL Documents

• Units 1 & 2 Control Room Logs

• NDAP-QA-0737, "Regulatory Performance Assessment"

• Technical Specification 3.4.4, "RCS Operational Leakage"

• SO-100/200-006, "Shiftly Surveillance Operating Log"

• SC-176/276-102, "Reactor Coolant Dose Equivalent Iodine-131"

• Units 1 & 2 Licensee Event Reports

Enclosure 2

19

Cornerstone: Occupational Radiation Exposure

a. Inspection Scope (71151 - 1 Sample)

The inspector reviewed all licensee performance indicators (PIs) for the Occupational

Exposure Cornerstone for follow-up. The inspector reviewed a listing of licensee event

reports for the period January 1, 2004 through November 28, 2004 for issues related to

the occupational radiation safety performance indicator, which measures non-

conformance with high radiation areas greater than 1R/hr and unplanned personnel

exposures greater than 100 mrem total effective dose equivalent (TEDE), 5 rem skin

dose equivalent (SDE), 1.5 rem lens dose equivalent (LDE), or 100 mrem to the unborn

child.

The inspector determined if any of these PI events involved dose rates greater than 25

R/hr at 30 centimeters or greater than 500 R/hr at 1 meter. If so, the inspector

determined what barriers had failed and if there were any barriers left to prevent

personnel access. For unintended exposures greater than 100 mrem TEDE (or greater

than 5 rem SDE or greater than 1.5 rem LDE), the inspector determined if there were

any overexposures or substantial potential for overexposure. This inspection activity

represents one sample.

b. Findings

No significant findings or observations were identified.

Cornerstone: Public Radiation Safety

c. Inspection Scope (71151 - 1 Sample)

The inspector reviewed a listing of licensee event reports for the period January 1, 2004

through November 28, 2004, for issues related to the public radiation safety

performance indicator, which measures radiological effluent release occurrences per

site that exceed 1.5 mrem/qtr whole body or 5 mrem/qtr organ dose for liquid effluents;

or 5 mrads/qtr gamma air dose, 10 mrads/qtr beta air dose; or 7.5 mrems/qtr organ

doses from I-131, I-133, H-3 and particulates for gaseous effluents. This inspection

activity represents one sample.

b. Findings

No significant findings or observations were identified.

4OA2 Identification and Resolution of Problems (71152 - 1 Annual Sample, 1 Semi-Annual

Sample)

a. Inspection Scope

Annual Sample Review - ESW Equipment Replacement/Flow Balance/Modeling Issues

(71152 - 1 Annual Sample)

Enclosure 2

20

Inspectors reviewed the effectiveness of corrective actions associated with the

Emergency Service Water (ESW) system flow balance and the associated emergency

heat sink safety function. This sample included a review of corrective actions

associated with valve seat leakage to reactor building closed cooling water, turbine

building closed cooling water and the alternate train of the E Emergency Diesel

Generator ESW cooling. NCV 2001005-001 identified leakage paths that were not

tested that could impact safety by diverting the cooling water flow from Emergency

Service Water to interfacing systems. Although the testing of these leakage paths was

implemented promptly in 2001 to assure system operability, several of the long-term

actions to restore system health by replacing these and other system boundary valves

were completed by PPL in 2004. Inspectors screened a collection of corrective actions

associated with maintaining the design cooling water flows to ESW cooled components.

Inspectors reviewed the conditions adverse to quality entered into the PPL corrective

action system and those in progress during the year to determine the aggregate impact

on the ability of the ESW system to perform safety functions.

Inspectors reviewed the results of the ESW system flow balance, TP-054-076, as well

as comprehensive pump testing results and compared this performance information to

the flow models used previously to evaluate system operability and system performance

trends. ESW measured flows were compared to FSAR assumptions and values used in

design calculations. Inspectors concentrated review on the corrective actions identified

by engineering or associated with recent field observations of equipment performance or

configuration such as unexpected valve throttle position. Corrective Action reports and

the other technical references reviewed are listed in the Attachment. The inspectors

found that concerns and issues for the ESW system were identified, documented and

properly evaluated through the PPL corrective action program.

Semi-Annual PI&R Trend Review (71152 - 1 Semi-Annual Sample)

The inspectors reviewed 221 action request (AR) items that were categorized as

Management sub type, Chemistry and Effluents, as part of the semi-annual baseline

inspection documented in this report. Fifteen of the ARs were reviewed in detail to verify

whether the full extent of the issues were adequately identified, appropriate evaluations

were performed, and reasonable corrective actions were identified. The inspectors

evaluated the ARs against the requirements of NDAP-QA-0702, "Action Request and

Condition Report Process," and 10 CFR 50, Appendix B. The 15 ARs reviewed in detail

were: 582584, 583122, 583526, 584603, 586479, 585323, 589980, 582686, 586411,

586411, 591296, 595712, 599809, 604772, and 612621.

Routine PI&R Review

The inspectors reviewed selected condition reports (CRs), as part of the routine

baseline inspection documented in this report. The CRs were assessed to verify

whether the full extent of the various issues were adequately identified, appropriate

evaluations were performed, and reasonable corrective actions were identified. The

inspectors evaluated the CRs against the requirements of NDAP-QA-0702, "Action

Request and Condition Report Process," and 10 CFR 50, Appendix B. During this

inspection period, the inspectors performed a screening review of each item that PPL

entered into their corrective action program, to assess whether there were any

Enclosure 2

21

unidentified repetitive equipment failures or human performance issues that might

warrant additional follow-up.

b. Findings and Observations

No findings of significance were identified.

4OA3 Event Follow-up (71153 - 1 Sample)

1. (Closed) LER 05000387/2004-004-00 Radiation Monitors Inoperable During Spent Fuel

Cask Transport - Operation Prohibited by Technical Specification

On August 20, 2004, PPL discovered that the Secondary Containment Zone 3 isolation

relays for both process radiation monitor in the central railroad access bay were

disabled. These trips had been disabled on July 16, 2004, when an Instrument &

Control Technician incorrectly executed steps in procedure IC-079-012, Railroad

Access Shaft Radiation Monitor Alarm / Trip Disabling. On August 2, and August 16,

2004, spent fuel storage casks had been moved in this area. Technical Specification 3.3.6.2, Secondary Containment Isolation Instrument, and 3.3.7.1, Control Room

Emergency Outside Air Supply System, require the railroad access shaft radiation

monitors be operable during movement of irradiated fuel in the railroad access shaft.

Corrective actions included reaffirm work standards with the individuals and a plan to

provide this information to all maintenance personnel. This finding is more than minor

because the radiation monitors would not have functioned automatically in response to a

radiological condition in the railroad access shaft (Zone 3 - spent fuel pool zone). The

finding affects the Barrier Integrity Cornerstone and was considered to have very low

safety significance (Green) using a Phase -1 SDP, because the finding only represented

a degradation of the radiological barrier for the control room and spent fuel pool zone.

The enforcement aspects of the violation are discussed in Section 4OA7. This LER is

closed.

4OA4 Cross Cutting Aspects of Findings

Cross Reference to Human Performance Findings Documented Elsewhere

Section 2OS1 describes an NCV where PPL did not complete an adequate survey to

identify a high radiation area. This resulted in an unposted high radiation area at the

horizontal spent fuel module B-5.

4OA6 Meetings, Including Exit

On January 13, 2005, the resident inspectors presented the inspection results to Mr. R.

Saccone, Vice President - Nuclear Operations, and other members of your staff, who

acknowledged the findings.

4OA7 Licensee-identified Violations

The following violation of very low safety significance (Green) was identified by PPL and

is a violation of NRC requirements which meet the criteria of Section VI of the NRC

Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.

Enclosure 2

22

C Technical Specification 3.3.6.2, Secondary Containment Isolation Instrument,

and 3.3.7.1, Control Room Emergency Outside Air Supply System, require the

railroad access shaft radiation monitors be operable during movement of

irradiated fuel in the railroad access shaft. Contrary to this on August 2, and

August 16, 2004, spent fuel storage casks had been moved in this area. This

was identified in the PPL corrective action program as CR 600250. This finding

is of very low safety significance because it only represented a degradation of

the radiological barrier for the control room and spent fuel pool zone.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure 2

A-1

SUPPLEMENTAL INFORMATION

KEY POINT OF CONTACT

1R04 Equipment Alignment

Kevin Daly - Lead Engineer

John Vandenberg - Backup Engineer

1R04 Equipment Alignment

Paul Capotos

Len Casella

John Rotha

Phil Brady

Eric Miller

1R11 Licensed Operator Requalification

B. Stitts, Susquehanna Training Department

2PS2 Radioactive materials Processing and Shipping

D. Davis, Technical Training Instructor

R. Hock, Radiological Operations Supervisor

J. Meter, Licensing Engineer

M. Micca, Health Physicist - Waste Shipping

V. Schuman, Radiation Protection Manager

V. Zukauskas, Jr., Health Physics Foreman

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

050000387, 388/2004005-02 NOV Failure to Complete 10 CFR 50.59 Analysis

Opened and Closed

05000387/2004005-01 NCV Reactor Recirculation and Residual Heat Removal

System Instrument Lines Outside of Secondary

Containment

05000387/2005/005-03 NCV Failure to Post Horizontal Spent Fuel Storage

Module B-5 as a High Radiation Area

05000387/2004005-04 NCV Failure to Correctly Package Waste Resin for

Shipment

Closed

05000387/2004-004-00 LER Radiation Monitors Inoperable During Spent Fuel

Cask Transport - Operation Prohibited by Technical

Specification

Attachment

A-2

LIST OF BASELINE INSPECTIONS PERFORMED

2PS2 Radioactive materials Processing and Shipping

7112101 Access Control 2OS1

7112202 Radioactive Material Processing and Shipping 2PS2

71151 Performance Indicator Verification 4OA1

LIST OF DOCUMENTS REVIEWED

(Not Referenced in the Report)

Section 1R04: Equipment Alignment

P&ID

Reactor Core Isolation Cooling - PPL drawing no E106254, AE drawing no -149, Rev 46

Reactor Core Isolation Cooling - PPL drawing no E106255, AE drawing no -150, Rev 26

Procedures & Checkoff list

RCIC manual injection with a loss of AC and DC power -ES 150(250)-003

Electrical - CL-150-0011 Rev - 11

Mechanical - CL-150-0012 Rev - 18

Containment - CL-150-0013 Rev 5

Notifications

CR 478799 CR 654600 CR 613953

CR 613952 CR 613776 CR 613573

CR 613555 CR 608809 CR 575709

CR 468503 CR 614504 CR 614407

CR 614319 CR 604479 CR 597589

CR 596983 CR 596900 CR 571749

CR 571046 CR 538717 CR 538717

Action Request and Change Request

CRA 491260 AR 354431354431 AR 616048616048AR 616053 AR 616056616056 AR 616057616057System Health Report

RCIC Unit 1 and Unit 2 dated 08/21/2004

Miscellaneous

UFSAR - 5.4.6 Reactor core isolation cooling

Info Rev 0, 03/28/83 - Reactor core isolation

Documents Calculations

EC-SBOR-0501 SBO Coping Assessment

EC-SBOR-0506, Rev 0, 5/19/94 SBO Required Coping Duration

EC-002-1031, Rev 5, 8/25/04125V DC Load Profiles

Attachment

A-3

EC-002-0505, Rev 13, 11/8/04 Unit 2, D Battery Load Profile Calculation

EC-002-0504, Rev 25, 11/15/04 Unit 2, B Battery Load Profile Calculation

EC-088-0526, Rev 2, 12/29/2000 Battery Room Hydrogen Generation

EC-013-0561, Rev 6, 1/2/01 Appendix R - HVAC Study

Design Basis

DBD001, Rev 4, 9/25/03 Design Basis Document for Class 1E DC Electrical

FSAR Section 8.3.2 DC Power Systems

Procedures/Surveillances

OP-202-001, Rev 13, 8/17/04 125V DC System Operation

EO-200-030, Rev 16, 1/14/04 Unit 2 Response to Station Blackout

SM-202-001, Completed 12/8/04 Weekly Battery Surveillance

SM-202-002, Completed 12/2/04 Quarterly Battery Surveillance

SM-202-D04, Completed 3/21/03 48-Month Modified Performance Test

AR/CRs

550022 Correction to Unit 1, A 125V battery load profile

550397 Review of all battery load profiles

473769 Battery testing documentation

339039 Battery charger voltage not within limits 3 times

221157 Replacement of mixed cells in Unit 2, D 125V battery

Generated as a result of this inspection

625328 Inaccuracy in FSAR section 8.3.2.1.1.5 regarding battery cell classification

627984 TS 3.8.4.7 is not met due to unreasonable 60 month exception note

Section 1EP4: Emergency Action Level (EAL) and Emergency Plan Changes

Susquehanna Emergency Response Plan and Implementing Procedures

Section 2PS2: Radioactive materials Processing and Shipping

Radioactive Material Shipments: 04-146; 04-151;04-154; 04-155;04-156

Quality Assurance Internal Audit Report No. 435295, Solid Radwaste

Self-Assessment HPS-04-02, EPRI Liquid Radwaste Management Assessment

Low Level Waste Characterization Study, October 2003

Radiological Profile Report, Unit 1 Thirteenth Cycle

Procedures: HP-TP-103, Rev 3, Plant Radiation Profile

HP-TP-721, Rev 3, Gamma-to-Alpha Ratio Determinations

NTP-QA-53.3, Rev 3, Hazardous Materials Handling, Packaging, Shipping and

Transportation Training Program

WM-PS-100, Rev 9, Shipment of Radioactive Waste

WM-PS-110, Rev 5, General Shipment of Radioactive Material

WM-PS-210, Rev 7, Packaging and Loading of DAW and Radioactive Material

WM-PS-310, Rev 3, Use of the 10-142B Shipping Cask

Lesson Plans: MST-320, Hazardous Material Shipping and Handling Large Quantities

MST-325, Hazardous Material/Shipping and Handling

MST-336, DOT Security Awareness and Plan

Attachment

A-4

HP-230, Receipt and Shipment of Radioactive Material

HS-053, Hazmat Employee Training for Container Loaders

EF-009, Qualified Loader of Radioactive Material

HP-242, Fundamentals of Radwaste Shipping

HP-246, Radwaste Shipping Technician Orientation

HP-248, Use of Shipping Document Computer Programs

Condition Reports: 621672; 613944; 602411; 597666; 594215; 593074; 600491; 600517;

603630; 610452; 616287

Section 4OA2: Identification and Resolution of Problems

Procedures

OP-054-001, Revision 22, Emergency Service Water System

SO-024-014,

TP-054-076

SO-054-002

AR/CRs

544629, 548869 550087 551225

552695 572573 593354 594262

604482 604960 621817

EWRs and Calculations

EWR # 552695

EWR # 329234

CRA # 550719

CRA # 557738

ESW-054-0511

EC-Valv-0571

FSAR

Tables 9.2-4 and 9.2-3

Miscellaneous

D107295, Schematic ESW Pump 0P504C

ESW System Health Report

Attachment

A-5

LIST OF ACRONYMS

ALARA As Low As Is Reasonably Achievable

ASME American Society of Mechanical Engineers

CEDE Committed Effective Dose Equivalent

CFR Code of Federal Regulations

CR Condition Report

EAL Emergency Action Level

ESW Emergency Service Water

FSAR [SSES] Final Safety Analysis Report

HP Health Physics

HSM Horizontal Storage Module

HVAC Heating, Ventilation and Air-Conditioning

KV Kilovolts

LDE Lens Dose Equipment

LER Licensee Event Report

NCV Non-cited Violation

NDAP Nuclear Department Administrative Procedure

NRC Nuclear Regulatory Commission

NVLAP National Voluntary Laboratory Accreditation Program

PI [NRC] Performance Indicator

PI&R Problem Identification and Resolution

PPL PPL Susquehanna, LLC

RCIC Reactor Core Isolation Cooling

RG [NRC] Regulatory Guide

RHR Residual Heat Removal

RR Reactor Recirculation

RSPS Risk Significant Planning Standard

SDE Skin Dose Equivalent

SDP Significant Determination Process

SSES Susquehanna Steam Electric Station

TEDE Total Effective Dose Equivalent

TLD Thermoluminescent Dosimeter

VHRA Very High Radiation Area

WO Work Order

Attachment