ML050310062: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
Line 1: Line 1:
{{Adams
#REDIRECT [[IR 05000387/2004005]]
| number = ML050310062
| issue date = 01/28/2005
| title = 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
| author name = Shanbaky M
| author affiliation = NRC/RGN-I/DRP
| addressee name = Shriver B
| addressee affiliation = PPL Generation, LLC
| docket = 05000387, 05000388
| license number = NPF-014, NPF-022
| contact person =
| document report number = IR-04-005
| document type = Inspection Report, Inspection Report Correspondence
| page count = 40
}}
See also: [[see also::IR 05000388/2004005]]
 
=Text=
{{#Wiki_filter: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
                                                                                Enclosure 2
 
                                                4
    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
 
                                                5
  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
 
                                              6
    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 617546, 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
 
                                                7
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
 
                                          8
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
 
                                              9
    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
 
                                              10
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
 
                                              11
    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
 
                                              12
    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
 
                                            13
  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
 
                                                14
    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
 
                                              15
    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 354431                      AR 616048
AR 616053                            AR 616056                      AR 616057
System 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
}}

Revision as of 04:38, 24 March 2020

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML050310062&oldid=2676754"