ML050310062
| 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
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
Enclosures:
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
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
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
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
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
Enclosure 2
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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
Enclosure 2
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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
Enclosure 2
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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
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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"
- 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
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
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,
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
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
RCIC Reactor Core Isolation Cooling
RG [NRC] Regulatory Guide
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