IR 05000247/2009005: Difference between revisions

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{{Adams|number = ML100400200}}
{{Adams
| number = ML100400177
| issue date = 02/09/2010
| title = IR 05000247-09-005, on 10/01/2009 - 12/31/2009, Indian Point Nuclear Generating (Indian Point) Unit 2; Integrated Inspection Report and Notice of Violation
| author name = Gray M K
| author affiliation = NRC/RGN-I/DRP/PB2
| addressee name = Pollock J E
| addressee affiliation = Entergy Nuclear Operations, Inc
| docket = 05000247
| license number = DPR-026
| contact person =
| case reference number = EA-09-296
| document report number = IR-09-005
| document type = Inspection Report, Letter, Notice of Violation
| page count = 46
}}


{{IR-Nav| site = 05000247 | year = 2009 | report number = 005 }}
{{IR-Nav| site = 05000247 | year = 2009 | report number = 005 }}


=Text=
=Text=
{{#Wiki_filter:UNITED NUCLEAR REGULATORY REGION 475 ALLENDALE KING OF PRUSSIA, PENNSYLVANIA February 9, 2010 EA-09-296 Mr. Joseph Site Vice President Entergy Nuclear Operations. Inc. Indian Point Energy Center 450 Broadway. GSB Buchanan, NY 10511-0249 INDIAN POINT NUCLEAR GENERATING UNIT 3 NRC INSPECTION REPORT 05000286/2009005 and NOTICE OF
{{#Wiki_filter:UNITED NUCLEAR REGULATORY REGION 475 ALLENDALE KING OF PRUSSIA, PENNSYLVANIA February 9, 2010 EA-09-296 Mr. Joseph Site Vice President Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB Buchanan, NY 10511-0249 INDIAN POINT NUCLEAR GENERATING UNIT 2 -NRC INTEGRATED INSPECTION REPORT 05000247/2009005 and NOTICE OF VIOLATION (EA-09-296)


==Dear Mr. Pollock:==
==Dear Mr. Pollock:==
On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Unit 3. The enclosed integrated inspection report documents the inspection results, which were discussed on January 21,2010, with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's 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 NRC Enforcement Policy included on the NRC's Web site at www.nrc.gov; select About NRC, How We Regulate, Enforcement, and 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. During the inspection, the NRC identified a violation involving Entergy's submittal of inaccurate information to the NRC related to the medical qualifications of licensed operators. Letters to the NRC certified that the operators had been medically examined and had met all medical qualifications, when, in fact, one test (namely, a tactile test) had not been performed. A tactile test is required to ensure that operators can distinguish among various shapes of control knobs and handles by touch. The test was not performed because your Medical Review Officer (MRO) was unaware that such a test was required. Further, the MRO considered his review of the operators' medical history records for neurological conditions to be sufficient to verify the operators' ability to feel, manipulate, and distinguish plant components when needed. Violations involving the provision of inaccurate or incomplete information are of particular concern to the NRC, and may be considered for escalated enforcement under certain circumstances. However, in this case, the NRC has classified this violation at Severity Level IV, after considering the guidance set forth in Section IV.A.3 of the Enforcement Policy because the inaccurate information did not invalidate the NRC licensing since all of the operators subsequently passed a tactile test when Entergy administered it shortly after the NRC identified the violation. Further, the actual and potential safety significance of this violation was very low in that the Medical Review Officer had conducted a neurological evaluation, albeit not a tactile test, and the operators had been observed successfully manipulating control knobs and handles by Entergy and NRC personnel in the conduct of their licensed duties. Nonetheless, this violation demonstrates the importance of taking all of the necessary steps and conducting all of the necessary reviews to assure that information submitted to the NRC is complete and accurate in all material respects. Although this violation has been placed in your corrective action program, a Notice of Violation is being issued and a response is being required to better understand: 1) what actions were taken in 2004 in response to NRC Information Notice (IN) 2004-20, "Recent Issues Associated with NRC Medical Requirements for Licensed Operators," which, in part, reminded facility licensees that licensed operators and the personnel who perform and interpret their medical examinations need to be familiar with the regulatory requirements and guidelines (it should be noted that this IN specifically described an instance in which a facility licensee had not conducted some tests required in the ANSI standard for any of its licensed operators); 2) why appropriate action was not taken in response to IN 2004-20 to identify appropriate tactile testing was being conducted; and 3) the corrective actions taken and planned at this time to assure all information submitted to the NRC is complete and accurate in alf material respects. 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. Based on the results of this inspection, this report also documents five additional findings of very , low safety significance. Three of these findings were determined to be violations of NRC I requirements. However, because of their very low safety significance, and because the findings were entered into your corrective action program, the NRC is treating these findings as violations (NCVs) consistent with Section VI. A. 1 of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a written response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory CommiSSion, AnN.: 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, DC 20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 3. In addition, if you disagree with the characterization of any finding. you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point Nuclear Generating Unit 3. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice, II a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report documents the inspection results, which were discussed on January 21, 2010 with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.


Sincerely,Projects Branch Division of Reactor Docket No. License No. DPR-64
Based on the results of this inspection, the NRC has determined that a Severity Level IV violation of NRC requirements occurred.


===Enclosures:===
The violation was evaluated in accordance with the NRC Enforcement Policy included on the NRC's Web site at www.nrc.gov; select About NRC, How We Regulate, Enforcement, and 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. During the inspection, the NRC identified a violation involving Entergy's submittal of inaccurate information to the NRC related to the medical qualifications of licensed operators.
(1) Notice of Violation (2) Inspection Report No. 05000286/2009005  
 
Letters to the NRC certified that the operators had been medically examined and had met all medical qualifications, when, in fact, one test (namely, a tactile test) had not been performed.
 
A tactile test is required to ensure that operators can distinguish among various shapes of control knobs and handles by touch. The test was not performed because your Medical Review Officer {MRO) was unaware that such a test was required.
 
Further, the MRO considered his review of the operators'
medical history records for neurological conditions to be sufficient to verify the operators'
ability to feel, manipulate, and distinguish plant components when needed. Violations involving the provision of inaccurate or incomplete information are of particular concern to the NRC, and may be considered for escalated enforcement under certain circumstances.
 
However, in this case, the NRC has classified this violation at Severity Level IV, after considering the guidance set forth in Section IV.A.3 of the Enforcement Policy because the inaccurate information did not invalidate the NRC licensing since all of the operators subsequently passed a tactile test when Entergy administered it shortly after the NRC identified the violation.
 
Further, the actual and potential safety significance of this violation was very low in that the Medical Review Officer had conducted a neurological evaluation, albeit not a tactile test. and the operators had been observed successfully manipulating control knobs and handles by Entergy and NRC personnel in the conduct of their licensed duties. Nonetheless, this violation demonstrates the importance of taking all of the necessary steps and conducting all of the necessary reviews to assure that information submitted to the NRC is complete and accurate in all material respects.
 
Although this Violation has been placed in your corrective action program. a Notice of Violation is being issued and a response is being required to better understand:
1) what actions were taken in 2004 in response to NRC Information Notice (IN) 2004-20, "Recent Issues Associated with NRC Medical Requirements for licensed Operators,fl which. in part, reminded facility licensees that licensed operators and the personnel who perform and interpret their medical examinations need to be familiar with the regulatory requirements and guidelines (it should be noted that this IN specifically described an instance in which a facility licensee had not conducted some tests required in the ANSI standard for any of its licensed operators);
2) why appropriate action was not taken in response to IN 2004-20 to identify appropriate tactile testing was being conducted; and 3) the corrective actions taken and planned at this time to assure all information submitted to the NRC is complete and accurate in all material respects.
 
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.
 
Based on the results of this inspection, this report also documents three additional findings of very low safety significance.
 
All of these findings were determined to be violations of NRC requirements.
 
However, because of their very low safety significance, and because the findings were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report. you should provide a written response within 30 days of the date of this inspection report. with the basis for your denial. to the Nuclear Regulatory CommiSSion, AnN.: Document Control Desk, Washington D.C.
 
with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory CommiSSion, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2. In addition, if you disagree with the characterization of any finding, you should provide a response within 30 days of the date of this inspection report, with the basis for your dlsagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point Nuclear Generating Unit 2. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS). accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html.
 
To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction. ! * ! I
 
Sincerely.
 
Projects Branch 2 Division of Reactor Projects Docket No.
 
License No. DPR-26 Enclosure 1: Notice of Violation Enclosure 2: Inspection Report No. 05000247/2009005  


===w/Attachment:===
===w/Attachment:===
Supplemental Information cc w/encl: Distribution via ListServ r I In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document or from the NRC's document system (ADAMS). accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html. To the extent possible, your response should not include any personal prrvacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
Supplemental Information cc w/enc!: Distribution via ListServ I nATF I OFFICE NAME DATE J. Poliack 3 In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html.
 
To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
 
Distribution w/encl: (via Ewmail) S. Collins, RA (R10RAMAIL RESOURCE)
M. Oapas, ORA (R10RAMAIL RESOURCE)
O.lew,ORP (R1DRPMAIL RESOURCE)
J. Clifford, DRP (R1DRPMAIL RESOURCE)
D. Roberts, DRS (R1DRSMaii Resource)
P. Wilson, DRS (R1DRSMaii Resource)
L. Trocine, RI OEDO M. Gray, DRP B. Bickett, DRP S. McCarver, DRP SUNSI Review Complete:
__
__
 
Sincerely,IRA! Mel Gray, Chief Projects Branch 2 Division of Reactor Projects M. Osborn, DRP E. Keighley, ORP K. Mangan, DRP, Acting SRI A. Ayegbusi, DRP t RI D. Hochmuth, DRP RidsNrrPMlndianPoint Resource RidsNrrDorlLpl1-1 Resource ROPreport Resource@nrc.gov (Reviewer'S Initials)
ML 100400177 DOCUMENT NAME: G:\DRP\BRANCH2\a -Indian Point 2\lnspection Reports\IP2 IR2009-005\IP2 2009 005rev3.doc After declaring this document "An Official Agency Record" it will be released to the Public. To recelvQ a copy of thii document, indicate In the box: 'C'" Copy without attachmentlenclosure
'E'" Copy with attachmentlenclosure
'N':; No copy RI/DRP RI/DRS I RI/DRP I RllEnf I I RI/RC I KManQanlKM BBicketU BB MMcLaughlinJ MMM MGray/MG KFarrar/KF 021 en 110 02/03 f10 02/04 110 0210R 110 02104/10 RIIDRP I I I I I DLew/DL 02/08/10 1 NOTICE OF VIOL ATION Entergy Nuclear Operations, Inc. Indian Point Unit 2 and Unit 3 Delcket No. 50-247 &50-286 License Nos. DPR-26 and DPR-64 EA-09-296 During an NRC inspection conducted from October 19 through October 22, 2009, a violation of NRC requirements was identified.
 
In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR 50.9 requires, in part, that information provided to the Commission by an applicant for a license or by a licensee or information r l 9quired by statute or by the Commission's regulations, Orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.
 
10 CFR 55.21 requires, in part, that an applicant for a license shall have a medical examination by a physician and the licensee shall have a medical examination by a physician every two years. The physician shall determine that the applicant or lic.:msee meets requirements of Section 55.33(a)(1}.
10 CFR 55.33(a)(1}
requires, in part, that an applicant's medical condition and general health will not adversely affect the performance of aSSigned operator job duties or cause operational errors endangering public health and safety. 10 CFR 55.23 requires, in part, that to certify the medical fitness of the applicant, an authorized representative of the facility licensee shall complete and sign NRC Form-396, "Certification of Medica! Examination by Facility Licensee." NRC Form*396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant and that the guidance contained in American National Standards Institute/American Nuclear Society (ANSIIANS)
Standard 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants" was followed in conducting the examination and making the determination of medical qualification.
 
ANSIIANS 3.4-1983, Section 5.4 provides specific minimum capacities required for medical qualifications.
 
Section 5.14 requires, "Tactile discrimination sufficient to distinguish among various shapes of control knobs and handles by touch." Contrary to the above, from October 20,2004 through October 22, 2009, Entergy Nuclear Operations, Inc. (Entergy)
provided information to the NRC that was not complete and accurate in all material respects.
 
Specifically.
 
Entergy had not completed medical examinations of licensed operators in accordance with ANSI/ANS 3.4-1983.
 
The licensee submitted numerous NRC Form-396s for renewal of senior reactor operator and reactor operator licenses and for initial license applicants that certified that the applicants met the medical requirements of ANSI/ANS 3.4-1983 when, in fact" tactile testing had not been conducted.
 
This is a Severity Level IV violation (Supplement VII) . . Pursuant to the provisions of 10 CFR 2.201. Entergy Nuclear Operations.
 
Inc. is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional EnClosure 1
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; EA-09-296" 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, 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 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.


Sincerely. IRAJ Mel Gray, Chief Projects Branch 2 Division of Reactor Projects Distribution w/encl: (via E-mail) S. Collins, RA (R10RAMAIL RESOURCE) M. Osborn, ORP M. Dapas, DRA (R10RAMAIL RESOURCE) E. Keighley, DRP D, Lew, DRP (R1 DRPMAIL RESOURCE) p, Cataldo, SRI, IP3 J. Clifford, DRP (R1DRPMAIL RESOURCE) M. Harter, ORP. RI, IP3 D. Roberts, DRS (R1DRSMaii Resource) D. Hochmuth, ORP P. Wilson, DRS (R1DRSMaii Resource) RidsNrrPMlndianPoint Resource L. Trocine, RI OEDO RidsNrrDorlLpl1-1 Resource M. Gray, DRP ROPReportsResources@nrc.gov B, Bickett, DRP S. McCarver. ORP SUNSI Review Complete: __:.;bb:<..-__(Revlewer's Initials) ML 100400200 DOC NAME: G;\ORP\BRANCH2\a -Indian Point 3\lnspection Reports\2009\IP3 005\IP3.2009,005.r3.doc After declaring this document "An'Official Agency Record" it will be released to the Public. To receive acopy of this document, indicate In the box: 'C' =Copy without atlachmenllenclosure 'EO =Copy with attachment/enclosure 'N'=Nocopy RIIDRS RI/RC PCataldo/BB for KFarrar/KF 02/04/10 02/05/10 RIIORP NAME OLewfDL DATE 02/08/10 NOTICE OF VIOLATION Entergy Nuclear Operations, Inc. Docket No. 50-247 & 50-286 Indian Point Unit 2 and Unit 3 License Nos. DPR-26 and DPR-64 EA-09-296 During an NRC inspection conducted from October 19 through October 22,2009, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR 50.9 requires, in part, that information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, Orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects. 10 CFR 55.21 requires, in part, that an applicant for a license shall have a medical examination by a physician and the licensee shall have a medical examination by a physician every two years. The physician shall determine that the applicant or licensee meets requirements of Section 55.33(a)(1). 10 CFR 55.33(a)(1) requires, in part, that an applicant's medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. 10 CFR 55.23 req uires, in part, that to certify the medical fitness of the applicant, an authorized representative of the facility licensee shall complete and sign NRC Form-396. "Certification of Medical Examination by Facility licensee." NRC Form-396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant and that the guidance contained in American National Standards Institute/American Nuclear Society (ANSI/ANS) Standard 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants" was followed in conducting the examination and making the determination of medical qualification. ANSI/ANS 3.4-1983, Section 5.4 provides specific minimum capacities required for medical qualifications. Section 5.14 requires, "Tactile discrimination sufficient to distinguish among various shapes of control knobs and handles by touch." Contrary to the above, from October 20. 2004 through October 22. 2009, Entergy Nuclear Operations, Inc. (Entergy) provided information to the NRC that was not complete and accurate in all material respects. Specifically, Entergy had not completed medical examinations of licensed operators in accordance with ANSIIANS 3.4-1983. The licensee submitted numerous NRC Form-396s for renewal of senior reactor operator and reactor operator licenses and for initial license applicants that certified that the applicants met the medical requirements of ANSIIANS 3.4-1983 when, in fact, tactile testing had not been conducted. This is a Severity Level IV violation (Supplement VII). Enclosure 1 Pursuant to the provisions of 10 CFR 2.201, Entergy Nuclear Operations, Inc. is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 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, 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 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 NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.govireading-rmiadams.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 9th day of February, 2010. Enclosure 1 Docket No.: License No.: Report No.: Licensee: Facility: Location: Dates: Inspectors: Approved By: U.S. NUCLEAR REGULATORY COMMISSION REGION I 50-286 DPR-64 05000286/2009005 Entergy Nuclear Northeast (Entergy) Indian Point Nuclear Generating Unit 3 450 Broadway, GSB Buchanan, NY 10511-0249 October 1, 2009 through Decem ber 31, 2009 P. Cataldo, Senior Resident Inspector -Indian Point 3 M. Halter. Resident Inspector -Indian Point 3 J. Commiskey, Health Physicist G. Newman, Reactor Inspector J. D'Antonio, Senior Operations Engineer C. Crisden, Emergency Preparedness Specialist D. Orr, Senior Reactor Inspector T. Fish. Senior Operations Engineer J. Schoppy, Senior Reactor Inspector Mel Gray, Chief Projects Branch 2 Division of Reactor Projects Enclosure 2 TABLE OF
Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS). accessible from the NRC Web site at http://www.nrc.gov/reading-rmiadams.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 protectetd and a redacted copy of your response that deletes such information, If you request withholding of such material, you !!llJ.&sect;! 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 9 th day of February 2010. Enclosure 1
Docket No.: License No.: Report No.: Licensee:
Facility:
Location:
Dates: Inspectors:
Approved By: 1 U.S. NUCLEAR REGULATORY COMMISSION REGION I 50-247 DPR-26 05000247/2009005 Entergy Nuclear Northeast (Entergy)
Indian Point Nuclear Generating Unit 2 450 Broadway, GSB Buchanan, NY 10511-0249 October 1, 2009 through December 31, 2009 G. Malone, Senior Resident Inspector  
-Indian Point 2 O. Ayegbusi, Resident Inspector  
-Indian Point 2 P. Cataldo, Senior Resident Inspector  
-Indian Point 3 J. D'Antonio, Senior Operations Engineer S. Barr. Senior Emergency Prep Inspector J. Commiskey.
 
Health Physicist C. Crisden, Emergency Preparedness Specialist T. Fish. Senior Operations Engineer J. Lilliendahl, Reactor linspector K. Mangan, Senior Reactor Inspector J. Nicholson, Health Physicist J. Schoppy, Senior Reactor Inspector Mel Gray, Chief Projects Branch 2 Division of Reactor Projects Enclosure 2
2


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
..........................................................................................................REPORT DET AI LS .....................................................................................................................REACTOR SAFETy ............................................................................................................1R01 Adverse Weather Protection ...................................................................................1R04 Equipment Alignment ..............................................................................................1R05 Fire Protection ........................................................................................................1R06 Flood Protection Measures .....................................................................................1R07 Heat Sink Performance ...........................................................................................1R11 licensed Operator Requalification Program .......................................................... 1R12 Maintenance Effectiveness ..................................................................................1R13 Maintenance Risk Assessments and Emergent Work Control ............................... 'IR15 Operability Evaluations .........................................................................................1R18 Plant Modifications ...............................................................................................1R19 Post-Maintenance Testing .................................................................................... 1R22 Surveillance Testing .............................................................................................1EP2 Alert and Notification System (ANS) Evaluation ..................................................... 1 EP3 Emergency Response Organization (ERO) Staffing and Augmentation System ... 1EP4 Emergency Action Level (EAL) and Emergency Plan Changes ............................1EP5 Correction of Emergency Preparedness Weaknesses ........................ , ................. 1EP6 Drill Evaluation ......................................................................................................RADIATION SAFETy ........................................................................................................20S1 Access Control to Radiologically Significant Areas ...............................................20S2 ALARA Planning and Controls ..............................................................................OTHER ACTIVITIES .........................................................................................................40A1 Performance Indicator Verification .......................................................................40A2 Identification and Resolution of Problems .............................................................40A3 Event Follow-up ....................................................................................................40A5 Other Activities ......................................................................................................40A6 Meetings, including Exit .........................................................................................40A7 licensee-Identified Violations ................................................................................SUPPLEMENTAL INFORMATION .................................................................................. KEY POINTS OF CONTACT ............................... " .......................................................... LIST OF ITEMS OPENED, CLOSED AND DISCUSSED ................................................. LIST OF DOCUMENTS REVIEWED ................................................................................ LIST OF ACRONyMS ....................................................................................................  
IR 05000247/2009005; 10/01/2009
-12131/2009;
Indian Point Nuclear Unit 2; Licensed Operator Requalification Program; Alert and Notification System (ANS) Evaluation;
Event Follow-Up; and Other Activities.
 
This report covered a three-month period of inspection by resident and region based inspectors.
 
Four finding of very low significance (Green) were identified.
 
The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process." The cross-cutting aspects for the findings were determined using IMC 0305, "Operating Reactor Assessment Program." Findings for which the significance determination process (SOP) does not apply may be Green, or be assigned a severity level (SL) after NRC management review. The NRC's program for overseeing safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.
 
===Cornerstone: Mitigating===
 
Systems SL IV. An NRC-identified Severity Level IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators.
 
The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by tOUCh) from the required licensed operator medical examinations.
 
The inspectors determined that Entergy's medical physician did not adequately test all licensed operators (both initial and renewal licensees)in accordance with 10 CFR 55.21 and 10 CFR 55.33 with respect to ANSlIANS-3.41983.


3 SUMMARY OF FINDINGS IR 05000286/2009005; 10/01/2009 -1213112009; Indian Point Nuclear Generating (Indian Point) Unit 3; Licensed Operator Requalification Program; Operability Evaluations; Alert and Notification System (ANS) Evaluation; Identification and Resolution of Problems; and Event Follow-Up. This report covered a three-month period of inspection by resident and region based inspectors. One Severity Level (SL IV) violation and five additional findings of very low safety significance (Green) were identified. Three of these findings were also determined to be non-cited violations (NCVs) of NRC requirements. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process." Findings for which the significance determination process (SOP) does not apply may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006,
However, Entergy had submitted medical information, as required by 10 CFR 55 for licensed operators and applicants that stated the testing had been performed satisfactorily.


===Cornerstone: Initiating Events ===
Following identification of the issue, Entergy entered the issue into the corrective action program (CR-IP3-2009-04487)and completed corrective actions to develop and administer an appropriate test. The inspectors noted that all licensed operators passed this new test and no new license conditions were required.
Green: A self-revealing finding (FIN) of very low safety significance was identified because Entergy personnel did not ensure adequate maintenance was conducted on the 31 and 32 main boiler feed pumps (MBFPs). Specifically, the inspectors determined that Entergy maintenance personnel did not implement maintenance procedures and utilize available vendor information to ensure the 32 MBFP coupling installation was conducted with appropriate tolerances; 32 steam generator water level (SGWL) controller reset times were not set appropriately; and high pressure governor valve stroke settings for 31 MBFP were appropriate. These maintenance performance issues in combination contributed to plant transients including an unplanned power reduction and an automatic reactor trip. The finding was more than minor because the finding was associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, maintenance performance issues resulted in reliability challenges to the non-safety related feedwater pumps and resulted in unplanned plant transients, The inspectors evaluated the finding using IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings,* and determined the finding was of very low safety Significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because Entergy personnel did not ensure effective supervisory and management oversight of maintenance and design control activities regarding the MBFPs. (H.4(c) per I MC0305) (Section 40A3)


===Cornerstone: Mitigating Systems SL IV, An NRC-identified SL IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators. The inspectors identified Entergy submittals to the NRC were===
Entergy's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9. Because this violation of 10 CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dis positioned using the traditional enforcement process. The finding was more than minor because documents which provided the information to the NRC were signed under oath by the company medical physician and the Site Vice President.


inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by touch) from the required licensed operator medical examinations. The inspectors determined that Entergy's medical physician did not adequately test all licensed operators (both initial and renewal licensees) in accordance with 10 CFR 55.21 and 10 CFR 55.33 with respect to ANSIIANS-3.41983. However, Entergy had submitted medical information, as required by 10 CFR 55 for licensed operators and applicants that stated the testing had been performed satisfactorily. Following identification of the issue, Entergy personnel entered the issue into the corrective action program (CR-IP3-2009-04487) and completed corrective actions to develop and administer an appropriate test. The inspectors noted that all licensed operators passed this new test and no new license conditions were required. Entergy's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9. Because this violation of 10 CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dispositioned using the traditional enforcement process, The finding was more than minor because documents which provided the information to the NRC were signed under oath by the company medical physician and the Site Vice President. Because there was no evidence that operators mis-operated equipment due to omitted tactile tests, the finding was determined to be of very low safety significance (SL IV). The applicability of cross-cutting aspects related to the performance deficiency of this finding will be determined after NRC review of Entergy's response to the Notice of Violation, (Section 1 R11.2)
The applicability of cross-cutting aspects related to the performance deficiency of this finding will be determined after NRC review of Entergy's to the Notice of Violation. (Section 1 R11.2)
: '''Green.'''
: '''Green.'''
An NRC-identified non-cited violation (NCV) of very low safety significance of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," was identified because Entergy personnel did not promptly identify and correct a condition adverse to quality regarding molded-case circuit breaker (MCCB) nonconformance. Specifically, in 2004, Entergy personnel determined that a population of MCC8s in safety related applications were beyond the design life as specified in Westinghouse Technical Bulletin, TB-04-13. However, Entergy's scheduled replacement timeframe (through 2011) for those affected safety related MCCBs was not consistent with the safety significance of the issue or adequately supported through an engineering justification considering, at that time, a number of the MCCBs were in service for greater than the 20-year design life. The finding was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the reliability of the electrical distribution system to respond to initiating events to prevent undesirable consequences. Specifically, the MCCB breakers that exceeded their expected design life could impact their reliability to respond to design basis events and plant transients. The inspectors determined the finding was of very low safety significance because the finding was a design qualification deficiency confirmed not to result in loss of operability or fUnction. Specifically, no actual loss of function could be attributed to operating with MCCBs greater than 20 years in service and the inspectors' review of an Entergy operability determination concluded the MCCBs were an operable but nonconforming condition.
A self-revealing non-cited violation (NCV) of very low safety significance of 10 CFR 50, Appendix B Criterion V "Instructions, Procedures, and Drawings," was identified because Entergy personnel did not perform work regarding replacement of a control room digital recorder.
 
As a result, during performance of the work, personnel inadvertently shorted a live wire resulting in a partial loss of control room indications and alarms related to the safety relief valve acoustic monitor flow indications, low range steam and feed flow indications, and inadvertent control rod movement.
 
Entergy personnel reset the breakers to restore control room indications and entered this issue into the corrective action program as 04860. Personnel subsequently replaced the digital recorder with the circuit breaker opened to eliminate the electrical hazard. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiatino events to prevent undesirable consequences.
 
Specifically, the grounded recorder power supply resulted in a loss of control room indications and alarms that could have impacted operations response to an event. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,>>
and determined it to be of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance related to work practices.
 
Specifically, Entergy personnel did not follow procedures during the replacement of a control room digital recorder.


This finding has a cross-cutting aspect in the area of Problem Identification and Resolution because on several occasions Entergy personnel did not thoroughly evaluate MCCB qualification issues including operability and functionality considerations. This included an opportunity to evaluate the condition in 200B when engineers identified residue indicative of grease breakdown. (P.1 (c) per IMC0305) (Section 40A2) Green: An NRC-identified NCVof very low safety significance of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions,* was identified because Entergy personnel did not adequately correct a condition adverse to quality to ensure the continued operability of emergency diesel generators (EDGs). Specifically, Entergy personnel did not ensure that contacts associated with EDG jacket water pressure switches for the air start systems were in the appropriate state following EDG operations to support EDG restart. Additionally, after identification of the spectfic cause, Entergy personnel did not implement continuity checks on the EDGs to ensure continued operability after EDG operation in a timely manner. The finding was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 33 EDG incurred unavailability hours and reliability was impacted during EDG standby conditions with one air start sub-system available. The inspectors determined the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of safety function, and was not risk significant with respect to external events. The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution because Entergy personnel did not implement adequate corrective actions to address continued EDG operability concerns with degraded jacket water pressure switches in a timely manner. (P.1(d) per IMC0305) (Section 1 R15) Green: A self-revealing finding (FIN) of very low safety significance was identified because Entergy personnel did not perform adequate post-maintenance functional testing to ensure 6.9kV breakers were able to perform intended safety functions. Specifically, in July 2009, during a planned maintenance activity, maintenance personnel installed a 6.9kV breaker without adequate post-maintenance testing. As a result, on August 10, 2009, following an automatic plant trip, a 6.9kV breaker failed to operate due to a bent lever and prevented the automatic transfer of a 4BO-Volt safety bus from its onsite electrical power source to its appropriate off-site electrical sources, as required. The finding was more than minor because the finding is associated with the procedure quality performance attribute of the Mitigating Systems cornerstone and affected the associated cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding was considered to be of very low safety significance (Green) in accordance with IMC 0609, Attachment 4, "Phase 1 -Initial Screening and Characterization of Findings," because the finding was not a design or qualification deficiency, did not result in the loss of a safety function, and was not risk significant due to external events.
[H.4(b) per IMC 0305] (Section 40A3.2)  


6 The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because Entergy personnel did not ensure adequate planning (work control) was performed to ensure post-maintenance functional testing was appropriate for the 6.9kV bustle breakers. [H.3(a) per IMC0305]. (Section 40A3)
===Cornerstone: Emergency===


===Cornerstone: Emergency Preparedness ===
Preparedness
: '''Green.'''
: '''Green.'''
A seff-revealing NCVof very low safety significance of 10 CFR 50.47(b)(5) was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be degraded for 56 days. The siren technicians did not use a detailed written procedure or work instruction to perform siren file updates, but instead relied on performing the task from memory. As a result, on September 16, 2009, Entergy conducted a full volume siren test during which a total of 18 sirens indicated a failure to function. Entergy personnel entered the siren failures into their corrective action process for resolution and performed a root cause of the event to determine the short and long term corrective actions. The finding was more than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of facilities and equipment, and impacted the cornerstone objective of ensuring that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated using 'MC 0609 Appendix B, "Emergency Preparedness Significance Determination Process (SOP)" and was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by siren technicians [H.4(c) per IMC 0305] (Section 1 EP2)  
A self-revealing NCV of very low safety significance of 10 CFR 50.47(b}(5}
was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be degraded for 56 days. The siren technicians did not use a detailed written procedure or work instruction to perform siren file updates, but instead relied on performing the task from memory. As a result, on September 16, 2009, Entergy conducted a full volume siren test durinu which a total of 18 sirens indicated a failure to function.
 
Entergy entered the siren failures into their corrective action process for resolution and performed a root cause of the event to determine the short and long term corrective actions. The finding was more than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of facilities and equipment, and impacted the cornerstone objective of ensuring that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.
 
This finding was evaluated using IMC 0609 Appendix B, "Emergency Preparedness Significance Determination Process (SOP)" and was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by siren technicians  
[H.4(c) per IMC 0305J (Section 1EP2)


===Other Findings===
===Other Findings===
Violations of very low safety significance. which were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and corrective actions are listed in Section 40A7 of this report.
SL IV, An NRC-identified Severity Level IV, NCV of 10 CFR 72.212(b)(2)(ii), was identified because Entergy personnel did not evaluate a change to the written evaluation described in its Holtec Updated Final Safety Analysis Report (UFSAR) prior to implementing the change. Specifically, inspectors identified that Entergy personnel were storing combustible material on the Independent Spent Fuel Storage Installation (ISFSI) pad which was contrary to the Holtec UFSAR and the Entergy 72.212 Evaluation Report which stated that transient combustibles will not be stored on the ISFSI pad. Following the inspectors'
questions, Entergy personnel determined the required evaluation in accordance with the requirements of 10 CFR 72.48(c) was not performed.


.1 7
Entergy personnel entered the issue into their corrective action program and verified that all combustibles had been removed from the pad. The Reactor Oversight Process (ROP) was not used for this finding because inspections of ISFSI activities are covered under NRC Manual Chapter 2690 and are not incorporated in the reactor safety cornerstones in the ROP's Significance Determination Process (SOP). It was determined that the failure to evaluate a change to the written evaluation required by 10 CFR 72.212 using the requirements of 10 CFR 72.48{c) was a performance deficiency that was reasonably within Entergy's ability to foresee and prevent The finding was determined to be a Severity Level IV violation based on Supplement VI, Example 0.2 of the NRC Enforcement Policy. A cross-cutting aspect was not assigned since the performance deficiency was applicable to evaluation in accordance with the ROP. (Section


=REPORT DETAILS=
=REPORT DETAILS=
Summary of Plant Status Indian Point Unit 3 operated at or near full reactor power (100%) throughout the inspection period.
Summary of Plant Status Indian Point Unit 2 began the inspection period operating at full reactor power (100%). On November 2, Unit 2 shutdown due to an automatic reactor trip due to a turbine-generator protective trip resulting from a loss of the generator exciter power supply. On November "1, operators returned the plant to 100% power. Unit 2 remained at or near full power during the remainder of the inspection period.


==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity  
Cornerstones:
{{a|1R01}}
Initiating Events, Mitigating Systems, and Barrier Integrity 1 R01 Adverse Weather Protection (71111.01 -1 sample)
==1R01 Adverse Weather Protection==
{{IP sample|IP=IP 71111.01|count=1}}
Cold Weather Preparedness


====a. Inspection Scope====
===.1 Station Readiness===
The inspectors performed a detailed review of Entergy procedures to address seasonal cold weather conditions. This review included an evaluation of deficiencies identified during the current seasonal preparations. and that adverse conditions were being adequately addressed to ensure the cold weather conditions would not have significant impact on plant operation and safety. The inspectors conducted plant and system walkdowns of the refueling water storage tank, the auxiliary feedwater building, service water intake structure. and the control building. Additionally, the inspectors conducted the review to verify that the station's implementation of OAP-OOB, "Severe Weather Preparations," and OAP-04B, "Seasonal Weather Preparation," appropriately maintained systems required for normal operation and safe shutdown conditions. The inspection satisfied one inspection sample for the seasonal weather preparations.
 
for Extreme Cold Conditions Inspection Scope The inspectors reviewed the readiness of risk-significant systems for winter cold weather conditions.
 
The inspectors reviewed Entergy's adverse weather procedures, operating experience.
 
corrective action program, UFSAR, Technical Specifications (TS). operating procedures.
 
and applicable plant documents to determine the types of adverse weather challenges to which the site is susceptible.
 
The inspectors also checked local area temperatures.
 
as well as the operability of ventilation and heating systems. to ensure the plant was prepared for cold weather conditions.
 
In addition, the following risk-significant systems that were required to be protected from adverse weather conditions were selected and collectively represented one inspection sample: Motor driven and turbine driven auxiliary feedwater system; Diesel generator fire pump; and 21, 22 and 23 emergency diesel generators (EDGs).


====b. Findings====
====b. Findings====
No findings of significance were identified. 1
No findings of significance were identified.
{{a|R04}}
{{a|1R04}}
==R04 Equipment Alignment==
==1R04 Eguipment==
{{IP sample|IP=IP 71111.04Q|count=3}}
 
Partial System Walkdowns
Alignment (71111.04Q -3 samples)
 
===.1 Partial System Walkdowns===


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed partial system walkdowns to inspect Entergy staffs performance in maintaining the proper equipment alignment of redundant or diverse trains and components during periods of system train unavailability. and where applicable, following return to service after maintenance. The inspectors referenced system procedures, the Updated Final Safety Analysis Report (UFSAR), and system drawings to verify that the alignment of the applicable system or component supported its required safety functions. The inspectors also reviewed applicable condition reports (CRs) or work orders (WOs) to ensure Entergy personnel identified and properly addressed equipment deficiencies that could potentially impair the capability of the available train(s). The documents reviewed during this inspection are listed in the Attachment. The inspectors performed partial walkdowns of the following systems or components, which represented three inspection samples: Enclosure 2
The inspectors performed partial system walkdowns to verify the operability of redundant or diverse trains and components during periods of system train unavailability or following periods of maintenance.


===.1 31 and 32 EDG during 33 EDG outage on November 2; 31 and 33 service water (SW) pump while 32 SW pump was out of service for maintenance on November 18; and 31 and 33 motor-driven auxiliary boiler feedwater pumps (ABFP) during maintenance on 32 turbine-driven ABFP on November 23.
The inspectors referenced system procedures.
 
UFSAR. and system drawings to verify the alignment of the available train suppolted its required safety functions.
 
The inspectors also reviewed applicable condition reports Enclosure
 
===.1 (CRs) and work orders to ensure Entergy personnel===
 
identified and properly addressed equipment discrepancies that could potentially impair the capability of the available train, as required by 10 CFR 50, Appendix B, Criterion XVI, Action." The documents reviewed during these inspections are fisted in the Attachment.
 
The inspectors performed a partial walkdown on the following systems, which represented three inspection samples: 22 EDG after planned outage; 22 residual heat removal (RHR) train when 21 RHR pump was out of service; and EDG fuel oil system following testing.


====b. Findings====
====b. Findings====
No findings of significance were identified . . Complete System Walkdown===
No findings of significance were identified . . Full System Walkdown (71111.04S  
{{IP sample|IP=IP 71111.04S|count=1}}
-1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a complete system walkdown of accessible portions of the safety injection (SI) system, to determine whether the eXisting equipment alignment was consistent with the required alignment for the current plant conditions. The inspectors reviewed operating procedures, surveillance tests, drawings, equipment lineup check-off lists, and the UFSAR, to determine if the SI system was appropriately aligned to perform its required safety functions. The inspectors reviewed a sample of CRs that were generated to address deficiencies associated with the SI system, and verified that these deficiencies were appropriately evaluated and/or resolved within the corrective action program. The documents reviewed during this inspection are listed in the Attachment. The walkdown of the SI system represented one inspection sample.
The inspectors performed a complete system walkdown of accessible portions of the component cooling water (CCW) system to identify discrepancies between the existing equipment lineup and the required lineup. The inspectors reviewed operating procedures, surveillance tests, piping and instrumentation drawings, equipment lineup check-off lists, and the UFSAR to verify the system was aligned to perform its required safety functions.
 
The inspectors reviewed a sample of eRs written to address deficiencies associated with the system to ensure they were appropriately evaluated and resolved.
 
The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. 1Fire Protection (71111.05Q -4 samples) Quarterly Fire Area Walkdowns
No findings of significance were identified. Fire Protection (71111.050 -5 samples) Resident Inspector Quarterly Walkdowns


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors conducted tours of selected Unit 3 fire areas to assess the material condition and operational status of applicable fire protection features. The inspectors reviewed, consistent with the applicable administrative procedures, whether: combustible material and ignition sources were adequately controlled; passive fire barriers, manual fire-fighting eqUipment, and suppression and detection eqUipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program. The inspectors also evaluated the fire protection program for conformance with the requirements of License Condition 2. K. The documents reviewed during this inspection are listed in the Attachment. Enclosure 2 This inspection represented four inspection samples and was conducted in the areas addressed by the following Pre-Fire Plans (PFP): PFP-306, Primary auxiliary building; PFP-307A1B, Charging Pump areas; PFP-385, Circulating and SW pump building; and PFP-352, Cable spreading/battery rooms.
The inspectors conducted tours of several fire areas to assess the material condition and operational status of fire protection features.
 
The inspectors verified, consistent with the applicable administrative procedures, that: combustibles and ignition sources were adequately controlled; passive fire barriers, manual firE!.fighting equipment, and suppression and detection equipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program. The inspectors evaluated the fire protection program for conformance with the requirements of License Condition 2.K. The documents reviewed during this inspection are listed in the Enclosure
.
This inspection represented five inspection samples for fire protection tours, and was conducted in the following areas: ISFSI pad area; Fire Zone (FZ) 25. 23 battery room; FZ 15 control room; FZ 90A, 91A spent fuel pool area; and FZ 252 cable spreading room.


====b. Findings====
====b. Findings====
No findings of significance were identified. Flood Protection Measures (71111.06 -1 sample)
No findings of significance were identified.
 
1 Heat Sink Performance (71111.07T -3 Samples)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the Unit 3 Individual Plant Examination, the UFSAR, and RPT-06-00071, "Indian Point Unit 3 Probabilistic Safety Assessment (PSA)," concerning internal flooding events. The inspectors assessed flood mitigation attributes within the ABFP building that are utilized to minimize potential impacts of flooding on the ABFPs and feedwater control valves. The inspectors also reviewed a surveillance test associated with the fire protection system to verify operators would have indication of system actuation. This inspection represented one sample for internal flood protection measures.
Based on a plant specific risk assessment, past inspection results, and resident inspector input, the inspectors selected the following heat exchanger samples: 22 CCW heat exchanger; 23 EDG jacket water and lube oil heat exchangers; and Ultimate heat sink (UHS), which included operation of the service water system and UHS. The inspectors reviewed whether potential common cause heat sink performance problems were identified and corrected by the licensee.
 
The inspectors also reviewed potential macro fouling (silt, debris, etc.) issues and biotic fouling issues to verify the issues were closely examined by Entergy personnel.
 
In response to Generic Letter 13, "Service Water System Problems Affecting Safety-Related Equipment," Entergy committed to performing frequent periodic cleaning of essential service water heat exchangers in lieu of testing for degraded performanCE!.
To ensure adequate implementation of Generic Letter 89-13 commitments, the inspectors reviewed Entergy's inspection, cleaning, and eddy current testing methods and frequency with the responsible system engineer.
 
The inspectors compared surveillance test and inspection data, including as found conditions and eddy current summary sheets, to the established acceptance criteria to verify that the results were acceptable and that system heat exchanger operation was consistent with design. The inspectors reviewed heat exchanger design basis values and assumptions, plugging limit calculations, and vendor information.
 
to verify whether Entergy personnel incorporated the information into the heat exchanger inspection and maintenance procedures.
 
The inspectors walked down the intake area, portions of the service water system, including the service water pump and strainer pits, CCW heat exchangers, and EDG heat exchangers, to assess the material condition and operational functioning of these systems and components.
 
The inspectors reviewed a sample of condition reports related to the service water system to ensure that station personnel were appropriately identifying, characterizing, and correcting problems related to these systems and components.
 
The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
===.1 9===
No findings of significance were identified. Heat Sink Performance (lP 71111.0n -2 samples) a. I nspection Scope Based on a plant specific risk assessment, previous inspections, recent operational experience, and resident inspector input, the inspectors selected the following areas for review: Operation and performance testing of the SW system; Performance of the ultimate heat sink (UHS), which included SW piping integrity and SW intake structure functionality; and EDG and component cooling water (CCW) heat exchangers (HXs). The inspectors reviewed the SW system design to evaluate the adequacy of system monitoring and performance testing. The inspectors reviewed a sample of SW pump and valve performance tests, system health and walkdown reports. and in-service test (1ST) vibration monitoring results for adverse trends and to verify that the system functioned as designed. The inspectors verified that Entergy personnel performed the pump and valve ISTs in accordance with American Society of Mechanical Engineers (ASME) Code requirements. In addition, the inspectors reviewed Entergy's monitoring, Enclosure 2 maintenance, and testing of interface valves between safety-related SW and non-safety related or non-seismic piping systems to ensure that adequate SW flow is available accident consistent wlth design basis assumptions. The inspectors reviewed Entergy's buried pipe inspection and monitoring program to independently assess the condition and structural integrity of the SW piping. The inspectors reviewed a risk-informed sample of Entergy's disposition of active wall pipe leaks, including completed or planned corrective actions and structural evaluations. The inspectors reviewed a sample of SW pipe nondestructive examination (NDE) records including ultrasonic tests, radiographic tests, visual tests, and available videos to ensure that Entergy personnel appropriately identified and dispositioned SW leakage or degradation. The inspectors performed an above ground walkdown of accessible areas containing buried SW piping to look for soil subsidence or other indications of piping leakage and/or degradation. The inspectors also directly observed the condition of SW piping in the accessible portions of the valve pits. The inspectors reviewed Entergy's procedures and processes to control macro fouling of the SW system. The inspectors also observed the condition of the SW bio-boxes used to monitor zebra mussel activity. The inspectors reviewed Entergy's procedures for SW and intake structure operation, abnormal SWoperations, adverse weather conditions, cold weather preparations, and for a loss of the SW system. The inspectors reviewed whether Entergy maintained these procedures consistent with their design and licensing basis and that plant operators could reasonably implement the procedures as written. The inspectors independently verified that SW and intake level instrumentation, which operators rely upon for decision making, was available and functional. The inspectors walked down control room instrument panels, the EDG and CCW HXs, accessible portions of SW piping in the reactor and turbine buildings, and SW intake area (including the SW pumps, strainers, and traveling water screens) to assess the material condition and configuration control of these structures, systems and components (SSCs). On December 9,2009, the inspectors performed an additional focused walkdown of the SW intake area to assess SW system functionality during adverse weather conditions. The inspectors also reviewed a sample of corrective action CRs related to the SW isolation valves, SW pumps, and SW piping integrity to ensure that Entergy appropriately identified, characterized, and corrected problems related to these essential SSCs. A list of documents reviewed is provided in the Attachment to this report.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified.
{{a|1R11}}
 
==1R11 Licensed Operator Regualification Program Quarterly Resident Inspector Evaluation (71111.11 Q w 1 sample)==
1Licensed Operator Requalmcation Program (71111.11 Q -1 sample) Quarterly Review


====a. Inspection Scope====
====a. Inspection Scope====
On October 5, 2009, the inspectors observed annual licensed operator requalification training examinations conducted in the plant-reference simulator, to verify appropriate operator performance, and that evaluators identified and documented crew performance Enclosure 2
On October 6, the inspectors observed licensed operator simulator training.


===.1 11 problems. as applicable. The inspectors evaluated the performance of risk significant operator actions. including the use of emergency operation procedures. The inspectors assessed the clarity and the effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operations, and the oversight and direction provided by the control room supervisor. The inspectors reviewed simulator fidelity to verify correlation with the actual plant control room, and to verify that differences in fidelity that could potentially impact training effectiveness were either identified or appropriately dispositioned. Licensed operator training was evaluated for conformance with the requirements of 10 CFR 55, "Operator Licenses." The documents reviewed during this inspection are listed in the Attachment. This observation of operator evaluations represented one inspection sample.===
which included simulated steam generator instrumentation failures and a large break coolant-accident (LBLOCA) coincident with the failure of several plant systems to automatically respond to adverse conditions, to verify operator performance was adequate and evaluators were identifying and documenting crew performance problems.
 
The inspectors evaluated the performance of risk-significant operator actions including the use of emergency operating procedures.
 
The assessed the clarity and effectiveness of communications, implementation of actions in response to alarms. performance of timely control board operation and manipulation, and the oversight and direction provided by the control room supervisor.
 
The inspectors also assessed simulator fidelity with respect to the actual plant. The inspectors evaluated licensed operator training for conformance with the requirements of 10 CFR 55, "Operator Licenses." The documents reviewed during this inspection are listed in the Attachment.
 
This observation of operator simUlator training represented one inspection sample.


====b. Findings====
====b. Findings====
No findings of significance were identified . . 2 Licensed Operator Requalification (71111.11 B-1 sample)
No findings of significance were identified . . Licensed Operator Requalification (71111.11 B-1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The following inspection activities were performed using NUREG 1021, "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement 1, and Inspection Procedure Attachment 71111.11, "Licensed Operator Requalification Program." The inspectors conducted a review of recent operating history documentation found in inspection reports, licensee event reports, and Enterg}ls corrective action program. The inspectors also reviewed specific events from Entergy's corrective action program which indicated possible training deficiencies, to verify that they had been appropriately addressed. The resident inspector staff was also consulted for inSights regarding licensed operators' performance. . The remediation plans for a crew or individual's performance were reviewed by inspectors to assess the effectiveness of the remedial training. Operators, instructors and training/operation's management were interviewed for feedback on their training program and the quality of training received. The inspectors observed simulator performance during the conduct of the examinations, and reviewed simulator discrepancy reports to verify facility staff were in compliance with the requirements of 10 CFR 55.46. Inspectors also reviewed a sample of simulator tests including transient, steady state, and malfunction tests. The inspectors reviewed a sample of records for requalification training attendance, program feedback, reporting, and medical examinations for compliance with license conditions, including NRC regulations. The operating tests for the weeks of September 21, October 5, and October 19, 2009. were reviewed for content, quality, and overlap. Likewise, three 2009 comprehensive written exams were reviewed for content, quality, and overlap. Enclosure 2 The inspectors observed the training staff administer operating exams to one shift crew during the onsite inspection week. which began October 19. The inspectors also observed post-scenario evaluations, and monitored exam security practices. On December 15,2009. the inspectors reviewed results of Indian Point Unit 32009 comprehensive written and annual operating tests to determine whether pass/fail rates were consistent with the guidance of NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process (SDP)." Inspectors verified the following: Crew failure rate on the dynamic simulator was less than (Failure rate was Individual failure rate on the dynamic simulator test was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the walkthrough test Gob performance measures) was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the 2009 comprehensive written exam was less than or equal to 20%. (Failure rate was 4.5%); and More than 75% of the individuals passed all portions of the exam (95.5% of the individuals passed all portions of the exam).
On December 15, the inspectors reviewed results of the 2009 comprehensive written and annual operating tests to determine whether pass/fail rates were consistent with the guidance of NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process (SOP)." Inspectors verified the following: Crew failure rate on the dynamic simulator was less than (Failure rate was Individual failure rate on the dynamiC simulator test was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the walkthrough test (job performance measures)was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the 2009 comprehensive written exam was less than or equal to 20%. (Failure rate was 0.0%): and More than 75% of the individuals passed all portions of the exam (100% of the individuals passed all portions of the exam). Enclosure


====b. Findings====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
An NRC-identified Severity Level IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators. The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by tOUCh) from the required licensed operator medical examinations.
An Severity Level IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators.
 
The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by touch) from the required licensed operator medical examinations.


=====Description:=====
=====Description:=====
The NRC's requirements related to the conduct and documentation of medical examinations for operators are contained in Subpart C, Medical Requirements, of 10 CFR 55, Operators' Licenses. Specifically, 10 CFR 55.21, Medical Examination, requires every operator be examined by a physician when he or she first applies for a license and every two years, thereafter, once the license is received. The medical examination is performed in order for the physician to determine whether the operator meets the requirements of 10 CFR 55.33(a)(1). The physician is to verify that the operator's medical condition and general health will not adversely affect the performance of assigned operator duties or cause operational errors that endanger public health and safety. The facility licensee (Entergy) must also certify which industry standard (Le., the 1983 or 1996 version of ANSIIANS-3.4. Medical Certification and Monitoring or Personnel Requiring Operator Licenses for Nuclear Power Plants. or other NRC-approved method) was used in making the fitness determination. For the medical examination performed for licensed operators at Indian Point Units 2 and 3, the inspectors determined that Entergy had stated on NRC Form 396 that the 1983 industry standard was used for the completion of the medical examination. The inspectors noted that ANSI-3.4 1983. Paragraph 5.4.14 "Neurological," requires licensed operators to have discrimination (Stereognosis) sufficient to distinguish among various shapes of control Enclosure 2 13 knobs and handles by touch." Additionally, the inspectors identified that the Form 396 was signed by both the medical review officer and Site Vice President, under oath, verifying the examination had been performed. During the medical records review, the inspectors determined that Entergy personnel had not been conducting tactile testing of its licensed operators. This omission had the potential for being significant since, during a transient aggravated by limited visibility, operators may be required to perform actions relying on their ability to distinguish, by touch, between different shapes of operating switches and knobs. Following identification of the issue Entergy personnel completed corrective actions to develop and administer an appropriate test. The inspectors noted that all licensed operators passed this new test, and no new license conditions were required.
The NRC's requirements related to the conduct and documentation of medical examinations for operators are contained in Subpart C, Medical Requirements, of 10 CFR 55, Operators' Licenses.
 
Specifically, 10 CFR 55.21, Medical Examination, requires every operator be examined by a physician when he or she first applies for a license and every two years, thereafter, once the license is received.
 
The medical examination is performed in order for the physiCian to determine whether the operator meets the requirements of 10 CFR 55.33(a)(1).
 
The physician is to verify that the operator's medical condition and general health will not adversely affect the performance of assigned operator duties or cause operational errors that endanger public health and safety. The facility licensee (Entergy)must also certify which industry standard (Le., the 1983 or 1996 version of ANSI/ANS-3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, or other NRC-approved method) was used in making the fitness determination.
 
For the medical examination performed for licensed operators at Indian Point Units 2 and 3, the inspectors determined that Entergy had stated on NRC Form 396 that the 1983 industry standard was used for the completion of the medical examination.
 
The inspectors noted that ANSI-3.4 1983, Paragraph 5.4.14 "Neurological," requires licensed operators to have "Tactile discrimination (Stereognosis)sufficient to distinguish among various shapes of control knobs and handles by touch." Additionally, the inspectors identified that the Form 396 was signed by both the medical review officer and Site Vice President, under oatl'i, verifying the examination had been performed.
 
During the medical records review, the inspectors determined that Entergypersonnel had not been conducting tactile testing of its licensed operators.
 
This omission had the potential for being significant since, during a transient aggravated by limited visibility, operators may be required to perform actions relying on their ability to distinguish, by touch, between different shapes of operating switches and knobs. Following identification of the issue Entergy personnel completed corrective actions to develop and administer an appropriate test. The inspectors noted that aI/licensed operators passed this new test, and no new license conditions were required.


=====Analysis:=====
=====Analysis:=====
The inspectors determined that a long-standing deficiency had existed at the Indian Point Units 2 and 3 in that the licensee's medical physician was not adequately testing all licensed operators (both initial and renewal licensees) in accordance with 10 CFR 55.21 and 55.33 with respect to ANSIIANS-3.4 1983. 10 CFR 55.23 requires that an authorized representative of the facility licensee shall certify the medical fitness of an applicant by completing and signing an NRC Form 396. NRC Form 396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant as required in 10 CFR 55.21, and that the guidance contained in ANSIIANS-3.4 1983 was followed in conducting the examination and making the determination of medical qualification. The licensee's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action by the NRC and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9 .and the issue was within the licensee's ability to foresee and prevent. Because a violation of 10 CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dispositioned using the traditional enforcement process. The finding was more than minor because the document which provided the information was provided to the NRC signed under oath by the company medical doctor and the site vice president. Because there was no evidence that operators mis-operated equipment due to omitted tactile tests, the finding was determined to be of very low safety significance (SL IV). The applicability of aspects related to the performance deficiency of this finding will be determined after NRC review of Entergy's response to the Notice of Violation.
The inspectors determined that a long-standing deficiency had existed at the Indian Point Units 2 and 3 in that the licensee's medical phYSician was not adequately testing all licensed operators (both initial and renewal licensees ) in accordance with 10 CFR 55.21 and 55.33 with respect to ANSI/ANS-3.4 1983. 10 CFR 55.23 requires that an authorized representative of the facility licensee shall certify the medical fitness of an applicant by completing and signing an NRC Form 396. NRC Form 396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant as required in 10 CFR 55.21, and that Enclosure the guidance contained in ANSI/ANS-3.4 1983 was followed in conducting the examination and making the determination of medical qualification.
 
The licensee's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action by the NRC and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9 and the issue was within the licensee's ability to foresee and prevent. Because a violation of 1 (I CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dispositioned using the traditional enforCElment process. The finding was more than minor because the document which provided the information was provided to the NRC signed under oath by the company medical doctor and the site vice president.
 
Because there was no evidence that operators mis-operated equipment due to omitted tactile tests, the finding was determined to be of very low safety significance (SL IV). The applicability of cross-cutting aspects related to trie performance deficiency of this finding will be determined after NRC review of Entergy's response to the Notice of Violation.


=====Enforcement:=====
=====Enforcement:=====
10 CFR 50.9 states, in part, "Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects." Contrary to this, from October 20, 2004 through October 22, 2009, Entergy submitted inaccurate information to the NRC on NRC Form 396 regarding the medical certification and testing of its licensed operators and initial applicants. This information was material to the NRC because the NRC relied on this certification to determine whether the applicant met the requirements to operate the controls of a nuclear power plant pursuant to 10 CFR 55. Enclosure 2 This issue has been entered into the facility corrective action program 04487) and is of very low safety significance. The licensee implemented immediate corrective action and satisfactorily performed the required test. The inspectors verified the adequacy and promptness of the licensee's corrective actions for the medical issue. These corrective actions included the development of a tactile test which required operators to identify by touch various control knobs and switch shapes within a bag. The new tests were administered to all licensed operators and senior licensed operators. All operators passed the test and no new deficiencies were identified. This violation is being treated consistent with other licensed operator medical examination findings and the NRC Enforcement Policy. (NOV 05000286/2009005*01, Incomplete Licensed Operator Medical Examinations) 1 R Maintenance Effectiveness (71111.12 -2 samples)
10 CFR 50.9 states, in part, "Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects." Contrary to this, from October 20,2004 through October 22,2009, Entergy submitted inaccurate information to the NRC on NRC Form 396 regarding the medical certification and testing of its licensed operators and initial applicants, This information was material to the NRC because the NRC relied on this certification to determine whether the applicant met the requirements to operate the controls of a nuclear power plant pursuant to 10 CFR 55. This issue has been entered into the facility corrective action program 04487) and is of very low safety significance.
 
The licensee implemented immediate corrective action and satisfactorily performed the required test. )"he inspectors verified the adequacy and promptness of the licensee's corrective actions for the medical issue. These corrective actions included the development of a tactile test which required operators to identify by touch various control knobs and switch shapes within a bag. The new tests were administered to all licensed operators and senior licensed operators.
 
All operators passed the test and no new deficiencies were identified, This violation is being treated consistent with other licensed operator medical examination findings and the NRC Enforcement Policy. (NOV 05000247/2009005*01, Incomplete Licensed Operator Medical Examinations)1 R12 Maintenance Effectiveness (71111.12Q -1 sample) a. Inspection" Scope The inspectors reviewed performance-based problems that involved structures, systems, and components (SSCs) to assess the effectiveness of maintenance activities.
 
When applicable, the reviews focused on: Enclosure Proper maintenance rule scoping in accordance with 10 CFR 50.65; Characterization of reliability issues; Changing system and component unavailability; 10 CFR 50.65(a)(1)and (a)(2) classification; Identifying and addressing common cause failures; Trending of system flow and temperature values; and Appropriateness of performance criteria for SSGs classified (a)(2). The inspectors also reviewed the system health report, maintenance backlogs, and maintenance rule basis document.
 
The inspectors evaluated maintenance effectiveness and monitoring activities against the requirements of 10 CFR 50.65. The documents reviewed during this inspection are listed in the Attachment.


====a. Inspection Scope====
The following component was reviewed and represented one inspection sample: Appendix R diesel generator coolant compatibility.
The inspectors reviewed performance-based problems that involved selected structures, systems, and components (SSCs) to assess the effectiveness of maintenance activities and to verify activities were conducted in accordance with site procedures and 10 CFR 50.65 (The Maintenance Rule). The reviews focused on: Evaluation of Maintenance Rule scoping and performance criteria; Verification that reliability issues were appropriately characterized; Verification of proper system and/or component unavailability; Verification that Maintenance Rule (a)(1) and (a)(2) classifications were appropriate; Verification that system performance parameters were appropriately trended; For SSCs classified as Maintenance Rule (a)(1), that goals and associated corrective actions were adequate and appropriate for the circumstances; and Identification of common cause failures. The inspectors also reviewed system health reports, maintenance backlogs, and Maintenance Rule basis documents. The documents reviewed during this inspection are listed in the Attachment. The following systems and/or components were reviewed and represented two inspection samples: Service water discharge valve disc erosion; and 32 main feed water regulating valve deficiencies.


====b. Findings====
====b. Findings====
No findings of significance were identified. 1 Maintenance Risk Assessments and Emergent Work Control (71111.13 -4 samples)
No findings of significance were identified. 1 R Maintenance Risk Assessments and Emergent Work Control (71111.13 -4 samples)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed maintenance activities to verify that the appropriate on-line risk assessments were performed prior to removing equipment for work as required by  
The inspectors reviewed scheduled and emergent maintenance activities to verify that the appropriate risk assessments were performed prior to removing equipment from service for maintenance or repair. The inspectors reviewed selected risk assessments to verify assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete.
 
When emergent work was performed, the inspectors reviewed the plant risk to ensure risk was promptly reassessed and managed. Documents reviewed during this inspection are listed in the Attachment.


15 CFR 50.65(a)(4). When planned work scope or schedules were altered to address emergent or unplanned conditions, the inspectors verified that the plant risk was promptly reassessed and managed by station personnel. The documents reviewed during this inspection are listed in the Attachment. The following activities represented four inspection samples: Planned risk during 138kV switchyard work regarding 33332 L&M line restoration on October 8; Planned risk during Undervoltage/Degraded Grid testing on October 22; Planned risk during 33 EDG maintenance on November 2; and Planned risk during emergency boration valve stroke testing on November 10.
The following activities represented four inspection samples: Emergent maintenance associated with the Appendix R diesel generator concurrent with maintenance on 138kV line 33332 L&M. power range nuclear instrumentation recalibrations, and preventative maintenance on the 22 containment spray pump on October 2; Planned maintenance associated with the 23 CCW pump and preventative maintenance on the 21 and 22 safety injection (SI) and RHR pump motor breakers on October 20; Planned maintenance associated with the 33332 L&M line, 23 CCW pump and 22 SW pump following an inadvertent trip of the 22 EDG output breaker on October 26; and Unplanned maintenance outage associated with the 22 EDG on NovembEtr 9 and 10.


====b. Findings====
====b. Findings====
No findings of significance were identified. 1 R Operability Evaluations (71111.15 -5 samples)
No findings of significance were identified.
 
===.1 1 R15 Operability===
 
Evaluations (71111.15 -2 samples) Resident Quarterly Review


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed operability evaluations to assess the acceptability of the evaluations, the use and control of compensatory measures when applicable, and compliance with Technical Specifications. These reviews were conducted to verify that operability determinations were performed in accordance with procedure ENN-OP-104, "Operability Determinations." The inspectors assessed the technical adequacy of the evaluations to ensure consistency with the UFSAR and associated design and licensing basis documents. The documents reviewed are listed in the Attachment The following operability evaluations were reviewed and represented five inspection samples: CR-IP3-2009-03911, stab from compartment 7FMLlFMR from B phase missed bus bar; CR-IP3-2009-04123, Neutron Flux Detector N-38 failing; CR-IP3-2009-04165, Through-wall leak on SW line upstream of SWN-213; CR-IP3-2009-04351104362, EDG east air start motor failure and potential kilowatt overload; and 32 ABFP oil, degraded bearing, and vibration issues.
The inspectors reviewed operability evaluations to assess the acceptability of the evaluations, the use and control of compensatory measures, when applicable, and compliance with Technical Specifications (TS). The inspectors' reviews included verification that operability determinations were performed in accordance with procedure 04, "Operability Determinations." The inspectors assessed the technical adequacy of the evaluations to ensure consistency with the TS, UFSAR, and associated design basis documents.
 
The documents reviewed are listed in the Attachment.
 
The following operability evaluations were reviewed and represented two inspecti<m samples:
* 21 EDG day tank level Indication; and
* 22 EDG jacket water heater breaker failure.


====b. Findings====
====b. Findings====
I ntroduction: An NRC-identified NCV of very low safety significance (Green) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," was identified because Entergy personnel did not adequately correct a condition adverse to quality to ensure the continued operability of EDGs. Specifically, Entergy personnel did not ensure that contacts associated with EDG jacket water pressure switches for the air start systems were in the appropriate state following EDG operations to support EDG restart.
No findings of significance were identified. \ 1 R19 Post-Maintenance Testing (71111.19 -8 samples)


=====Description:=====
====a. Inspection Scope====
On November 4, 2009, during post-work testing on the 33 EDG, the 33 EDG failed to start on demand due to air start motor that had not operated properly. Subsequent troubleshoottng by Entergy personnel resulted in the discovery on November 14, that the direct cause of the air start motor failure was pressure switch PS-Enclosure 2 16 2282, a jacket water pressure switch associated with the air start system, did not change to the appropriate state as expected. The 33 EDG air start system utilizes two pressure switches, one for each air start motor, to ensure the air start motors operate and allow for proper startup and operation of the EDG. PS-2282 is associated with the east air start motor, and utilizes jacket cooling water system pressure to operate as the 33 EDG comes up to speed. Contact No. 1 of the pressure switch, which is open during EDG operation, is required to close as the pressure switch responds to decreasing pressure in the jacket water system at a predetermined setpoint. For this November 4 event, Entergy personnel determined that the pressure switch had not fully returned to its at-rest state following a preceding EDG performance run during post-maintenance testing, which should have placed contact No. 1 in the desired closed position to ensure the 33 EDG was ready for operation upon demand, whether in manual or automatic mode. The inspectors identified that previous failures of jacket water pressure switches occurred in 2008 at the station, and that Entergy personnel previously determined the cause to be inadequate contact material selection resulted in micro-welding of closed contacts during operation coupled with electrical circuit deficiencies and possibly setpoint drift. Subsequently, newer models with increased trip setpoints associated with three of five jacket water pressure switches on all three EDGs were installed by station personnel in 2008 as a result of these pressure switch design issues and contact material deficiencies. However, the two remaining pressure switches associated with the air start motors (two per EDG for a total of six pressure switches) were not replaced by Entergy personnel at that time due to station considerations regarding planned EDG air receiver design and licensing basis changes that were unrelated to the switch problem. The inspectors evaluated the operability of the 33 EDG following the discovery of both the original air start motor failure on November 4. and Entergy personnel's identification of the direct cause ofthe failure on November 14. The inspectors concluded that the operability of the EDGs were not assured as a result of this known, degraded condition of the jacket water pressure switches without implementation of appropriate compensatory corrective actions. Specifically, the inspectors identified that Entergy personnel had previously recognized a potential, failure mode due to a condition adverse to quality (contact micro-welding) associated with EDG air start motor operation that would not be self-revealing until an EDG demand start was required. The inspectors noted that no failures had occurred in the jacket water pressure switches associated with the air start system in recent history prior to the failure on November 4. However, the inspectors concluded the reliability of the 33 EDG was reduced because only one air start sub-system was available between November 4 and November 11, due to the east air start motor being isolated to support troubleshooting. Additionally, between November 14 and December 17, the inspectors identified Entergy personnel did not effectively implement corrective actions to assure that the degraded switches on the air start system were in a state to support an emergency start of the EDGs. On December 17, Entergy personnel instituted measures to ensure continued EDG operability following EDG operation. These measures included post-run verification that the contacts associated with the air start systems were in the appropriate position to ensure subsequent EDG operation, when needed. Additionally, Entergy personnel implemented corrective actions to install improved pressure switches following resolution Enclosure 2
The inspectors reviewed post-maintenance test procedures and associated testing activities for selected risk-significant mitigating systems, and assessed whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel.


===.1 17 of licensing basis considerations associated with the EDG air receiver. These corrective actions are detailed in CR-IP3-2009-04819.
The Inspectors verified that: test acceptance criteria were clear and the test demonstrated operational readiness consistent with design basis documentation; test instrumentation had current calibrations with the appropriate range and accuracy for the application; and the tests were performed as written, with applicable prerequisites satisfied.


=====Analysis:=====
Upon completion of the tests, the inspectors reviewed whether equipment was returned to the proper alignment necessary to perform its safety function.
The inspectors determined that station personnel did not implement corrective measures in a timely manner for a degraded EDG air start system pressure switch condition and that this constituted a performance deficiency. The inspectors determined this issue was more than minor because it affected the equipment performance attribute of the Mitigating System cornerstone and affected its objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 33 EDG incurred unavailability and reliability was impacted during 33 EDG standby conditions with one air start SUb-system available prior to Entergy's implementation of appropriate compensatory measures. The inspectors evaluated the finding using IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings.n The inspectors determined the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of safety function, and was not risk significant with respect to external events. The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution because Entergy personnel did not implement adequate corrective actions to address continued EDG operability concerns with degraded jacket water pressure switches in a timely manner. (P.1{d) per IMC0305)


=====Enforcement:=====
Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix B, Criterion XI, "Test Control." The documents reviewed are listed in the Attachment.
10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that the conditions adverse to quality, such as deficiencies and defective material and equipment are promptly identified and corrected. Contrary to the above, Entergy personnel did not correct a condition adverse to quality associated with jacket water pressure switches or implement corrective measures in a timely manner. Specifically, Entergy personnel did not implement actions in 2008 to replace jacket water pressure switches or take actions from November 4 through December 17, 2009, to ensure that contacts for the 33 EDG pressure switches for the air start systems were in the appropriate state following EDG operation. Because this violation is of very low safety significance and has been entered into the Entergy's corrective action program, 2009-01550, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000286/2009005w02, Untimely Compensatory Measures for Degraded EDG Pressure Switches) 1R18 Plant Modifications===
{{IP sample|IP=IP 71111.18|count=1}}
33 Emergency Diesel Generator East Side Jacket Water Heater Disabled


====a. Inspection Scope====
The following post-maintenance testing activities were reviewed and represented eight inspection samples:
The inspectors reviewed the design documentation associated with the temporary disabling of the east side jacket water heater. The inspectors verified the adequacy of the temporary modification and reviewed the associated temporary procedure changes. This verification included review of the system impacts of reduced heating capacity during standby and review of operator round records to ensure jacket water temperature was within specificatrons to support an emergency start of the diesel generator. The inspectors also reviewed the work package that installed this temporary modification. Enclosure 2
* 23 CCW pump overhaul;
* Motor operated valve (MOV) SI-18526 motor and actuator overhaul;
* Starting air system maintenance and output breaker inspection on the 22 EDG;
* Preventative maintenance of the 24 fan cooler unit (FCU) service water flange;
* Service water valve MOV SWN-41 16 motor and actuator overhaul;
* Cable pull and repair splicing of the L&M 33332 line;
* Internal inspection of EDG 23 heat exchangers; and
* Replacement of diesel fire pump reHef valve. Enclosure


====b. Findings====
====b. Findings====
No findings of significance were identified. 1
No findings of significance were identified.
{{a|R19}}
 
==R19 Testing==
1 Surveillance Testing (71111.22
{{IP sample|IP=IP 71111.19|count=5}}
-6 samples)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed post-maintenance test procedures and associated testing activities for selected risk-significant mitigating systems, and assessed whether the effect of maintenance on plant systems was adequately addressed by control room and plant personnel. The inspectors verified that: test acceptance criteria were clear; tests demonstrated operational readiness and were consistent with design basis documentation; test instrumentation had current calibrations and appropriate range and accuracy for the application; tests were performed as written; and applicable test prerequisites were satisfied. Upon completion of the tests, the inspectors verified whether equipment was returned to the proper alignment necessary to perform its safety function. Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix 6, Criterion XI, "Test ControL" The following post-maintenance activities were reviewed and represented five inspection samples:
The inspectors observed performance of portions of surveillance tests and/or reviewed test data for selected risk-significant structures, systems, and components (SSCs) to assess whether tests satisfied TS, UFSAR, Technical Requirements Manual, and Entergy procedure requirements.
* 32 SW pump motor replacement on October 1;
 
* 33 Charging pump repack on October 4;
The inspectors verified that: test acceptance criteria were clear, demonstrated operational readiness, and were consistent with design basis documentation; test instrumentation had accurate calibration, and appropriate range and accuracy for the application; and tests were performed as written, with applicable prerequisites satisfied.'
* 32 EDG maintenance outage testing on October 7;
Following the tests, the inspectors verified that the equipment was capable of performing the required safety functions.
* Condensate storage tank level switch replacement on October 20; and
 
* 34 FeV (flow control valve) motor lead replacements on November 5.
The inspectors evaluated the surveillance tests against the requirements in TS. The documents reviewed during this inspection are listed in the Attachment.
 
The following surveillance tests were reviewed and represented six inspection samples: Feedwater valves FCV-405 A-D test (1ST); 23 EDG load test; 21 SI pump 1ST; 21 RHR pump 1ST; Condensate storage tank guided wave evaluations of underground portions of the condensate and SW piping; and 23 station battery quarterly surveillance.


====b. Findings====
====b. Findings====
No findings of significance were identified.  
No findings of significance were identified.
{{a|1R22}}
 
==1R22 Surveillance Testing==
===Cornerstone:===
{{IP sample|IP=IP 71111.22|count=5}}
 
Emergency Preparedness (EP) Alert and Notification System fANS) Evaluation (71114.02 -1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk-significant structures, systems, and components, to assess whether test results satisfied Technical Specifications. UFSAR, technical requirements manual, and Entergy procedure requirements. The inspectors verified that: test acceptance criteria were sufficiently clear; tests demonstrated operational readiness and were consistent with design basis documentation; test instrumentation had accurate calibrations and appropriate range and accuracy for the application; tests were performed as written; and applicable test prerequisites were satisfied. Following the tests, the inspectors verified whether equipment was capable of performing the required safety functions. The documents reviewed during this inspection are listed in the Attachment. The following surveillance tests were reviewed and represented five inspection samples, which included an in-service testing (1ST) surveillance: Enclosure 2 3-PT -Q132, Emergency Boration Flow Path Valve CH-MOV-333, on November 11 (1ST): 3-PT-Q080, Pressurizer Block Valve Timing Test RC-MOV-535 and 536, on November 13; 3-PT-M62A. 480-Volt Undervoltage I Degraded Grid Protection System Bus 2A and 3A Functional, on November 19; 0-SOP-lEAKRATE-001, RCS Leakrate Surveillance, Evaluation, and Leak Identification. on November 22; and 3-PT-Q120B, 32 ABFP (Turbine Driven) Surveillance and 1ST, on November 25.
An onsite review was conducted to assess the maintenance and testing of Indian Point Energy Center's (lPEC) current ANS. During the inspection, the Inspector interviewed the Entergy staff responsible for overseeing the ANS testing and maintenance of the system. The inspector reviewed ANS procedures and the ANS design report to ensure Entergy's compliance with design report commitments for system maintenance and testing. The inspector reviewed CRs pertaining to the ANS for causes, trends, and corrective actions. The inspector also reviewed Entergy's root cause report related to siren test results conducted in September 2009. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment


====b. Findings====
===2. Planning Standard, 10 CFR 50.47(b)(5) ===
No findings of significance were identified. 1 Alert and Notification System (ANS) Evaluation (71114.02 -1 sample) a. Inspection ScoQe An onsite review was conducted to assess the maintenance and testing of Indian Point Energy Center's (JPEC) current ANS. During the inspection, the inspector interviewed the Entergy staff responsible for overseeing the ANS testing and maintenance of the system. The inspector reviewed ANS procedures and the ANS design report to ensure Entergy's compliance with design report commitments for system maintenance and testing. The inspector reviewed CRs pertaining to the ANS for causes, trends, and corrective actions. The inspector also reviewed Entergy's root cause report related to siren test resUlts conducted in September 2009. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 2. Planning Standard, 10 CFR 50,47(b)(5) and the related requirements of 10 CFR 50, Appendix E, were used as reference criteria. The documents reviewed during this inspection are listed in the Attachment.
 
and the related requirements of 10 CFR 50, Appendix E, were used Enclosure as reference criteria.
 
The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
A self-revealing NCVof very low safety significance (Green) of 10 CFR 50,47(b)(5) was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be in a degraded condition for 56 days.
A self-revealing NCVof very low safety significance (Green) of 10 CFR 50.47(b)(5)was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be in a degraded condition for 56 days.


=====Description:=====
=====Description:=====
The new ANS siren system is comprised of 172 sirens located throughout the four counties within the 10 mile Emergency Planning Zone (EPZ). Of the 172 sirens, 13 are capable of volce reproduction. The voice enabled sirens are located in areas, such as Harriman State and Croton Point Parks, where the population may not have access to media that would transmit Emergency Alert Messages. The inspector's review of Entergy's root cause evaluations determined that, in July 2009, Entergy received new voice chips along with two data files (one for voice and one for non-voice sirens) along with instructions for installation of the chips and data files frpm Enclosure 2 20 the siren system vendor. The new voice chips and software provided an upgrade to the previous voice message. On July 15, 2009, Entergy personnel discussed the task of installing new voice chips on the digital message boards (OMB) for the 13 voice enabled sirens and installing the updated voice data file for each siren. The first voice chip installation and data file update was performed on July 20, 2009. Although the siren system vendor provided the installation instructions for the data file, the instructions were not included in the Entergy work instructions nor were they provided to the technician performing the upgrade. On July 22, 2009, technicians continued to update all voice sirens with the new voice chip and the new data file. While updating a single voice siren data file, the UPDATE ALL command was inadvertently invoked three times within a short period of time. The technician recognized the error and proceeded to abort the process all three times. A similar data file update error had previously occurred on July 20, 2009. While actions were taken to recover from the error, a CR was not documented and no actions were taken to prevent reoccurrence. Between July 22 and July 29, 2009, the technicians continued to update the remaining voice sirens with the new voice chips and data file with no additional instances of the UPDATE ALL command being invoked. The installation of voice chips and the voice data files was completed on July 29, 2009. All voice sirens were updated and verified with the voice chips and the new data file. The post maintenance testing for this activity would not have identified the latent error with the non-voice enabled sirens because it was not intended to have modified these sirens during this work activity. As a result of the data file update error on July 22, 2009, 14 non-voice sirens were inadvertently configured as voice sirens. After the technician made the file update error on July 22, 2009, the technician did not verify that the correct data files were installed for all non-voice sirens (three non-voice sirens were verified as having the correct files after the July 20. 2009 data update error). This error caused 14 non-voice sirens to be left in a condition where the sirens would function (annunciate); however, the indication at the siren activation points would indicate that the sirens had failed (red-dots versus dot for successful activation). In August 2009, routine polling, silent tests and annual Preventive Maintenance (PM) were conducted by Entergy. The annual PM procedure requires verification if the individual siren's data file is correct for the type of siren (voice or non-voice). During the PMs, several siren data files were found to be incorrect and were corrected during the PM. The last four PMs conducted on non-voice sirens in the August/September timeframe each began with a non-voice siren verification failure. This failure was an indication that the non-voice siren was configured with a voice siren data file. The Entergy Root Causereport determined that the failure should have been identified by the technician and indicated that there was a more significant problem with the siren data files. This problem was neither documented in a CR nor was it reported to management. The silent tests that were conducted would not have identified voice data file configuration errors. On September 16, 2009, Entergy conducted a full volume test of the siren system. Of the 172 sirens activated during the test, 18 siren failures were observed (red dots on displays indicating Siren failures). Of the 18 failures observed, four were reported as amplifier (AMP) failures and 14 were reported as OMB errors. The inspector did not identify a performance deficiency associated with the four AMP siren failures. The  
The new ANS siren system is comprised of 172 sirens located throughout the four counties within the 10 mile Emergency Planning Zone (EPZ). Of the 172 sirens, 13 are capable of voice reproduction.
 
The voice enabled sirens are located in areas, such as Harriman State and Croton Point Parks, where the population may not have access to media that would transmit Emergency Alert Messages.
 
The inspector's review of Entergy's root cause evaluations determined that, in July 2009, Entergy received new voice chips along with two data files (one for voice and one for non-voice sirens) along with instructions for installation of the chips and data files from the siren system vendor. The new voice chips and software provided an upgrade to the previous voice message. On July 15, 2009, Entergy personnel discussed the task of installing new voice chips on the digital message boards (OMS) for the 13 voice enabled sirens and installing the updated voice data file for each siren. The first voice chip installation and data file update was performed on July 20, 2009. Although the siren system vendor provided the installation instructions for the data file, the instructions were not included in the Entergy work instructions nor were they provided to the technician performing the upgrade. On July 22, 2009, technicians continued to update all voice sirens with the new voice chip and the new data file. While updating a single voice siren data file, the UPDATE ALL command was inadvertently invoked three times within a short period of time. The technician recognized the error and proceeded to abort the process all three times. A similar data file update error had previously occurred on July 20, 2009. While actions were taken to recover from the error, a CR was not and no actions were taken to prevent reoccurrence.
 
Between July 22 and July 29, 2009, the technicians continued to update the remaining voice sirens with the new voice chips and data file with no additional instances of the UPDATE ALL command being invoked. The installation of voice chips and the voice data files was completed on July 29, 2009. All voice sirens were updated and verified with the voice chips and the new data file. The post maintenance testing for this activity would not have identified the latent error with the non-voice enabled sirens because it was not intended to have modified these sirens during this work activity.
 
As a result of the data file update error on July 22, 2009. 14 non-voice sirens werH configured as voice sirens. After the technician made the file update error on July 22,2009, the technician did not verify that the correct data files were installed for all non-voice sirens (three non-voice sirens were verified as having the correct files after the July 20, 2009 data update error). This error caused 14 non-voice sirens to be left in Enclosure 16 a condition where the sirens would function (annunciate);
however, the indication at the siren activation points would indicate that the sirens had failed (red-dots versus dot for successful activation).
 
In August 2009, routine polling, silent tests and annual Preventive Maintenance (PM) were conducted by Entergy. The annual PM procedure requires verification if the individual siren's data file is correct for the type of siren (voice or non-voice).
 
During the PMs, several siren data files were found to be incorrect and were corrected during the PM. The last four PMs conducted on non-voice sirens in the August/September timeframe each began with a non-voice siren verification failure. This failure was :an indication that the non-voice siren was configured with a voice siren data file. The Entergy Root Cause report determined that the failure should have been identified by the technician and indicated that there was a more significant problem with the siren data files. This problem was neither documented in a CR nor was it reported to management.
 
The silent tests that were conducted would not have identified voice data file configuration errors. On September 16, 2009, Entergy conducted a full volume test of the siren system. Of the 172 sirens activated during the test, 18 siren failures were observed (red dots on displays indicating siren failures).
 
Of the 18 failures observed, four were reported as amplifier (AMP) failures and 14 were reported as DMS errors. The inspector did not identify a performance deficiency associated with the four AMP siren failures.
 
The 14 DMS errors were due to an incorrect data file being installed for the siren. The sirens indicating an error were non-voice sirens that were installed with the voice data file. According to procedure IP-EP-AD30, IPEG ATI Siren System Administration, maintenance on the siren system will be performed using procedures, step lists, and checklists per IP-EP-AD31, IPEC Siren System Maintenance Administration Procedure.


21 DMB errors were due to an incorrect data file being installed for the siren. The sirens indicating an error were non-voice sirens that were installed with the voice data file. According to procedure IPEC ATI Siren System Administration, maintenance on the siren system will be performed using procedures, step lists, and checklists per IP-EP-AD31, IPEC Siren System Maintenance Administration Procedure. IP-EP-A031 states checklist and procedures will be used if the work is beyond the skill of the craft or the vendor tech manuals. Contrary to IP-EP-A030, the inspectors determined the technician did not use detailed written procedures nor work instructions to perform the siren updates. Instead the technician relied on performing the task from memory. As a result, on September 16,2009,14 OMB failures occurred due to an incorrect data file being installed for the sirens. Troubleshooting testing conducted following the September 16, 2009, full volume test, demonstrated that while the 14 sirens indicated that they had failed to function, the sirens most likely sounded based on this subsequent testing. In the case of a siren indicating failure during an actual event, Entergy would use an installed reverse calling system to notify the affected public. Following the siren test failures, Entergy diagnosed the data file error, installed the correct data file, and had all 14 sirens returned to an operable status on the day of the test. On October 22, 2009, a subsequent full volume test demonstrated 100 percent successful siren activation.
IP-EP-AD31 states checklist and procedures will be used if the work is beyond the skill of the craft or the vendor tech manuals. Contrary to IP-EP-AD30, the inspectors determined the technician did not use detailed written procedures nor work instructions to perform the siren updates. Instead the technician relied on performing the from memory. As a result, on September 16, 2009, 14 DMB failures occurred due to an incorrect data file being installed for the sirens. Troubleshooting testing conducted following the September 16, 2009. full volume test, demonstrated that while the 14 sirens indicated that they had failed to function, the sirens most likely sounded based on this subsequent testing. In the case of a siren indicating failure during an actual event, Entergy would use an installed reverse calling system to notify the affected public. Following the siren test failures, Entergy diagnosed the data file error, installed the correct data file, and had all 14 sirens returned to an operable status on the day of the test. On October 22, 2009, a subsequent full volume test demonstrated 100 percent successful siren activation.


=====Analysis:=====
=====Analysis:=====
The inspector determined that Entergy's failure to use procedures, step lists or checklists while performing maintenance on the siren system was a performance deficiency resulting in approximately 8% of the system to be degraded for 56 days. The finding is greater than minor because it is associated with the emergency preparedness (EP) cornerstone attribute of Facilities and Equipment (Maintenance of Equipment) and affected the EP cornerstone objective of ensuring the capability to implement adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Sheet 1, "Failure to Comply." The finding is associated with the failure to meet or implement a regulatory requirement (planning standard). The finding is not more than Green because it did not result in a Risk Significant Planning Standard (RSPS) function being lost or degraded. The SOP defines degradation of this RSPS to be, "the public alert and notification system (e.g., sirens, other supporting primary notification methods) has design flaws or deficiencies in the test program, maintenance program, or procedures that degrade a portion of the system for a significant period from the time of discovery (e.g., 100% over 25 days, greater than 48% over 45 days, greater than 24% over 90 days, greater than 12% over 6 months)." In this case, approximately 8% of sirens were degraded for over 45 days; therefore, it was concluded that the RSPS was not degraded (as defined by the SOP) and the finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by station personnel and siren technicians (H.4(c)).
The inspector determined that Entergy's failure to use procedures, step lists or checklists while performing maintenance on the siren system was a performance deficiency resulting in approximately 8% of the system to be degraded for 56 days. The finding is greater than minor because it is associated with the emergency preparedness (EP) cornerstone attribute of Facilities and Equipment (Maintenance of Equipment)and affected the EP cornerstone objective of ensuring the capability to implement adequate measures to protect the health and safety of the public in the event of a radiological emergency.
 
This finding was evaluated using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Sheet 1, "Failure to Comply." The finding is associated with the failure to meet or implement a regulatory requirement (planning standard).
 
The finding is not more than Green because it did not result in a Risk Significant Planning Standard (RSPS) function being lost or degraded.
 
The SOP defines degradation of this RSPS to be, "the public alert and notification system (e.g., sirens, other supporting primary notification methods) has design flaws or deficiencies in the test program, maintenance program, or procedures that degrade a portion of the system for a significant period from the time of discovery (e.g., 100% over 25 days, greater than 48% over 45 days, greater than 24% over'90 days, greater than 12% over 6 months)." In this case, approximately 8% of sirens were degraded for over 45 days; therefore, it was concluded that the RSPS was not dewaded (as defined by the SOP) and the finding was determined to be of very low safety significance (Green). This finding has a aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by station personnel and siren technicians (H,4(c>>).  


=====Enforcement:=====
=====Enforcement:=====
1 0 CFR 50.54(q) states in part that the facility licensee shall follow and maintain in effect emergency plans which meet the standards in 50.47(b) and the requirements in Appendix E of this part. Planning Standard 10 CFR 50.47 (b)(5) requires, in part, that a means to provide early notification and clear instruction to the Enclosure 2 populace within the plume exposure pathway EPZ have been established. Contrary to the above, from July 22, 2009 until September 16. 2009. a means to provide early notification and clear instruction to the populace within the plume exposure pathway EPZ had not been established in the areas adjacent to the 14 non-functional sirens. A contributing cause for this violation was the failure to use procedures, step lists or checklists during a siren maintenance activity conducted on July 22, 2009. Because this violation was of very low safety significance and it was entered into Entergy's corrective action program this violation is being treated as an NCV, consistent with Section VI,A.1 of the NRC Enforcement Policy. (NCV 0500028612009005*03, Siren Test Failure) 1 Emergency Response Organization (ERO) Staffing and Augmentation System (71114.03 -1 sample)
1 0 CFR 50.54(q) states in part that the facility licensee shall follow and maintain in effect emergency plans which meet the standards in 50,47(b) and the requirements in Appendix E of this part. Planning Standard 10 CFR 50,47 (b)(5) requires, in part, that a means to provide early notification and clear instruction to the populace within the plume exposure pathway EPZ have been established.
 
Contrary to the above, from July 22, 2009 until September 16, 2009, a means to provide early notification and clear instruction to the populace within the plume exposure pathway EPZ had not been established in the areas adjacent to the 14 non-functional sirens. A contributing cause for this violation was the failure to use procedures, step lists or checklists during a siren maintenance actiVity conducted on July 22, 2009. Because this violation was of very low safety significance and it was entered into Entergy's corrective action program (CR-IP2-2009-3701);
this violation is being treated as an NCV, consistent with Section VLA.1 of the NRC Enforcement Policy. (NCV 05000247/2009005*02, Siren Test Failure) 1 Emergency Response Organization (ERO) Staffing and Augmentation System (71114.03 -1 sample) Inspection Scol2e The inspector conducted a review of IPEC's ERO augmentation staffing requirements and the process for notifying and augmenting the ERO. This was performed to ensure the readiness of key licensee staff to respond to an emergency event and to ensure Entergy's ability to activate their emergency facilities in a timely manner. The inspector reviewed the IPEC ERO roster, sampling of training records, and CRs related to the ERO staffing augmentation system. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment
 
===3. Planning Standard, 10 CFR 50,47(b)(2) ===
 
and related requirements of 10 CFR 50, Appendix E, were used as reference criteria.
 
The documents reviewed during this inspection are listed in the Attachment. Findings No findings of significance were identified.
 
1EP4 Emergency Action Level (EALl and Emergency Plan Changes (71114.04
-1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspector conducted a review of IPEC's ERO augmentation staffing requirements and the process for notifying and augmenting the ERO. This was performed to ensure the readiness of key licensee staff to respond to an emergency event and activate their emergency facilities in a timely manner. The inspector reviewed the IPEC ERO roster, sampling of training records, and CRs related to the ERO staffing augmentation system. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 3. Planning Standard, 10 CFR 50.47(b )(2) and related requirements of 10 CFR 50. Appendix E, were used as reference criteria.
Since the last NRC inspection of this program area, Entergy implemented various changes to different sections of their emergency plan. Entergy had determined that, in accordance with 10 CFR 50.54(q), any change made to the emergency plan, and its lower-tier implementing procedures, had not resulted in any decrease in effectiveness of the plan, and that the revised plan continued to meet tl1e standards in 50.47(b) and the requirements of 10 CFR 50 Appendix E. The inspector reviewed all emergency plan changes, including the changes to lower-tier emergency plan implementing procedures, to evaluate for any potential decreases in effectiveness of the emergency plan. However, this review by the inspector was not documented in an NRC Safety Evaluation Report and does not constitute formal NRC approval of the changes. Therefore, these changes remain subject to future NRC inspection in entirety.
 
The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment  
 
===4. The requirements ===
 
in 10 CFR 50.54(q) were used as reference criteria.
 
The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. 1 Emergency Action Level (EAL) and Emergency Plan Changes (71114.04 -1 sample)
No findings of Significance were identified.
 
1 EP5 Correction of Emergency Preparedness Weaknesses (71114.05 -1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
Since the last NRC inspection of this program area. Entergy personnel implemented changes to different sections of their emergency plan. Entergy personnel had determined that, in accordance with 10 CFR 50.54(q), any change made to the emergency plan, and its lower-tier implementing procedures, had not resulted in any decrease in effectiveness of the plan, and that the revised plan continued to meet the standards in 50.47(b) and the requirements of 10 CFR 50 Appendix E. The inspector reviewed all emergency plan changes, including the changes to lower-tier emergency plan implementing procedures, to evaluate for any potential decreases in effectiveness of the emergency plan. However, this review by the inspectors was not documented in an NRC Safety Evaluation Report and does not constitute formal NRC approval of the changes. Therefore, these changes remain subject to future NRC inspection in their entirety. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 4. The requirements in 10 CFR 50.54(q) were used as reference criteria. The documents reviewed during this inspection are listed in the Attachment.
The inspectors reviewed a sampling of self-assessment procedures and reports to assess Entergy's ability to evaluate their EP performance and programs.
 
The inspectors reviewed a sampling of CRs from December 2007 through November 2009, initiated by Entergyat IPEC from drills and audits. Additionally, the inspectors reviewed 10 CFR 50.54(t) audits; and self*assessment reports. This inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment  
 
===5. Planning Standard, 10 CFR 50.47{b)(14) ===
 
and the related requirements of 10 CFR 50. Appendix E, were used as reference criteria.
 
The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
====b. Findings====
No findings of significance were identified. Enclosure 2 23 'I
No findings of significance were identified.
{{a|EP5}}
 
==EP5 Correction of Emergency Preparedness Weaknesses==
===2. RADIATION ===
{{IP sample|IP=IP 71114.05|count=1}}
 
SAFETY Cornerstone:
Occupational Radiation Safety (OS) 20S1 Access Contro! to Radiologically Significant Areas (71121.01 15 samples)


====a. Inspection Scope====
====a. Inspection Scope====
The inspector reviewed a sampling of procedures and reports to assess Entergy's ability to evaluate their EP performance and programs. The inspector reviewed a sampling of CRs from December 2007 through November 2009, initiated by Entergy at IPEC from drills and audits. Additionally, the inspectors reviewed 10 CFR 50.54(t} audits; and self-assessment reports. This inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 5, Planning Standard, 10 CFR 50.47(b)(14) and the related requirements of 10 CFR 50, Appendix E, were used as reference criteria. The documents reviewed during this inspection are listed in the Attachment.
During September 28 through October 2, 2009, the inspectors conducted activities to verify that Entergy staff at IPEC were properly implementing physical, engineering, and Enclosure 19 administrative controls for access to high radiation areas (HRAs), and other radiologically controlled areas, and that workers were adhering to these controls when working in these areas. Implementation of the access Gontrol program was reviewed against the criteria contained in 10 CFR 20, site technical speCifications, and Entergy's procedures required by the Technical Specifications as criteria for determining compliance.
 
During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The documents reviewed during this inspection are listed in the Attachment.
 
The inspectors performed independent radiation dose rate measurements and reviewed the following items: Plant Walk Downs and RWP Reviews The inspectors reviewed exposure significant work areas within radiation areas, HRAs, and airborne areas in the plant to assess licensee controls and surveys for adequacy.
 
Work reviewed included 3R15 Refueling Outage and On-Une work activities: U2 Rep Platform Entry (Oil AddiUon) U2 Vapor Containment, Replace 21 CRD Fan Motor radiation work permit (RWP) 2009-2033 U2 Fuel Moves, RWP 2009-2043 U2 Dry Cask Storage & Associated Work, RWP 2009-2029 Radiation protection support for locked HRA (LHRA) Entries, RWP 3501 Maintenance Support, RWP 2009-3506 Waste Management, RWP 2009-3504 Scaffolding, RWP 2009-3518 Outage Valve Work, RWP 2009*3520 Reactor Disassembly
& Reassembly, RWP 2009-3521 Split Pin Repair & Associated Work, RWP 2009-3530 RCP Pump & Motor Work, RWP 2009-3534 With a survey instrument and assistance from a Health Physics qualified individual, the inspectors walked down various areas to determine:
whether the RWP, procedure, and engineering controls were in place and whether surveys and postings were adequate.
 
The inspectors reviewed RWPs that provide access to exposure-significant areas of the plant. Specified electronic personal dosimeter alarm set points were reviewed by inspectors with respect to current radiological condition appllcability and workers were queried to verify their understanding of plant procedures governing alarm response and knowledge of radiological conditions in their work area. The inspectors determined there were no RWPs for airbome radioactivity areas with the potential for individual worker internal exposures of >50 millirem (mrem) committed effective dose equivalent (CEDE). Additionally, the inspectors determined there were no internal dose assessments that resulted in actual internal exposures greater than 50 mrem CEDE. Enclosure Problem Identrfication and Resolution The inspectors reviewed access control-related eRs generated since the last inspection in this area was conducted.
 
Staff members were interviewed and documents reviewed to determine that follow-up activities are being conducted in an effective and timely manner, commensurate with their safety and risk. For repetitive deficiencies or significant individual deficiencies in problem identification and resolution, the inspectors determined jf the licensee's assessment activities addressed the repetitive aspects. The inspectors reviewed events to determined whether there existed performance indicator occurrences that involved dose rates greater than 25 Rem/hour at 30 em, dose rates greater than 500 Rem/hour at 1 meter, unintended exposures greater than 100 mrem total effective dose equivalent (TED E), greater than 5 Rem shallow dose equivalent (SDE), or greater than 1.5 Rem lens dose equivalent (LDE). Job-in-Progress Reviews The inspectors observed aspects of various on-going activities to confirm that radiological controls, such as required surveys, area postings, job coverage, and job site preparations were conducted.
 
The inspectors verified that personnel dosimetry was properly worn and that workers were knowledgeable of work area conditions.
 
The inspectors attended briefing meetings for U2 Badger Testing and ISFSI related activities.
 
High Risk Significant.
 
High Dose Rate High Radiation Areas and Very HRA (VHR82 Controls . Key control associated with LHRA and VHRAs were reviewed by inspectors to assess Entergy's controls and inventory and to verify accessible LHRAs were properly seicured and posted during plant tours. The inspectors discussed with radiation protection supervision the adequacy of high dose rate HRA and \lHRA controls and procedures and verified that no programmatic or procedural changes have occurred that reduce the effectiveness and level of worker protection, Radiation Worker Performance During observation of the work activities listed above, the inspectors evaluated radiation worker performance with respect to the specific radiation protection work requirements and their knowledge of the radiological conditions in their work areas. The inspectors reviewed eRs related to radiation worker performance to determine if an observable pattern traceable to a similar cause was evident Radiation Protection Technician Proficienc),:
During observation of the work activities listed above, inspectors evaluated radiation protection technician work performance with respect to' their knowledge of the radiological conditions, the specific radiation protection work requirements and radiation protection procedures.
 
The inspectors reviewed eRs related to radiation protection technician performance to determine if an observable pattern traceable to a similar cause was evident. Enclosure


====b. Findings====
====b. Findings====
No findings of significance were identified. 1
No findings of significance were identified 20S2 AlARA Planning and Controls (71121.02
{{a|EP6}}
-10 samples)
==EP6 Drill Evaluation==
{{IP sample|IP=IP 71114.06|count=1}}


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors evaluated an emergency classification conducted on October 5, 2009, during a licensed-operator requalification examination conducted in the plant-reference simUlator. The inspectors observed an operating crew respond to simulated initiating events and malfunctions that ultimately resulted in the simulated implementation of the site emergency plan. In particular, the inspectors verified the adequacy and accuracy of the simulated emergency classification of 'Site Area Emergency.' The inspectors verified this initial classification was appropriately credited as an opportunity toward NRC performance indicator data. The inspectors observed the management evaluation and training critique following termination of the scenarios, and verified that performance deficiencies were appropriately identified and addressed within the critique and, as applicable, within the corrective action program. Also, the inspectors reviewed the summary performance report for the evaluation and verified that appropriate attributes of drill performance including deficiencies were captured. This evaluation constituted one inspection sample.
During September 28 through October 2, 2009, the inspectors conducted the foUowing activities to verify that Entergy staff were properly maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the AlARA program was reviewed for conformance with the criteria contained in 10 CFR 20, applicable industry standards, and Entergy's procedures.
 
The documents reviewed during this inspection are listed in the Attachment.
 
Inspection Planning The inspectors reviewed pertinent infonmation regarding cumulative exposure history, current exposure trends, and on-going activities to assess current perfonmance and outage exposure challenges.
 
The inspectors determined the site's 3-year rolling collective average exposure.
 
The inspectors reviewed work performed during the inspection period, the associated ALARA plans, RWPs, AlARA Committee Reviews, exposure estimates, actual exposures and post job reviews. Jobs reviewed included those listed earlier in this report in Section 20S1. The inspectors reviewed implementing procedures associated with maintaining occupational exposures AL.ARA. This included a review of the processes used to estimate and track work activity exposures.
 
Radiological Work Planning With respect to the work activities reviewed, the inspectors reviewed dose summary reports, related post-job ALARA reviews, related RWPS, exposure estimates and actual exposures, and ALARA Committee meeting paperwork.
 
The inspectors reviewed ALARA work activity evaluations, exposure estimates, and exposure mitigating requirements were reviewed for work packages.
 
The inspectors' review was to verify whether the licensee has established procedures and work controls, based on sound radiation protection principles.
 
The inspectors compared the results aChieved with the intended dose that was established in the planning of the work. The inspectors evaluated the basis for inconsistencies between the intended and actual work activity doses and station management awareness and involvement.
 
Job Site Inspections and ALARA Controls The inspectors reviewed work activities that present the highest radiological risk to workers. The inspectors evaluated the licensee's use of engineering controls to achieve dose reductions and to verify that procedures and controls are consistent with ALARA reviews. Associated ALARA Plans and RWPS were reviewed by inspectors to determine if appropriate exposure and contamination controls were being employed.
 
Radiation Worker Performance Through observations and interviews, the inspectors reviewed whether workers and technicians were found to be knowledgeable of the work area radiological conditions and low dose waiting areas. Declared Pregnant Workers The inspectors reviewed information associated with declared pregnant workers (juring the assessment period and whether appropriate monitoring and controls were being utilized to ensure compliance with 10 CFR 20. Problem Identification and Resolution The inspectors reviewed elements of the licensee's corrective action program related to implementing radiological controls to determine if problems are being entered into the program for timely resolution.
 
No findings of significance were
 
==OTHER ACTIVITIES==
40A 1 Performance Indicator Verification (71151 -8 samples) Inspection Scol2e The inspectors reviewed performance indicator (PI) data for the cornerstones listed below and used Nuclear Energy Institute
'Regulatory Assessment Performance Indicator Guideline," Revision 6, to verify individual PI accuracy and completeness, The inspectors reviewed the PI data and supporting documentation from the fourth quarter of 2008 through the third quarter of 2009 to verify the accuracy of the reported data, The documents reviewed during this inspection are listed in the Attachment.


====b. Findings====
Barrier Integritv Cornerstone
No findings of significance were identified.
* Reactor Coolant Identified Mitigating Systems Mitigating System Performance Index Heat Removal Systems; and
* Mitigating System Performance Index Cooling Water Occupational Radiation Safety Occupational Exposure Control Effectiveness.
 
Encfosure
 
===.1 Public Radiation===


==RADIATION SAFETY==
Safety Cornerstone Radiological Effluent Technical Specifications (RETS)fOffsite Dose Calculation Manual (ODCM) Radiological Effluent Occurrences.


===Cornerstone:===
Emergency Preparedness Cornerstone Drill and Exercise Performance (DEP); ERO Drill Participation; and ANS Reliability.
Occupational Radiation Safety (OS) 20S1 Access Control to Radiologically Significant Areas (71121.01 -15 samples) a' Inspection Scope During September 28 through October 2, 2009, the inspectors conducted activities to verify that Entergy staff at IPEC were properly implementing physical, engineering, and Enclosure 2 24 administrative controls for access to high radiation areas (HRAs), and other radiologically controlled areas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the criteria contained in 10 CFR 20, site technical specifications, and licensee's procedures required by the Technical Specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The documents reviewed during this inspection are listed in the Attachment. The inspectors performed independent radiation dose rate measurements and reviewed the following items: Plant Walk Downs and RWP Reviews The inspectors reviewed exposure significant work areas within radiation areas, HRAs, and airborne areas in the plant to assess licensee controls and surveys for adequacy. Work reviewed included 3R15 Refueling Outage and On-Line work activities: U2 RCP Platform Entry (Oil Addition) U2 Vapor Containment, Replace 21 CRD Fan Motor radiation work permit (RWP) 2009-2033 U2 Testing and Fuel Moves, RWP 2009-2043 U2 Dry Cask Storage &Associated Work, RWP 2009-2029 Radiation protection support for locked HRA (LHRA) Entries, RWP 3501 Maintenance Support, RWP 2009-3506 Waste Management, RWP 2009-3504 Scaffolding, RWP 2009-3518 Outage Valve Work, RWP 2009-3520 Reactor Disassembly & Reassembly, RWP 2009-3521 Split Pin Repair & Associated Work, RWP 2009-3530 RCP Pump & Motor Work, RWP 2009-3534 With a survey instrument and assistance from a Health Physics qualified individual, the inspectors walked down various areas to determine: whether the RWP, procedure, and engineering controls were in place and whether surveys and postings were adequate. The inspectors reviewed RWPs that provide access to exposure-significant areas of the plant. Specified electronic personal dosimeter alarm set points were reviewed by inspectors with respect to current radiological condition applicability and workers were queried to verify their understanding of plant procedures governing alarm response and knowledge of radiological conditions in their work area. The inspectors determined there were no RWPs for airborne radioactivity areas with the potential for individual worker internal exposures of >50 mil/kern (mrem) committed effective dose equivalent (CEDE). Additionally, the inspectors determined there were no internal dose assessments that resulted in actual internal exposures greater than 50 rnrem CEDE. ' Enclosure 2 25 Problem Identification and Resolution The inspectors reviewed access control-related CRs generated since the last inspection in this area was conducted. Staff members were interviewed and documents reviewed to determine that follow-up activities are being conducted in an effective and timely manner, commensurate with their safety and risk. For repetitive deficiencies or significant individual deficiencies in problem identification and resolution, the inspectors determined if the licensee's assessment activities addressed the repetitive aspects. The inspectors reviewed events to determined whether there existed performance indicator occurrences that involved dose rates greater than 25 Rem/hour at 30 cm, dose rates greater than 500 Rem/hour at 1 meter, unintended exposures greater than 100 mrem total effective dose equivalent (TEDE), greater than 5 Rem shallow dose equivalent (SDE), or greater than 1.5 Rem lens dose equivalent (LDE). Job-In-Progress Reviews The inspectors observed aspects of various on-going activities to confirm that radiological controls, such as required surveys, area postings, job coverage, and job site preparations were conducted. The inspectors verified that personnel dosimetry was properly worn and that workers were knowledgeable of work area conditions. The inspectors attended briefing meetings for U2 Badger Testing and ISFSI related activities. High Risk Significant. High Dose Rate High Radiation Areas and Very HRA (VHRA) Controls Key control associated with LHRA and VHRAs were reviewed by inspectors to assess Entergy's controls and inventory and to verify accessible LHRAs were properly secured and posted during plant tours. The inspectors discussed with radiation protection supervision the adequacy of high dose rate HRA and VHRA controls and procedures and verified that no programmatic or procedural changes have occurred that reduce the effectiveness and level of worker protection. Radiation Worker Performance During observation of the work activities listed above, the inspectors evaluated radiation worker performance with respect to the specific radiation protection work requirements and their knowledge of the radiological conditions in their work areas. The inspectors reviewed eRs related to radiation worker performance to determine if an observable pattern traceable to a similar cause was evident. Radiation Protection Technician Proficienc-x: During observation of the work activities listed above, inspectors evaluated radiation protection technician work performance with respect to their knowledge of the radiological conditions, the specific radiation protection work reqUirements and radiation protection procedures. The inspectors reviewed eRs related to radiation protection technician performance to determine if an observable pattern traceable to a similar cause was evident.


====b. Findings====
====b. Findings====
No findings of significance were identified Enclosure 2 20S2 ALARA Planning and Controls (71121.02 -10 samples)
No findings of significance were identified. Identification and Resolution of Problems (71152 -2 sam pies) Resident Inspector Daily Review of Conditions Reports


====a. Inspection Scope====
====a. Inspection Scope====
During September 28 through October 2, 2009, the inspectors conducted the following activities to verify that Entergy staff were properly maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the ALARA program was reviewed for conformance with the criteria contained in 10 CFR 20, applicable industry standards, and Entergy procedures. The documents reviewed during this inspection are listed in the Attachment. Inspection Planning The inspectors reviewed pertinent information regarding cumulative exposure history, current exposure trends, and on-going activities to assess current performance and outage exposure challenges. The inspectors determined the site's 3-year rolling collective average exposure. The inspectors reviewed work performed during the inspection period, the associated ALARA plans, RWPs, ALARA Committee Reviews, exposure estimates, actual exposures and post job reviews. Jobs reviewed included those listed earlier in this report in Section 20S1. The inspectors reviewed implementing procedures associated with maintaining occupational exposures ALARA. This included a review of the processes used to estimate and track work activity exposures. Radiological Work Planning With respect to the work activities reviewed, the inspectors reviewed dose summary reports, related post-job ALARA reviews, related RWPS, exposure estimates and actual exposures, and ALARA Committee meeting paperwork. The inspectors reviewed ALARA work activity evaluations, exposure estimates, and exposure mitigating requirements were reviewed for work packages. The inspectors' review was to verify whether the licensee has established procedures and work controls, based on sound radiation protection prinCiples. The inspectors compared the results achieved with the intended dose that was established in the planning of the work. The inspectors evaluated the basis for inconsistencies between the intended and actual work activity doses and station management awareness and involvement. Job Site Inspections and ALARA Controls The inspectors reviewed work activities that present the highest radiological risk to workers. The inspectors evaluated the licensee's use of engineering controls to achieve dose reductions and to verify that procedures and controls are consistent with ALARA reviews. Associated ALARA Plans and RWPS were reviewed by inspectors to determine jf appropriate exposure and contamination controls were being employed. Radiation Worker Performance Through observations and interviews. the inspectors reviewed whether workers and technicians were found to be knowledgeable of the work area radiological conditions and low dose waiting areas. . Enclosure 2 27 Declared Pregnant Workers The inspectors reviewed information associated with declared pregnant workers during the assessment period and whether appropriate monitoring and controls were being utilized to ensure compliance with 10 CFR 20. Problem Identification and Resolution The inspectors reviewed elements of the licensee's corrective action program related to implementing radiological controls to determine if problems are being entered into the program for timely resolution. No findings of significance were
As required by Inspection Procedure 71152, "Identification and Resolution of Problems," and to identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for CRs and attending CR group screening meetings.


==OTHER ACTIVITIES==
In accordance with the baseline inspection modules, the inspectors selected corrective action program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for further follow-up and review. The inspectors assessed Entergy personnel's threshold for problem identification, adequacy of the causal analysis, extent of condition reviews, and operability determinations.
Performance Indicator Verification (71151 -6 samples) Inspection Scope The inspectors reviewed performance indicator (PI) data for the cornerstones listed below and used Nuclear Energy Institute 99-02, 'Regulatory Assessment Performance Indicator Guideline," to verify individual PI accuracy and completeness. The inspectors reviewed the PI data and its supporting documentation from the fourth quarter of 2008 through the third quarter of 2009 to verify the accuracy of the reported data. The documents reviewed during this inspection are listed in the Attachment. Barrier Integrity Cornerstone
* Reactor Coolant System Occupational Radiation Safety
* Occupational Exposure Control Public Radiation Safety Radiological Effluent Technical SpeCifications (RETS)/Offsite Dose Calculation Manual (ODCM) Radiological Effluent Occurrences Emergency Preparedness Cornerstone Drill and Exercise Performance (DEP): ERO Drill Participation; and ANS Reliability. Enclosure 2


===.1 28
and timeliness of the associated corrective actions.


====b. Findings====
====b. Findings====
No findings of significance were identified. 40A2 Identification and Resolution of Problems===
No findings of significance were identified . . Semi-Annual Trend Review (71152 -1 sample)
{{IP sample|IP=IP 71152|count=3}}
Routine Problem Identification and Resolution (PI&R) Program Review


====a. Inspection Scope====
====a. Inspection Scope====
As required by Inspection Procedure 71152, "Identification and Resolution of Problems," and to identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for eRs and attending condition report screening meetings. In accordance with the baseline inspection modules, the inspectors selected corrective action program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for further follow-up and review. The inspectors assessed Entergy personnel's threshold for problem identification, the adequacy of the cause analysis, extent of condition reviews, operability determinations, and the timeliness of the associated corrective actions.
The inspectors performed a semi-annual review of Unit 2 issues, to identify trends that might indicate the existence of more significant safety issues. The inspectors included in this review, repetitive or closely-related issues that may have been documented by Entergy outside of the corrective action program, such as trend reports, performance indicators, major equipment problem lists. maintenance rule assessments, and maintenance or corrective action program backlogs.
 
The inspectors also reviewed Entergy's corrective action program database for the tl1ird and fourth quarters of 2009, to assess CRs written in various subject areas (eqUipment problems, human performance
 
===.3 issues, etc.), as well as individual===


====b. Findings====
issues identified during the NRC's daily CR review. The inspectors reviewed Entergy's quarterly trend report for the third quarter of 2009, and specific inputs from the Engineering Department that were included in the site trend report, to verify the existence or absence of, identified trends and the adequacy of existing corrective actions to address these trends. The inspectors also reviewed 121, "Entergy Trending Process," to verify that Entergy was appropriately evaluating and trending adverse conditions in accordance with applicable procedures.
No findings of significance were identified . . 2 Semi-Annual Trend Review (71152 -1 sample)
 
The documents reviewed during this inspection are listed in the Attachment.
 
b. Assessment and Observations No findings of Significance were identified.
 
The inspectors identified several issues and events that occurred over the course of the past year, and more specifically, the third and fourth quarters of 2009, which could objectively be considered adverse trends. The inspectors verified that these issues were either addressed within the scope of the corrective action program, or through department review and documentation in the quarterly trend report for overall assessment.
 
For example, the inspectors reviewed the following issues: IP2-2009-04306
-Root Cause Evaluation:
Adverse Trend* Centrifugal Pump Rework; and IP2-2009*02629
* Recent events involving weaknesses in supplemental personnel work practices; No adverse trends were identified by the inspectors that were not previously addressed by Entergy personnel.
 
Aggregate Impact of Operator Workarounds (71152 -1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a semi-annual review of site issues, to identify trends that might indicate the existence of more significant safety issues. The inspectors included in this review, repetitive or closely-related issues that may have been documented by Entergy outside of the corrective action program, such as trend reports, performance indicators, major equipment problem lists, maintenance rule assessments, and maintenance or corrective action program backlogs. The inspectors also reviewed Entergy's corrective action program database for the third and fourth quarters of 2009, to assess CRs written in various subject areas (equipment problems, human performance issues, etc.), as well as individual issues identified during the NRC's daily CR review. The inspectors reviewed Entergy's quarterly trend report for the third quarter of 2009. and specific inputs from the Engineering Department that were included in the site trend report, to verify the existence or absence of. identified trends and the adequacy of existing corrective actions to address these trends. The inspectors also reviewed 121. "Entergy Trending Process," to verify that Entergy personnel were appropriately evaluating and trending adverse conditions in accordance with applicable procedures.
The inspectors conducted a review of the aggregate impact of operator burdens and workarounds.
 
The inspectors reviewed Entergy's implementation of procedures 45, "Operator Burden Program." The inspectors conducted control room walkdowns and interviewed plant operators to determine the impact of defiCiencies on operator response to plant events. Additionally.
 
the inspectors reviewed operator logs, CRs and performed system walkdowns to verify that there were no risk significant operator actions that had not been evaluated by Entergy personnel.


====b. Findings and Observations====
====b. Findings and Observations====
No findings of significance were identified. Enclosure 2 The inspectors reviewed issues that occurred over the course of the past year, and more specifically, the third and fourth quarters of 2009, which could objectively be considered potential adverse trends. The inspectors verified these issues were either addressed within the scope of Entergy's corrective action program. or through department review and documentation in the quarterly trend report for overall assessment. For example, the inspectors reviewed the following issues: IP2-2009-04306 -Root Cause Evaluation: Adverse Trend -Centrifugal Pump Rework; IP3-2009-02554 -Isolated phase bus duct cooling issues; IP2-2009-02629 -Recent events involving weaknesses in supplemental personnel work practices; IP3-2009-03928 102983/04847 -Fuse failures on control building exhaust fans; and IP3-2009-03626/02539 -Unit 3 Plant Scrams. The inspectors did not identify an adverse trend that was not previously identified by Entergy personnel. However, the inspectors identified that Entergy personnel appeared inconsistent in their implementation of the Corrective Action Program as it related to ensuring an adverse trend evaluation was performed by station personnel conSistently with respect to recent CRs regarding control building fans. (1) On July 4, 2009, the 32 control building fan did not start due to fuse failures on the B &C phases in the electrical starting circuitry. The Condition Review Group (CRG) requested an assessment to determine if an adverse trend existed; (2) On September 27,2009, the 31 control building fan did not start due to fuse failures on all three electrical phases. The inspectors identified that Entergy personnel closed this issue to "track and trend" after application of trend codes, but did not document a search of the corrective action database to evaluate for adverse trending; .and (3) On December 21, 2009, the 31 control building fan experienced a blown fuse on a single electrical phase. Entergy replaced the fuse and no further action appeared to have been taken within the corrective action program. The inspectors' assessment did not determine that an adverse trend existed or Entergy's actions were in violation of procedure requirements. However, the inspector's determined that an implementation threshold for conducting adverse trend evaluations with respect to this small sample were not consistently considered or implemented by station personnel. Annual Sample: 31 Pressurizer Backup Heater Molded Case Circuit Breaker Issues
No findings of significance were identified.
 
The inspectors verified that operator workarounds and burdens were entered into the corrective action program at an appropriate threshold and that corrective actions were planned or taken commensurate with their safety significance.
 
===.1 40A3 Event Follow-Up===
 
(71153 -2 samples) Reactor Trip on November 2, 2009. Due to a Turbine-Generator Exciter Protective Trip
 
====a. Inspection Scope====
The inspectors responded to the control room on November 2,2009, following an automatic reactor trip due to a turbine-generator protective trip resulting from a loss of the Generrex power supply. The inspectors observed Entergy's post-trip response in the control room to determine if plant equipment responded as expected, and to ensure that operating procedures were being appropriately implemented.
 
The inspectors attended post-trip review and forced outage meetings, and discussed the event, plant response and corrective actions with plant management.
 
The purpose of the reviews was to confirm that Entergy had taken appropriate actions during and foHowing the event, and had taken appropriate corrective actions for the trip prior to commencing restart activities.
 
The documents reviewed during this inspection are listed in the Attachment.
 
====b. Findings====
No findings of significance were identified . . 2 Partial Loss of Control Room Indication During NI-41 Recorder Replacement


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed Entergy's corrective actions to address an indication of oil external to circuits 7 and 8 on the 31 pressurizer backup heater molded case circuit breaker (MCCB) panel as well as a subsequent failure of the circuit 8 MCCB. The inspectors reviewed corrective actions associated with CR-IP3-2004-02896, 2008-01108, CR-IP3-2008-01235, and CR-IP3-2008-01287, and assessed each Enclosure 2
The inspectors responded to the control room on November 23,2009, following notification by the shift manager that there had been a partial loss of control room annunciators and alarms associated with safety relief valve acoustic monitor indication, low range steam and feedwater flow indication, and inadvertent control rod movement Entergy personnel determined that the partial loss of c(mtrol room indications and control rod movement was due to inadvertent grounding of a live feed wire during the replacement of a control room digital recorder.
 
The grounding caused the recorder's associated breaker to open and the 21 instrument bus to from its normal source (static inverter)to its alternate source (transformer).
 
The inspectors verified that Entergy operations and maintenance personnel had taken appropriate actions following the inadvertent grounding of the wire and resultant control room indications.
 
The inspectors' review included verification that applicable TS limiting conditions of operation (LCO) were entered by operations personnel for the eqUipment made inoperable by the partial loss of control room indications/alarms.
 
Finally, the inspectors performed system walkdowns, interviewed personnel, reviewed applicable CR's, work packages, plant procedures, operating experience and corrective actions associated with the apparent cause evaluation performed by Entergy personnel to independently assess the causes of the partial loss of control room annunciators.


===.3 30 condition report to ensure apparent cause evaluations appropriately identified causes and corrective actions were adequate and appropriate for the circumstances and risk significance.===
The documents reviewed during this inspection are listed In the Attachment.


====b. Findings and Observations====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
An NRC-identified NCVof very low safety significance (Green) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," was identified because Entergy personnel did not promptly identify and correct a condition adverse to quality regarding molded-case circuit breaker (MCCB) non-conformance.
A self-revealing NCV of very low safety significance (Green) of 10 CFR 50, Appendix B Criterion V "Instructions, Procedures, and Drawings," was identified because Entergy personnel did not perform work in accordance with instructions associated with the replacement of a control room digital recorder.
 
As a result, during performance of Enclosure the work, Entergy personnel shorted a live wire which resulted in a partial loss of control room indications and alarms, and inadvertent control rod movement.


=====Description:=====
=====Description:=====
On October 27,2009, the inspectors identified that Entergy personnel did not maintain or provide a basis for extending the qualification basis for original equipment Westinghouse MCCBs. Entergy personnel previously identified 2704) that a portion of the population of MCCBs in safety related applications exceeded the defined design life as specified in Westinghouse Technical Bulletin (TB), TB-04-13, "Replacement Solutions for Obsolete Classic Molded Case Circuit Breakers, UL Testing Issues, Breaker Design Life and Trip Band Adjustment" dated June 28, 2004. The inspectors noted that the MCCBs had been In service for greater than 29 years compared to the 20 year design life specified in the TB-04-13. The TB-04-13 documented the defined service life for these MCCBs as being 20 years and provided operating experience that would indicate MCCB replacement at 20 years was necessary and more frequent preventive maintenance activities to ensure the MCCBs maintained reliability. The TB-04-13 further described aging of the lubricants applied internaHy to Westinghouse MCCBs during manufacturing as a key limiting factor for continued MCCB operability. Previously, when Entergy personnel evaluated TB-04-13 in condition report 2704, station personnel initiated corrective actions to replace all safety related and important to safety MCCBs at refueling outage intervals with final expected completion date by 2011. The MCCB replacements were administratively tracked by preventive maintenance change request (PMCR)-04-480V-IMD-121. The inspectors noted that Unit 2 previously began replacement of its MCCBs installed in safety related applications in the early 1990s. Notwithstanding Entergy's actions to complete MCCB replacement by 2011, the inspectors determined that Entergy personnel did not take timely corrective actions to replace or evaluate qualification of breakers In service which already exceeded their year design life. The inspectors noted that MCCBs not yet replaced included breakers associated with the 31 instrument bus, 31 DC distribution panel, 31 DC power panel. and several panels for EDG auxiliary components. Additionally, the inspectors identified additional opportunities in 2008 for Entergy personnel to reasonably revisit the adequacy of unit specific actions to address the MCCB aging and lubrication issues regarding Westinghouse MCCBs. For example. on two separate occasions in May 2008, circuit 8 on the 31 pressurizer backup heater MCCB panel was discovered tripped free during operator rounds (CR-IP3-2008-01108; CR-IP3-2008-01235). Entergy personnel did not determine the cause of the tripped free condition on either occasion or initiate actions to review work history and PM frequency for the subject breaker. On June 24,2008, during MCCB circuit 8 replacement activities, Entergy staff identified a thick oily, dielectric residue covering the MCCB and initiated Enclosure 2 i
On November 23, 2009, during the replacement of control room safety related digital recorder NR-41, electrical maintenance personnel inadvertently grounded the recorder's live power lead on the bracket of the recorder.
* CR-IP3-2008-01287. Engineering provided verbal input that the MCCB contained a conductive grease susceptible to break down leaving a residue. Entergy personnel did not fully evaluate the failed MCCB. Entergy's operability review only considered the tripping function of the MCCBs and the potential impact on continuity of electric power to safety related loads was not reviewed. The extent of condition review was limited to the pressurizer heater backup panels and a potential issue with similarly aged MCCBs in safety related applications was not reviewed. The inspectors concluded these were reasonable opportunities indicative of current performance for which Entergy did not thoroughly evaluate adverse conditions regarding MCCBs. Based upon inspector questions, Entergy personnel initiated CR-IP3-2009-04262 to address the concerns with the MCCB aging issues. Entergy personnel completed a reasonable expectation of operability on October 30,2009, and an operability determination on November 5, 2009. Entergy personnel concluded that the MCCBs beyond 20 years in service were operable but nonconforming. Entergy's evaluations for continued operability were supplemented by compensatory actions to visually inspect the MCCBs once per shift and thermography once per week. The inspectors determined the operability evaluations supplemented by additional interim inspections and remaining corrective actions for MCCB replacement by the 2011 spring refuel outage were appropriate.
 
NR-41 provides control room operators indication for reactor power from the power range upper detectors.
 
Entergy personnel determined this resulted in the NR-41 circuit breaker opening and the power supply for the 21 instrumentation bus auto transferring from normal (static inverter)power to the alternate (transformer)power supply requiring entry into TS limiting condition of operation (LCO) 3.8.7. The opening of the circuit breaker caused a partial loss of control room annunciators retated to acoustic monitors for safety relief valves, and low range steam and feed flow indication.
 
In addition, operations personnel observed control rods (control bank '0') move in half a step. Entergy personnel determined that the control rod movement occurred because of the power transient associated with the 21 instrument bus transferring from its static inverter to an alternate power supply. The inspectors determined that Work Order (WO) 163807 provided instructions for replacing NR-41 and required the performance of a pre-job brief. WO 163807 idEmtified that working on live circuits was a "safety hazard" and an "error likely situation." The WO instructed maintenance personnel to "tape all areas where feed wires present. if applicable." The inspectors determined that EN-HU-102, "Human Performance Tools," requires an acceptable defense against an error likely situation and taping of all areas was identified in the WO as the human performance tool to address the error likely situation.
 
Entergy determined that the maintenance personnel did not apply the electrical barriers to prevent the inadvertent ground of the live power supply prior to performing the work. Following the event. inspectors observed that Entergy personnel replaced the digital recorder with the circuit breaker opened to eliminate the electrical hazard. Entergy entered the issue into the corrective action program (CR-IP2-2009-04860)and implemented corrective actions which included supplemental training for station personnel regarding the station's requirements to follow procedural direction.


=====Analysis:=====
=====Analysis:=====
The inspectors determined that not promptly correcting non-conforming MCCBs which exceeded their design qualification life was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the reliability of the electrical distribution system to respond to initiating events to prevent undesirable consequences. Specifically, not maintaining qualified components in the electrical distribution system could impact the ability of certain MCCBs to function as necessary during design basis events and plant transients. The inspectors evaluated the finding in accordance with IMC 0609. Significance Determination Process, Attachment 0609.04, Phase 1 -Initial Screening and Characterization of Findings. The finding screened as very low safety significance (Green) because the finding was a design qualification deficiency confirmed not to result in loss of operability or function. Specifically, the inspectors determined there was no actual loss of function that could be attributed to operating with MCCBs greater than 20 years in service. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution because on several occasions Entergy personnel did not thoroughly evaluate MCCB qualification issues including operability and functionality considerations. This included an opportunity for Entergy personnel to evaluate the condition in 2008 when engineers identified residue indicative of lubrication breakdown. (P.1 (c) per IMC0305)
The inspectors determined that a performance deficiency associated with this finding was that Entergy maintenance personnel did not follow instructions provided in the WO to install electrical protective barriers when working on live equipment.
 
The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences.
 
Specifically, the grounded recorder power supply resulted in a loss of control indication and alarms that would impact operations' response to an event. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A.
 
Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green) because it did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events. Enclosure
 
===.1 The inspectors===
 
determined that this finding had a cross"cutting aspect in the area of human performance related to work practices because Entergy personnel did not follow procedures during the replacement of a control room digital recorder. (H.4(b))


=====Enforcement:=====
=====Enforcement:=====
10 CFR 50, Appendix 6. Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality, such as nonconformances, are promptly identified and corrected. Contrary to the above, on August 2, 2004. Entergy did not promptly identify and correct a population of safety-related MCCBs that were nonconforming to the vendor's 20 year qualified design life as defined in Westinghouse Technical Bulletin, 04-13, Replacement Solutions for Obsolete Classic Molded Case Circuit Breakers, UL Enclosure 2 Testing Issues, Breaker Design Life and Trip Band Adjustment, dated June 28, 2004. Specifically, Entergy personnel did not correct the condition or provide an engineering basis for continued MCCB operability until October 30, 2009. Because this violation was of very low safety significance and was entered into Entergy's corrective action program as CR-IP3-2oo9-04262, this violation is being treated as an NCV, consistent with Section VLA.1 of the NRC Enforcement Policy (NCV 05000286/2009005-04: Failure to promptly identify and correct a MCCB service life nonconformance ) . Operator Workarounds Review
10 CFR 50, Appendix'B Criterion V "Instructions, Procedures, and Drawings" in part, requires that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions.
 
Contrary to the above, on November 24,2009, maintenance personnel did not follow the instructions provided in the WO during replacement of the safety-related digital recorder NR-41. Because this issue was of very low safety significance and was entered into Entergy's correctivH action program (IP2-2009-04860).
 
this violation -is being treated as an NeV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000247/2009005*03, Partial Loss of Control Room Indication during NI-41 Recorder Replacement)40A5 Other Activities Quarterly Resident Inspector Observations of Security Personnel and Activities


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors conducted a review of the aggregate impact of operator workarounds on the ability of operators to implement abnormal and emergency operating procedures, and to ensure that mitigating systems that are impacted remain capable of performing the associated safety functions. This review included operator burdens, as well as control room alarms and deficiencies. The inspectors reviewed the prioritization, assessment, and disposition of the inputs to the aggregate impact that is accomplished through the site's Unit Reliability Team. and the implementation and assessment of the Operations Aggregate Indicator, which is described in EN-OP-115, "Conduct of Operations," and OAP-045, "Operator Burden Program." The inspectors conducted plant and control panel walkdowns, as applicable, reviewed the corrective action program database, and discussed various deficiencies with Entergy personnel, to determine the overall impact the deficiencies would have on operator response to plant events.
During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with site security procedures and regulatory requirements relating to nuclear plant security, These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.


====b. Findings and Observations====
====b. Findings====
No findings of Significance were identified. Event Follow-up (71153 -6 samples) (Closed) LER 05000286/2008-006-01. and 2009-001-01, Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt Safeguards Bus 6A. On October 9, 2008, and again on January 2, 2009, during surveillance testing associated with undervoltage and degraded relays for 480V safety bus No. 6A, the normal supply breaker for the bus No. 6A opened unexpectedly. which caused various perturbations, including various loads that were de-energized/re-energized, including the 32 emergency diesel generator to start and load onto its associated bus No. 6A. As previously.documented in NRC inspection report (IR) 05000286/2009-004, Entergy personnel performed evaluations to both determine the cause of the transients, and to mitigate the potential for recurrence of these vital bus transients during surveillance activities. During this inspection period, the inspectors determined that Entergy personnel considered additional information to determine the causes of both events during a comprehensive revision to the root cause evaluation within the corrective action program under CR-IP3-2009-00011. The inspectors determined that station personnel Enclosure 2 33 had reasonably evaluated and considered a wide-range of potential causes for the transients. In addition, Entergy personnel collected information to suggest that human error was a contributor to the transients. The inspectors also concluded that a likely cause may have been inadvertent contact between a relay terminal connection and a test probe. The inspectors verified that Entergy personnel had initiated appropriate procedure changes and required insulated test probes with minimally-exposed contact area to be utilized. The inspectors determined these changes would reasonably minimize any potential contact between exposed terminal connections and the probes, and as a result, mitigate the potential for future transients. As a result, the inspectors did not identify a finding of significance or violation of NRC requirements. These LERs are closed . . (Closed) LER 05000286/2009-005-00, Technical Specification (TS) Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable 480-Volt Undervoltage/Degraded Grid Relay Caused by Personnel Error Based on a 2009 NRC inspection regarding past surveillance data related to 4BO-Volt bus 3A degraded voltage safety-related time delay relay, 62-1/3A, Entergy personnel determined the 62-1/3A relay had been TS inoperable during the time period October 11 through November 8, 2007. Entergy personnel determined the relay exceeded TS 3.3.5.2 surveillance requirement time delay value of equal to or less than 45 seconds. Entergy personnel determined that, at that time, personnel did not fully identify and evaluate the abnormal relay drift during 2007. Entergy personnel implemented corrective actions at that time to replace the degraded relay in November 2007. Entergy also captured the recent performance aspects in CR-IP3-2009-02664 to address management expectations and clarify guidance with respect to staff evaluation of found surveillance failures. The inspectors reviewed the LER to verify its accuracy based on NRC identification of the issue during the May 2009 NRC Problem Identification and Resolution team inspection. The inspectors reviewed Entergy's evaruation and corrective actions as documented in CR-IP3-2009-02664. The inspectors determined the performance aspects that contributed to this issue were previously evaluated by the NRC and dispositioned as a Green NCV in NRC inspection report IR 50-286/2009-008. There were no additional findings of significance or violations of NRC requirements identified. This LER is closed . w. (Closed) LER 05000286/2009006-00, Automatic Reactor Trip Due to a Generator Trip Caused by Actuation of the Generator Protection System lockout Relay During a Severe Storm with Heavy Lightning The inspectors reviewed the LER submitted by Entergy following the plant trip on August 10, 2009, which occurred during severe thunderstorms. The inspectors reviewed the LER and correctlve action program documents to evaluate whether performance issues contributed to, or complicated any subsequent operator or expected equipment action during the event, and whether appropriate corrective actions were identified, as appropriate. This LER is dosed.
No findings of significance were identified . . 2 Independent Spent Fuel Storage Installation
{60855 and 60855.1)


=====Introduction:=====
====a. Inspection Scope====
A self-revealing finding (FIN) of very low safety significance (Green) was identified because Entergy personnel did not perform adequate post-maintenance Enclosure 2 testing for a 6.9kV breaker that is utilized to transfer electrical power for a safety bus to credited off-site power sources following a plant trip.
On December 14, 2009, Entergy personnel completed its dry cask loading campaign for Unit 2. The inspectors reviewed documents and records associated with the operation of the Indian Point Energy Center (IPEG) Independent Spent Fuel Storage Installation (ISFSI), including training records for personnel involved with loading dry cask canisters.


=====Description:=====
The inspectors met with reactor engineering personnel to review the fuel selection process and associated documentation.
On August 10, 2009, the unit experienced a plant trip due to the results of severe thunderstorms, most likely from a lightning strike that caused protective relays susceptible to the effects of such strikes to initiate a main unit generator trip. While the unit is operating at power, 6.9kV buses that provide power to large loads, such as reactor coolant pumps, and downstream 480V safety buses via transformers, are powered through the Unit Auxiliary Transformer (UAT) from the main transformers. Following a plant trip, an automatic transfer normally occurs that transfers this input power to the 6.9kV buses to the station auxiliary transformer (SAT), a dedicated off-site power source required by technical specifications. Following the plant trip on August 10th, the 6.9kV bus 2 failed to have its normal power source automatically transfer from the UAT to the SAT via 6.9kV bus No.5, due to the failure of bus tie breaker 52/UT2ST5 to close on demand, as required. This resulted in the de-energization of 480V safety bus 5A, and the automatic start of emergency diesel generator No. 31 to repower 480V safety bus 5A and its associated loads. The inspectors reviewed the licensee event report, and the apparent cause evaluation performed under condition report (CR) IP3-2009-03380. The inspectors noted that the breaker was installed during a two-year preventive maintenance (PM) activity conducted on or about July 7,2009, and that applicable maintenance procedures and work order instructions did not ensure a functional test was performed (Le., breaker cycled) that verified the breaker would be able to perform its required functions, when required. WM-107, "Post-Maintenance Testing (PMT)," is utilized as guidance by Entergy personnel to ensure criterion for selection or scope, implementation requirements, and documentation of post-maintenance testing is performed effectively. In particular, the inspectors noted the scope of the program, especially in the planning stages of work package generation for maintenance activities, requires that post-maintenance testing activities are selected to ensure that equipment are capable of performing their intended functions. In this case, Entergy's review determined that planned maintenance, and subsequent movement and installation of the spare breaker into the breaker cubicle, resulted at some point, in damage to a trip cam lever that operates a relay required for proper automatic operation of the breaker during a plant trip. This damaged lever was visually identified upon removal of the 6.9kV breaker from the cubicle following the event, and Entergy personnel subsequently determined that the damaged lever prevented the latch trip relay from functioning to cause breaker closure. The inspectors determined that Entergy personnel did not conduct adequate post-maintenance functional testing that ensured proper breaker functioning per PMT standards'. The inspectors verified that Entergy personnel subsequently re-installed an operable breaker, revised maintenance documents to ensure adequate PMT requirements were in place, and changed the performance of this 2-year PM to an outage-related activity to ensure functional testing would be performed, as appropriate.


=====Analysis:=====
The video recording of the fuel bundles placed into the last loaded canister was reviewed by inspectors to verify proper bundle location.
Inadequate post-maintenance functional testing to ensure risk significant breakers are able to perform intended functions was considered a performance deficiency. The inspectors determined that the performance deficiency was more than minor. Specifically, the failure to perform appropriate post-maintenance functional Enclosure 2


===.4 35 testing is associated with the procedure quality performance attribute of the Mitigating Systems comerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was considered to be of very low safety significance (Green) in accordance with lMC 0609.04, "Phase 1 -Initial Screening and Characterization of Findings," because the finding was not a design or qualification deficiency, did not result in the loss of any safety function, and was not risk significant due to external events. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because Entergy personnel did not ensure that adequate maintenance functional testing appropriate for the circumstances was performed on a 6.9kV bus tie breaker. [H.3(a) per IMC0305].
The inspectors review also included verification of the annual inventory and the location of each dry cask canisters on the ISFSI pad. The inspectors interviewed radiation protection personnel to review total dose per canister, AlARA goals, and neutron dose determinations.


=====Enforcement:=====
The inspectors also interviewed fire protection personnel to determine the follow up to assess the adequacy of the evaluation Entergy performed for all transient combustibles to be stored on the ISFSI pad. The documents reviewed during this inspection are listed in the Attachment.
The inspectors determined that Entergy did not perform appropriate maintenance functional testing of 6.9kV breaker 52fUT2ST5PMT after replacement and performance of a 2-year preventive maintenance activity on July 7, 2009. Because this finding does not involve a violation of regulatory requirements and has very low safety significance, this issue is being treated as a FIN (FIN 05000286/2009005-05, Inadequate post-maintenance testing and resultant failure of 6.9kV breaker transfer following plant trip) . (Closed) LER 05000286/2009-006-00, Automatic Reactor Trip Due to a Turbine Trip as a Result of a Turbine Autostop Oil Actuation Caused by a Failed Autostop Oil Fitting. On August 10, 2009, an unplanned turbine and reactor trip occurred, which Entergy personnel later determined to be caused by a non-safety related failed fitting associated with the turbine autostop oil system. The location of the failed fitting resulted in a lowering 011 pressure below a turbine trip setpoint. The inspectors reviewed the root cause evaluation performed within the corrective action program under condition report CR-IP3-2009-03592. The inspectors noted that Entergy personnel determined that the fitting configuration did not fully consider adequate thread engagement aspects that could result in a long term high cycle fatigue concern. Additionally, Entergy personnel determined that station personnel during a 1995 system modification maintained the original fitting configuration upon system restoration from that modification. Ultimately, Entergy personnel determined this fitting failed due to high cycle fatigue. Entergy maintenance personnel replaced the fitting, performed inspections of tubing and other fittings for extent of condition checks, and initiated other actions for long-term tracking of this configuration to ensure reliability going forward on both Units 2 and 3. The inspectors concluded that this latent fitting configuration issue was not reasonably within Entergy's ability to foresee and correct. Because this non-safety related fitting is normally in a non-accessible area and there are limited outage-related opportunities to identify and observe this condition, the inspectors did not Identify a performance deficiency. No findings of significance or violations of NRC reqUirements were identified; therefore, this LER is closed . . (Closed) LER 05000286/2009-004-00, Automatic Reactor Trip Due to a High Steam Generator 32 Water Level Caused by Inadequate 31 Main Feedwater Pump Governor Valve Setting and 32 Main Steam Generator Level Controller Set-up. Enclosure 2===


36 The inspectors reviewed the lER submitted by Entergy following the plant trip on May 28, 2009, which occurred during a downpower to address a problem with the 32 main boiler feedwater pump. The inspectors reviewed the lER and corrective action program documents to evaluate whether performance issues contributed to, or complicated any subsequent operator or expected equipment action during the event, and whether appropriate corrective actions were identified, as appropriate. This lER is closed.
====b. Findings====


=====Introduction:=====
=====Introduction:=====
A self-revealing finding (FIN) of very low safety significance (Green) was identified because Entergy personnel did not conduct maintenance in accordance with maintenance procedures and processes on the 31 and 32 main boiler feedwater pumps (MBFP). The inadequate maintenance resulted in an unexpected downpower and subsequent reactor trip.
An NRC-identified NCV of very low safety significance (SLlV) of 10 CFR 72.212(b)(2)(ii), was identified because Entergy personnel did not evaluate a change to the written evaluation required by this paragraph using the requirements of 10 CFR 72.48(c), prior to storing transient combustibles on the ISFSI pad. The Holtec Final Safety Analysis Report (FSAR) and the Entergy 72.212 Evaluation Report, state that transient combustibles will not be stored on the ISFSI pad.


=====Description:=====
=====Description:=====
On May 28, 2009, a plant operator identified abnormal noises and vibration emanating from the 32 MBFP while the unit was operating at full power. Control room operators commenced a rapid downpower to approximately 63% power and removed the pump from service. Operators subsequently identified and responded to high steam generator water level (SGWl) but were not able to stabilize SGWl before high SGWL trip setpoints initiated an automatic reactor trip by design. The inspectors reviewed the root cause evaluations that addressed the 32 MBFP noise and vibration (CR-IP3-2009-02518) and the automatic reactor trip (CR-IP3-2009-0271 0). Entergy personnel determined that improper high pressure governor valve stroke settings and degraded/worn components associated with the 31 MBFP resulted in flow rates that caused inconsistencies between programmed and actual steam demand for 63% power. This resulted in feedwater regulating valves on all four SGs to cycle full open, as expected for that condition. However, coincident with this governor valve setting and degraded condition, Entergy personnel later determined the controller settings for the 32 (SGWl) control system were not consistent with the SGWl Controller Original Equipment Manufacturer Setpoint document for the plant configuration presented; thus, the controller was unable to prevent an feedwater overshoot to the 32 SG, and this resulted in a reactor trip on high SGWL. During the root cause investigation, Entergy personnel identified that the reset time internal to the 32 SGWL controller was established at 90 seconds. This instrument setting resulted in the controller being saturated during the onset of the level transient, which in effect, inhibited the normal operation of the controller. As a result, the combination of these two configuration control issues directly caused the automatic reactor trip on May 28, 2009. Entergy personnel also identified a circumferential crack on the shaft located near the keyway in the location of the 32 MBFP pump-side coupling hub. This condition caused increased vibration and abnormal noises on the 32 MBFP. A subsequent failure analysis was performed, which determined that the crack was caused by high cycle fatigue. Entergy's analysis further stated that the fatigue cracks were initiated by a loose coupling hub, and during pump operation, continued to propagate through the shaft. Additionally, the inspectors identified that main lessons learned from a previous failure with similar crack location, morphology and causes from 1992 had not been incorporated into site procedures. During 3R15, 32 MBFP planned replacement of the rotating assembly (pump impeller and shaft), Entergy maintenance personnel installed the pump-side coupling hub without the vendor-recommended 80% contact with the shaft. Entergy personnel determined Enclosure 2 from maintenance documentation that approximately 50% contact was between the two surfaces, incorrectly determined to be acceptable by the station, coupled with an undersized keyvvay block, resulted in repetitive and high impacts and its ultimate failure. The inspectors noted that Entergy's conduct was not consistent with guidance in 3-PMP-032-BFP, "Inspection, Overhaul, and/or Replacement of the Main Boiler Feed . Subsequently, the inspectors concluded that Entergy staff identified an appropriate root cause for the 32 MBFP shaft failure, as well as the SGWL controller settings that in conjunction with the degraded 31 MBFP high pressure governor valve, contributed to the SG water level deviations that resulted in the automatic reactor trip. The inspectors determined that the corrective actions were appropriate for the circumstances, which included procedure revisions to ensure appropriate MBFP coupling installation tolerances were achieved. and actions to address the degraded MBFP governor valve. However, the inspectors identi'ned that Entergy personnel had not addressed various performance and technical aspects within the original root cause evaluation for the May 28 trip. For example, work activities regarding the degraded governor valve were not performed in accordance with 0-TUR-402-MFW, "Main Boiler Feed Pump Turbine Inspection," and/or documented within the normal work control process. which were not identified by Entergy's root cause team. In addition, the inspectors identified discrepancies in the timeUne for certain activities detailed in the LER. and the associated root cause evaluation that provided input into this regulatory document. The inspectors discussed the performance root cause evaluation with Entergy management, and subsequent corrective actions were implemented under CR-IP3-2009-04393, 04853, and 04640. The inspectors noted that the subsequent evaluations performed under IP3-2009-04393, corrected the timeline of events, the programmatic and performance aspects that contributed to the events described in the subject LER, and contributed to the improved and revised root cause evaluation with adequate corrective actions appropriate for the circumstances. The inspectors noted that the corrective action plan included improvements to the Corrective Action Review Group, as well as continued actions to address the MBFP governor valve and control system problems. Additionally, the inspectors verified that Entergy personnel will evaluate the revised root cause evaluation to ensure appropriate information is submitted to the NRC in a supplement to the original LER reviewed in this report.
On September 28, 2009, inspectors questioned whether a mobile lift was allowed per procedures to be stored on the ISFSI pad adjacent to unloaded STORMs. Entergy personnel issued condition report O. Station personnel removed the mobile lift off the pad but other transient combustibles.
 
such as plywood and pallets, remained on the pad. During follow-up inspection related to the condition report, the inspectors determined that Entergy personnel were operatinfl under an incorrect assumption that there were active and non-active portions of the ISFsr pad, and that it was acceptable to store transient combustibles and fueled vehicles on the ISFSI pad as long as they were kept at a minimum of 30 feet from loaded casks. The inspectors determined that there was no description of active and non-active portions of the ISFSI pad in Entergy procedures relating to dry cask storage. Entergy uses the Holtee dry cask system under the Certificate of Compliance number 1014 issued to Holtec. The inspectors identified that the Holtec FSAR and the Entergy 72.212 Evaluation Report stated that there will be no combustibles stored on the ISFSI pad. The Holtee FSAR also provided design information that included a worst case fire analysis which concluded that a 50 gallon fuel tank fire (from the vertical cask transporter fuel tank) would result in only minor impact on the HI-STORM.
 
The inspectors reviewed Entergy's control of combustibles corporate procedure which identified the Indian Point ISFSI pad as a Level 1 combustible control zone. The procedure defines Level 1 as a fire sensitive area where transient combustible loading is prohibited unless evaluated and approved via this procedure.
 
In accordance with the Entergy corporate procedure, a transient combustible evaluation (TCE) should be performed prior to the combustibles being stored on the ISFSI pad. A TCE was performed by Entergy on October 19,2009, after the inspectors identified and informed Entergy personnel that combustibles were being stored on the ISFSI pad. The TCE determined that the fire hazard from the combustibles stored on the pad presented less of a fire hazard than the scenario analyzed in the Holtec FSAR. The inspectors questioned whether the TCE was appropriate considering the licensing basis documentation in the Holtec FSAR and 72.212 Evaluation Report allowed no combustibles on the pad. The inspectors determined an evaluation in accordance with the requirements of 10 CFR 72.212(b}(2}{ii}
was required to store combustibles on the ISFSI pad. Subsequent to inspector questions, Entergy personnel entered the issue into the corrective action program and all transient combustibles were removed from the pad.


=====Analysis:=====
=====Analysis:=====
The inspectors determined there was a performance deficiency because Entergy staff did sufficiently implement maintenance procedures O-TUR-402-MFW and 3-PMP-032-BFP to ensure appropriate maintenance was performed on the 31 and 32 MBFPs during the refueling outage in 2009. The inspectors concluded the finding is more than minor because the finding was associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, maintenance performance issues resulted in reliability challenges to the non-safety related feedwater pumps and resulted in unplanned plant transients including an automatic reactor trip. The inspectors evaluated the finding using IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings," and determined the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. Consequently, the finding is of very low safety significance. Enclosure 2
The Reactor Oversight Process (ROP) was not used in this case because inspections of ISFSI activities are covered under NRC Manual Chapter 2690 and are not considered applicable to evaluation under a reactor safety cornerstone in the ROP's Significance Determination Process (SOP). It was determined that the failure to evaluate a change to the written evaluation required by 10 CFR 72.212 in accordance with requirements of 10 CFR 72.48(c} is a Enclosure
 
===.3 29 performance===


===.1 38 The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because Entergy personnel did not maintain effective control over the configuration of the plant due to inadequate supervisory and management oversight of maintenance and design control activities. (H.4{c) per IMC0305).
deficiency that was reasonably within Entergy's ability to foresee and prevent. The finding was determined to be a Severity Level IV violation based on Supplement VI, Example D.2 of the NRC Enforcement Policy.


=====Enforcement:=====
=====Enforcement:=====
Enforcement action does not apply because the performance deficiency was related to non-safety related equipment, processes and procedures and did not involve a violation of regulatory requirements. Because this finding does not involve a violation of regulatory requirements and has very low safety significance, this issue is being treated as a FIN. (FIN 05000286/2009005-06: Inadequate maintenance on MBFPs resulted in an unexpected downpower transient and reactor trip.) Other Activities Quarterly Resident Inspector Observations of Security Personnel and Activities===
10 CFR 72.212(b )(2)(H) requires in part that a licensee shall evaluate any changes to the written evaluations required by this paragraph using the requirements of 10 CFR 72.48{ c}. Contrary to the above, prior to September 28, 2009, Entergy personnel did not evaluate changes to the written evaluation required by this paragraph.
 
Specifically, the Entergy's 10 CFR 72.212 evaluation report determined that a fire suppression system is not used at the IPEC ISFSI pad because there are no combustible materials stored on the ISFSI. However, between September 28, 2009 and December 17, 2009, combustibles were stored on the ISFSI pad and the licensee did not evaluate this change using the requirements of 10 CFR 72.48(c).
 
Because this violation was of very low safeW significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as an NCV consistent with the NRC Enforcement Policy. (NCV 05000247/2009*005*04, Transient Combustibles Stored on the ISFSI Pad) Temporary Instruction 2515/175:
Emergency Response Organization, Drill/Exercise Performance Indicator.
 
Program Review The inspectors performed NRC Temporary Instruction (TI) 2515/175, ensured the completeness of the licensee's completed 1 from the TI, and forwarded that data to NRC Headquarters.


====a. Inspection Scope====
40A6 Meetings.
During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with site security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integraf part of the inspectors' normal plant status review and inspection activities.


====b. Findings====
including Exit On January 21, 2010, the inspectors presented the inspection results to Mr. Joseph Pollock and other Entergy managers and staff, who acknowledged the inspection results. Entergy staff identified documents which were to be considered proprietary and handled as such. ATTACHMENT:  
No findings of significance were identified . . Temporary Instruction 25151175: Emergency Response Organization, Drill/Exercise Performance Indicator, Program Review The inspector performed NRC Temporary Instruction (TI) 2515/175, which ensured the completeness of the TI's Attachment 1, and have forwarded the data to NRC Headquarters. Meetings. including Exit On January 21, 2010, the inspectors presented the integrated inspection results to Mr. Tony Vitale, and other Entergy managers and staff, who acknowledged the inspection results. Entergy staff did not identify documents which were to be considered proprietary. Enclosure 2 40A7 Licensee-Identified Violations The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs. 10 CFR 20.1501 requires that surveys be made to comply with the regulations in 10 CFR Part 20, including 10 CFR 20.1902(b) for posting of high radiation areas (defined as an area greater than 100 mrem/hr at 30 centimeters). Contrary to this, on March 10, 2009, the licensee did not survey changes in radiological conditions during an authorized waste gas transfer from the Volume Control Tank {VCT} to the 36 Small Gas Decay Tank (SGDT). This resulted in the failure to post the SGDT as a high radiation area. Surveys performed apprOXimately 2 to 3 hours after the transfer indicated up to 170 mrem/hr @ 30 cm. The area was subsequently posted and controlled as a high radiation area. This event is documented in the licensee's corrective action program as CR-IP3-2009-00709. 10 CFR 20.1501 requires that surveys be made to comply with the regulations in 10 CFR Part 20. Contrary to this requirement, on April 1, 2009, one of two partially filled yellow drums on the 46' Vapor Containment (VC) walkway, outside crane wan, was identified by an Entergy quality assurance (QA) personnel as not surveyed or labeled. A radiation protection technician surveyed the drum and identified a contact reading of 70 mrem/hr and 20 mrem/hr at 30 cm. Entergy personnel subsequently labeled the drum, informing personnel of the radiological conditions associated with the drum, and relocated the drum to reduce personnel exposure. This event is documented in the licensee's corrective action program as CR-JP3-2009-01527. Failure to adhere to the regulatory requirements specified in 10 CFR 20.1501 constitutes a performance deficiency. These findings are more than minor because they are associated with the Occupational Radiation Safety cornerstone attribute pertaining to exposure control, and the performance deficiency adversely affected the cornerstone objective in that the failure to survey and subsequently inform personnel of the associated radiological conditions did not ensure adequate protection of worker health and safety from exposure to radiation from the radioactive materials. These findings are of very low safety significance (Green) because they did not involve a personnel exposure, i.e., exposure in excess of regulatory limits, or substantial potential for such over-exposure; and the ability to assess personnel exposure was not compromised. AITACHMENT:  


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL
INFORMATION=


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
Entergy Personnel
Site Vice President
: [[contact::J. Pollock Site Vice President A. Vitale General Manager]], Plant Operations  
General Manager, Plant Operations
: [[contact::K. Davison Assistant General Manager]], Plant Operations  
Assistant
: [[contact::P. Conroy Director of Nuclear Safety Assurance T. Orlando Director]], Engineering  
General Manager, Plant Operations
: [[contact::B. Sullivan Emergency Planning Manager A. Williams Site Operations Manager S. Verrochi System Engineering Manager H. Anderson licensing Specialist R. Christman Training Manager J. Cottam Fire Protection G. Dahl Licensing Specialist J.Dinelli Assistant Operations Manager E. Goetchius Training Instructor G. Hocking Supervisor]], Radiation Protection Support D.Loope Manager, Radiation Protection  
Director of Nuclear Safety Assurance
: [[contact::T. McCaffrey Acting Director]], Nuclear Safety Assurance  
Director, Engineering
: [[contact::T. McKee LOR Program Administrator B.Osmin Senior Lead Engineer S. Quinn Security Supervisor J. Reynolds Acting Manager]], Corrective Actions & Assessment  
Emergency
: [[contact::S. Sandike Specialist]], Effluent & Environmental Monitoring A. Singer Licensed Operator Requalification Training Supervisor  
Planning Manager Site Operations
==LIST OF ITEMS==
Manager System Engineering
OPENED, CLOSED AND DISCUSSED Opened 050002NOV Incomplete Licensed Operator Medical Examinations  
Manager Licensing
Specialist
Training Manager Fire Protection
Licensing
Specialist
Assistant
Operations
Manager Training Instructor
Supervisor, Radiation
Protection
Support Training Instructor
Manager, Radiation
Protection
Acting Director, Nuclear Safety Assurance
LOR Program Administrator
Senior Lead Engineer Security Supervisor
Acting Manager, Corrective
Actions & Assessment
Dry Cask Superintendent
Specialist, Effluent & Environmental
Monitoring
Licensed Operator Requalification
Training Supervisor  
 
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
: 05000247/2009-005-01
NOV Incomplete
Licensed Operator Medical Examinations  
 
===Opened and Closed===
===Opened and Closed===
: 05000286/2009005-02 NCV Untimely Compensatory Measures for Degraded EDG Pressure Switches
: 05000247/2009-005-02
: 05000286/2009005-03 NCV Siren Test Failure
NCV Siren Test  
: 05000286/2009005*04 NCV Failure to Promptly Identify and Correct a MCCB Service Life Nonconformance
: 05000247/2009-005-03
: 05000286/2009005-05 FIN Inadequate Post-Maintenance Testing and Resultant Failure of 6.9kV Breaker Auto-Transfer Following Plant Trip Attachment 
NCV Partial Loss of Control Room Indication
: 05000286/2009005-06 FIN Inadequate Maintenance on MBFPs Resulted In an Unexpected Downpower Transient and Reactor Trip.
during Nl 41 Recorder Replacement
===Closed===
: 05000247/2009-005-04
: [[Closes LER::05000286/LER-2008-006]]-01 LER Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing Due to Inadvertent De-Engergization of the Normal Supply Breaker to 480 Volt
NCV Transient
: [[Closes LER::05000286/LER-2009-001]]-01 LER Automatic Actuation of an Emergency Diesel Generator and Two Auxiliary Feedwater Pumps During Surveillance Testing due to Inadvertent De-Energization of the Normal Supply Breaker to 480 Volt
Combustibles
: [[Closes LER::05000286/LER-2009-004]]-00 LER Automatic Reactor Trip Due to a High Steam Generator 32 Water Level Caused by Inadequate 31 Main Feedwater Pump Governor Valve Setting and 32 Steam Generator Level Controller Set-up
Stored on the ISFSI pad Attachment
: [[Closes LER::05000286/LER-2009-005]]-00 LER Technical SpeCification (TS) Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable 480 Volt Undervoltage/Degraded Grid Relay Caused by Personnel Error
: [[Closes LER::05000286/LER-2009-006]]-00 LER Automatic Reactor Trip Due to a Turbine-Generator Trip Caused by Actuation of the Generator Protection'System Lockout Relay During a Severe Storm with Heavy Lightning
: [[Closes LER::05000286/LER-2009-007]] -00 LER Automatic Reactor Trip Due to a Turbine Trip As a Result of Turbine Autostop Oil Actuation Caused by a Failed Autostop Oil Fitting
==LIST OF DOCUMENTS REVIEWED==
Section 1 R01: Adverse Weather Preparations Procedures
: OAP-48, Seasonal Weather Preparation, Rev, 5 Condition Reports (CR-I P3-) 2009-04488 2009-04491 Attachment Section 1 R04: Equipment Alignment 3-COL-FW-2, Auxiliary Feedwater System, Rev. 3-COL-RW-002, Service Water System, Rev. 3-COL-SI-001, Safety Injection System, Rev. Other Safety Injection System Health Report 3rd Quarter 2009 Section 1 R05: Fire Protection Procedures Pre-Fire Plan, Rev. 5
: AP-64.1, Fire Protection/Appendix R Systems and Components Governed by Technical Requirements Manual and Technical
: EN-DC-161, Control of Combustibles, Rev. Condition Reports (CR-IP3-) 2009-03917 2009-04705 Drawings 9321-H-17053, Primary Auxiliary Building Fire Barrier at EI. 41'-0, Rev. 0 Work Orders
: 51700409
: 51559549
: 52024760
: 51469319
: 52219526 Other IP3 Fire Hazards Analysis, Rev. 3


==Section 1R06: Internal Flooding Inspection Work Orders
==LIST OF DOCUMENTS==
: 52207496 Other Individual Plant Examination==
REVIEWED Common Documents Used Indian Point Unit 2 Updated Final Safety Analysis Report, Rev. 21 Indian Point Unit 2 Individual Plant Examination of External Events, August 1992 Indian Point Unit 2 Technical Specifications and Bases, Amendment
: IP-RPT-06-00071, Indian Point Unit 3 Probabilistic Safety Assessment (PSA), Rev. 2 Section 1 R07: Heat Sink Performance Audits and Self-Assessments
: 2 Indian Point Unit 2 Technical Requirements Manual, Rev. 9 Indian Point Unit 2 Control Room Narrative Logs Indian Point Unit 2 Plan of the Day Section 1 R01: Adverse Weather Protection 
: LO-IP3LO-2009-00019, IPEC Heat Sink Performance Snapshot Self Assessment Report, dated 6/17/09 Calculations 6604.266-8-SW-021, SWS Model Input Data Calculations and Output Results for Ingersol Rand Pumps, Rev. 6
===Procedures===
: IP-CALC-08-00120, Evaluation of Wall Thinning at Extent of Condition Location 02-010, 150,
: 2-S0P-30.1.
: PAB-151,
: Electric Heat Tracing, Rev. 25 2-COL-11.5, Space Heating and Winterization, Rev. 28
: PAB-152 and
: COL 30.1, Electric Heat Tracing, Rev. 25 2-S0P-11.5, Space Heating and Winterization, Rev. 32 Condition Reports (CR-) IP2*2006-01308
: PAB-153, Rev. 0 Attachment Completed Surveillance Test Prooedures 3-PT-CS032A, Flow Test of SW HDR CK VLVS and Flow Test of Underground Portions of Line 409, performed 3/28/07 and 4/11/09 3-PT-CS032B, Flow Test of SW HDR CK VLVS and Flow Test of Underground Portions of Line 408, performed 3/28/07 and 4/11/09 3-PT-Q016, EDG and VC Temperature Valves
: IP2-2006-04980
: SWN-FCV-1176 & 1176A and
: IP2-2007 -00883
: SWN-TCV-1104 & 1105, performed 3-PT-Q058, 38 Back-Up Service Water Pump Test, performed 3-PT-Q092A, 31 Service Water Pump, performed 3-PT-Q092B, 32 Service Water Pump, performed 3-PT-Q092C, 33 Service Water Pump, performed 3-PT-Q092D, 34 Service Water Pump, performed 3-PT-Q092E, 35 Servioe Water Pump. performed 3-PT-Q092F, 36 Service Water Pump, performed Condition Reports (CR-IP3-) 2007-00453 2007-03961 2007-04274 2007-04411 2007-04542 2008-00276 2008-00120 2008*00745 2008-00873 2008-02026 2008-02185 2008-02193 2008-02358 2008-02383 2008-02514 2009-00411 2009-00535 2009-00682 2009-01538 2009-01618 2009-02115 2009-02327 2009-02408 2009-03808 2009-04165 2009-04705 2009-04713 2009-04726 2009-04738 2009-04739 Design & Licensing Bases
: IP2-2009-00729
: IP3-DBD-304, Design Basis Document for the Service Water System (SWS), Rev. 3
: IP2-2007 -03368 Section 1 R04: Equipment Alignment 
: IPN-90-004, NYPA Letter to USNRC, Service Water System Problems Affecting Safety Related Equipment Generic Letter 89-13, dated 2/6/90
===Procedures===
: IPN-92-040, NYPA Letter to USNRC, Service Water System Problems Affecting Safety Related Equipment Generic Letter 89*13, dated 9/9/92 NRC Generic letter 89-13, Service Water System Problems Affecting Safety Related Equipment, dated 7/18/89 Drawings 9321-F-20223, Flow Diagram Service Water System Nuclear Steam Supply Plant, Rev. 42 9321-F-20333 Sh. 1, Flow Diagram Service Water System, Rev. 49 9321-F-20333 Sh. 2, Flow Diagram Service Water System, Rev. 27 9321-F-20350, Yard Area Installation of Mechanical Seals in Service Water Piping Line No. 408 Piping Isometric, Rev. 1 Miscellaneous
: 2-PT-Q024C.
: OAP-048, Seasonal Weather Preparation, updated through 12/9/09 0-VLV-446-VCK, Inspection and Repair of 24," 18" and 14" Dual Check Valves, performed 3/27/09
: EDG Fuel Oil Transfer Pump, Rev. 9 2-COL-27.3.1, Diesel Generators, Rev. 25 2-COL-4.2.1.
: CEP-BPT-0100, Buried Piping and Tanks Inspection and Monitoring, Rev. 0 Control Room Log Entries Report, dated 11/1/09 -11/8/09 IPEC Emergency Action Levels Chart, Rev. 06-01 IPEC
: Residual Heat Removal System, Rev. 26 2-COL-4.1.1, Component Cooling System, Rev. 22 2-S0P-4.1.2, Component Cooling System Operation.
: GL 89-13 Program: IP3 Inspection History Report, dated 12/9/09
: Rev. 34 Condition Reports (CR-)
: ISYS-APL-08-001, Site Intake Infrastructure and Material Condition Management, Rev. 1
: IP2-2009-05261
: LO-IP3LO-2007 -00258, Effectiveness Review for
: IP2-2009-02977
: CR-IP3-2007 -00453, dated 10/17/07 Remote Visual I nspection Report of 10" Service Water Line #1099 during 3R15. dated 4/13/09 Attachment Report No. 0900235.401.RO, Structural Integrity Associates, Inc., G-Scan Assessment of Various Buried Piping, dated 11/16/09 Sampling Results for June-September for the Indian Point Zebra Mussel Monitoring Program, dated
: IP2-2008-02406
: SEP-SW-001, NRC Generic Letter 89-13 Service Water Program, Rev.
: IP2-2008-01705
: TS-MS-027, Specification for Service Water Piping & Piping Components, Rev. Visual Inspection Report of 24" Service Water Line #408 during 3R15, dated Nondestructive Examination
: IP2-2009-03666
: IP3-PT-09-038, Liquid Penetrant Examination of 34 FCU Return Piping, dated
: IP2-2008-02091
: IP3-RT-09-002. Radiographic Examination of SW Weld 01-010, dated
: IP2*2008-02037
: IP3-RT -09-007, Radiographic Examination of SW Weld 03-001, dated
: IP2-2008-02054 
: IP3-RT-09-01S, Radiographic Examination of SW Weld
===Drawings===
: VC-34-S, dated
: 9321-LL-3133-20, Diesel Generator Compressor Fuel, Oil Pump and Jacket Water and Lube, Oil Heaters, Sheet 4 IP2--S-000284-14.
: IP3-UT-07-038, Ultrasonic Examination of
: Schematic for Diesel Generator
: SW 18" Une #406, dated
: 9321-F-2030-39, Fuel Oil to Diesel Generators
: IP3-UT-07-146, Ultrasonic Examination of
: 21-F-3220-23, Wiring Diagram. Diesel Generators
: SW 18" Line #409/SWN-39, dated
: 21-22-23.
: IP3-UT-08-016, Ultrasonic Examination of
: Sheet 4 9321-F-3217-06, Wiring Diagram, Diesel Generators
: SW 18" Line #408/SWN-38, dated
: 21-22-23*.
: IP3-UT-OS-019, Ultrasonic Examination of SW 1S" Line #409/SWN-39, dated
: Sheet 1 A227781-82, Flow Diagram Auxiliary Coolant System. Sheet 1 Miscellaneous
: IP3-UT-08-026, Ultrasonic Examination of
: 2-ARP-003, Diesel Generator, Low Fuel Level, Rev. 6 2-ARP-SHF, CCR Electrical, Rev. 29 2-IC-PC-I-L-1207S, Diesel Generator Fuel Oil Day Tank No. 21 Level. Rev. 3 Maintenance Rule BasiS Document Component Cooling Water (CCW). Rev. 02 Attachment Section 1 R05: Fire Protection 
: SW 18" Line #409/SWN-39, dated
===Procedures===
: IP3-UT-OS-035, Ultrasonic Examination of
: 2-S0P-27.1.6, Instrument Bus DC Distribution System, Rev. 39 Condition Reports (CR-)
: SW 18" Line #409/SWN-39, dated
: IP2-2009-04233
: IP3-UT-OS-062, Ultrasonic Examination ofSW 18" Line #409/SWN-39, dated
: IP2-2009-05007* 
: IP3-UT -09-011, Ultrasonic Examination of SW 1S" Line #409/SWN-39, dated
===Miscellaneous===
: IP3-UT-09-021, Ultrasonic Examination of SW Weld
: IP2-RPT-03-00015, IP2 Fire Hazards Analysis, Rev. 2
: PAB-34, dated
: PFP-253, (PEC Pre-Fire Plans, Rev. 0
: IP3-VT-09-042, Visual Examination (VT-2) of SWWeids
: PGI-00433, Combustible Loading Calculation, Rev. 4 Pre Fire Plan
: SWN-42-1,
: PFP-252, Cable Spreading Room -Control Building, Rev. 0 Section 1 R07: Heat Sink Performance 
: SWN-43-1,
===Calculations===
: SWN-216 &
: FFX-00361-00, Minimum Wall Thickness 
: SWN-217, dated 11/3/09 Normal and Special (Abnormal) Operations Procedures 3-AOP-SW-1, Service Water Malfunction, Rev. 2 3-AOP-SWL-1, Low Service Water Bay Level, Rev. 1 3-ARP-012, Panel SJF -Cooling Water and Air, Rev. 47 3-ARP-049, Panel Local -Intake Structure, Rev. 5 3-COL-RW-002, Service Water System, Rev. 43 3-S0P-RW-002, Intake Structure Operation, Rev. 24 3-S0P-RW-005, Service Water System Operation, Rev. 34
===Calculations===
: OAP-OOS, Severe Weather Preparations, Rev. 6
: OAP-04S, Seasonal Weather Preparation, Rev. 5 Operating Experience Generic Service Water System Risk-Based Inspection Guide, NUREG/CR-5865
: EGG-2674 NUREG/CR-054S Ice Blockage of Water Intakes,
: LO-NOE-2007-00078, OE Impact Evaluation for NRC Information Notice 2007-06, dated 4/30/07
: LO-NOE-200S-00173, OE Impact Evaluation for NRC Information Notice 200S-11 , dated 11/10/0S NRC Information Notice 2007-06: Potential Common Cause Vulnerabilities in Essential Service Water Systems, dated 2/9/07 NRC Information Notice 2008-11: Service Water System Degradation at Brunswick Steam Electric Plant Unit 1, dated 6/1S/08 Operating Experience Feedback Report -Service Water System Failures and Degradations,
: NUREG-1275 Vol. 3 Attachment 
: 0-GNR-406-ELC, Emergency Diesel Generator 6-Year Inspection, Revs. 1 & 3-PT-Q058, 37 and 39 Backup Service Water Pumps Test, Rev.
: EN-DC-150, Condition Monitoring of Maintenance Rule Structures, Rev.
: EN-DC-340, Microbiologically Influenced Corrosion (MIC) Monitoring Program, Rev.
: EN-DC-343, Buried Piping and Tanks Inspection and Monitoring Program, Rev.
: ENN-DC-185, Through-Wall Leaks in ASME Section XI Class 3 Moderate Energy Systems, Rev.
: IP-RPT-09-00070, IP 3 Fourth Ten-Year Interval In-service Testing Program Plan, Rev.
: VLV-025-GEN, Inspection and Repair of 24," 18>> and 14" Dual Check Valves, Rev. Risk and Margin Risk-Informed Inspection Notebook for Indian Point Nuclear Power Plant Unit 3, Rev. System Health Reports, Maintenance History & 3PT -C01 Attachment 2, Service Water Leakage Summary Sheet (in VC only), dated 3/30/05,3/27/07, and 3rd Quarter, 2008 Engineering Department Quarterly Trend Report, dated Equipment History Summary Report for
: SWN-4,
: SWN-5,
: SWN-6,
: SWN-35-1 , SWN100-3,
: SWN-100-4, dated 12/7/09 IP3 Service Water System Leak History Graph, dated 11/20/09 IPEC Service Water System Component Leak History -Unit 3, dated 11/20/09
: IP-RPT-07-00078, Maintenance Rule Structural Monitoring Inspection Report (Third Cycle) for Intake Structure, dated 7130/07 Special Log 09-020, Upstream
: SWN-34-1 Service Water Leakage, dated 12/9/09 Special Log 09-026,
: SWN-33-1/SWN-32 Service Water Leakage, dated 12/9/09 Special Log 09-052, SW Leak Upstream of
: SWN-213, dated 12/9/09 Special Log 09-064,
: SWN-62-3 Leakage, dated 12/9/09 SW Bay Level Graph, 10/14/09 -12/9/09 Unit 3 Service Water System Health Report, 3rd Qtr 2009 Unit 3 Service Water Walkdown Report (Intake Structure and SW Pump and Strainer Pits), dated 10/22/09 Unit 3 Service Water Walkdown Report (Turbine Building HXs and Piping), dated 9/17/09 Section 1 R11: Licensed Operator Regualification Miscellaneous IPEC Simulator Evaluated Scenario, IP3 Lesson Plan 13SX-LOR-SES004, Rev. 1 Radiological Emergency Data Forms, Part 1, Notifications #1 and #2


==Section 1R12: Maintenance Effectiveness Condition Re[2orts 2005-05180 2009-04523 Attachment==
for Tubes of Jacket Water Cooler and Lube Oil Cooler for EDG, Rev. 0 .
: FMX-00295-00, Tube Plugging Limits for EDG LO and JW Cocllers, Rev. 0
: PGI-00087-00, EDG LO Cooler Sizing, Rev. 0
: PGI-00387-00, Testing of the CCW Heat Exchangers at Power Operation, Rev. 0 Vendor Manual 755C, Instruction Book for ConEd Component Cooling HXS, Rev. 0 Test Results 2-HTX-004-CCW, Component Cooling Water Heat Exchanger Maintenance, dated 2/23/05. 12112/06, and 2/17/09 O-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, dated 1/8107. 7/16/08, 1/4109, and
: 1217109 Eddy Current I nspection Reports for 22 CCW HX, dated 11/15/02,
: 2115105, 12/13/06, and
: 2110109 Eddy Current Inspection Reports for 23 EDG Jacket Water and Lube Oil HXs, dated 11/6/02, and 12/7109 Modifications
: EC10675, Timed Operation of the Zurn Strainer Circuitry, Rev. 4 EC12566, Material Upgrade of Service Water Strainer Slowdown System, Rev. 5 Conditions Reports (CR-)
: IP2-2004-05064
: IP2-2006-03916
: IP2-2006-03917
: IP2-2006-03929
: IP2-2006-03941
: IP2-2006-03962
: IP2-2006-03964
: IP2-2006-03965
: IP2-2006-03974
: IP2-2006-07009
: IP2-2009-03355
: I P2-2009-02085
: I P3-2009-04739*
: System Health Reports Unit 2 Service Water System, First Quarter 2009, Second Quarter 2009, Third Quarter 2009 Attachment  
===Drawings===
: 9321-F-2033, Service and Cooling Water, Rev. 80 9321-F-2722, Nuclear Service Water System (sheet 1), Rev. 117
: Nuclear Service Water System (sheet 2), Rev. 69 A234191, Flow Diagram Closed Cooling Water System. Rev. 45 Procedures
: 2-AOP-CCW-1, Loss of Component Cooling Water, Rev. 1 2-AOP-SW-1, Service Water Malfunction.
: Rev. 3 2-AOP-UCCW-1, Leakage into CCW System. Rev. 3 2-S0P-24.1, Service Water System Operation.
: Rev. 57 2-S0P-27.3.1.3, 23 EDG Manual Operation, Rev. 19 2-S0P-4.1.2, Component Cooling System Operation.
: Rev. 34 Program Documents
: EN-DC-150, Condition Monitoring of Maintenance Rule Structures, Rev. 0
: IP3-RPT-UNSPEC-03499, Indian Point Units 2&3 Eddy Current Program, Rev. 1
: SEP-SW-001, NRC Generic Letter 89-13
: Service Water Program, Rev. 2
: EN-DC-340, Microbiologically Influenced Corrosion Monitoring Program, Rev. 0 Miscellaneous ConEd Letter, S. Bram to NRC, dated 2/2/90, Response to
: GL 98*13 ConEd Letter, S. Bram to NRC, dated 7/19/91, Implementation Status of
: GL 98-13 ConEd Letter, S. Bram to NRC. dated 2111/92, Updated Implementation Status of
: GL 98-13 ConEd Letter, S. Bram to NRC, dated 9/7/94, Service Water System Operational Performance Inspection
lO-IP3-2009-00019, IPEC Heat Sink Performance, dated 6117/09 Section 1 R11: Licensed Operator Regualification Program Procedures
: 2-E-0, Reactor Trip or Safety Injection, Rev. 2 2-E-1, Loss of Reactor or Secondary Coolant, Rev. 0 2-ES-1.3, Transfer to Cold Leg Recirculation, Rev. 2 2-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, Rev. 0 2-AOP-LEAK-1, Sudden Increase in Reactor Coolant System Leakage, Rev, 7 2-AOP-INST
-1, Instrument/Controller Failures.
: Rev. 5 Miscellaneous
: LRQ-SES-22, SG Pressure Channel Failure, RCS Leak, LBLOCA, Transition to
: ES-1.3 with Equipment Failures, Rev. 2 Radiological Emergency Data Form Drill, Notification
#2, 10/6109 at 8:02 Radiological Emergency Data Form Drill, Notification
#3,10/6109
at 8:28


===Work Orders===
==Section 1R12: Maintenance==
: 51548713 Other Service Water System Health Report 3'd Quarter 2009
: Effectiveness 
===Procedures===
: 2-PT-M110, Appendix R DG Functional Test, Rev. 2 2-PT-M110, Appendix R DG Functional Test, Rev. 1, performed
: 06/12/08 Attachment
===Condition Reports===
(CR-)
: IP2-2009-03053
: IP2-2009-00721
: IP2-2009-00199
: IP2-2009-04021
: I P2-2009-04038
: I P2-2009-4042
: I P2-2009-04259
: IP2-2009-04
: 744 I P2 -2009-04806 
===Drawings===
: 501424, Station Blackout & Appendix R Diesel Generator Set PY800 Manual Double Wall U/L Listed -Fuel Oil Day Tank Mechanical, Rev. 0
: 501425, Station Blackout & Appendix R Diesel Generator Set Wiring Digram Fuel Oil Day Tank Electrical, Rev. 0 Miscellaneous Operators Risk Report, dated 10/2/09 Section 1 R13: Maintenance Risk Assessments and Emergent Work Control Procedures
: IP-SMM-WM-101, On-Line Risk Assessment, Rev. 2
: IP-SMM*WM-103, Control of Maintenance Activities Under Allowable Outage Time (AOT) Action Statements, Rev. 1 Condition Reports (CR-)
: IP2-2009-04420 
===Miscellaneous===
: Operator Risk Report for October 5, 2009


==Section 1R13: Maintenance Risk Assessment and Emergent Work Control Other==
==Section 1R15: Operability==
: IPTE Supplemental Controls and Pre-Job Brief Sheets for 138KV Feeder 33332LM HIPOT including Bus Bar Removal Operator's Risk Report, U3 Thursday 15:00 to 19:00 Section 1 R15: Operability Evaluations Condition Reports 2008-006982008-007172009-031082009-03911 2009-04165 2009-04351 2009-04362 2009-04499 Procedures 0-GNR-404-ELC, Emergency Diesel Generator 2-Year Inspection, Rev. 2 Work Orders
: Evaluations
: 210601 Other ASME Code Case N-513-2 Section 1 R18: Plant Modifications Condition Reports (CR-IP3-) 2009-04498 Work Orders
===Procedures===
: 00215794 Other
: EN-OP-111, Operational Decision-Making Issue (ODMI) Process, Rev. 3
: EC-18677 Section 1 R19: Post-Maintenance Testing Work Orders
: OAP-005, Narrative Logs. Rev. 2
: 198373
: OAP-017, Plant Surveillance and Operator Rounds, Rev. 6 Condition Reports (CR-)
: 210691 Other
: IP2-2009-05300
: EC 5000038856 (DC 97-3-039), IP3 Foxboro to NUS Bistable Replacement Program, Rev. 0
: IP2-2009-3527*
: IP2-2009-2469
: IP2-2009-3564
: IP2-2008-4212 
===Drawings===
: A208088-43, 480 Vac. Switchgears
& 22, Bus 2A, 3A, 5A & 6A 9321-F-2030, Flow Diagram Fuel Oil Diesel Generators, Rev. 39 Miscellaneous
: 3.8 Electrical Power, Technical Requirements Manual (TRM), Rev. 1
: EC 5000033794, IP2 Station Blackout and Appendix R Diesel Generator Set, Rev. 1 Calculations
: IP-CALC-06-00299, EDG Fuel Oil Day Tank Low Level Analytical Limit, Rev. 0 Work Orders
: 00199669 Attachment


==Section 1R22: Surveillance Activities Procedures==
==Section 1R19: Post-Maintenance==
: 3-PT-Q83, RWST Level Instrument Check and Calibration (UC-921), Rev. 28 Attachment 
: Testing Procedures
: 3-PT-Q120B, 32 ABFP (Turbine Driven) Surveillance and 1ST, Rev. 13 0-PMP-411-BFP, Turbine Driven Auxiliary Boiler Feed Pump Overhaul/Inspection, Rev. 1 0-SOP-LEAKRATE-001, RCS Leakrate Surveillance. Evaluation, and Leak Identification, Rev. 1 3-PT-Q080, Pressurizer Block Valve Timing Test
: 2-PMP-008-CCW, Inspection/Repair of the Component Coofing Pump, Rev. 2
: RC-MOV-535 and 536, Rev. 6 3-PT-M62A. 480V Undervoltage / Degraded Grid Protection System Bus 2A and 3A Functional, Rev. 7 Condition Reports (CR-IP3-) 2009-03179 2009-04592
: MSl-B-007-A, Chesterton Seals (Series 123), Rev. 7
: CR-HQN-2009-00953 Work Orders
: CUP-B-002-A, Falk Type T10ff20 Steefflex Couplfng, Rev. 8 2-PT-Q034B,
: 52191478 52216249
: PCV-1310A
and
: PCV-1310B
: Nitrogen Supply, Rev. 6 2-PT-Q034, 22 Auxiliary Feed Pump, Rev. 26
: Fire Protection System Operation, Rev. 22 O-Vl V-413-MOV, Motor Operated Valve Minor Preventive Maintenance, Rev. 4
: PT-M54, Fan Cooler Units Operation, performed
: 11/16/09 2-PT -M021 B, Emergency Diesel Generator load Test, performed
: 11/10/09 2-PT-2Y018F, Transfer Switch
: EDD-6 (22 EDG) Test, performed
: 11110109 O-Vl V-465-VSR,
: CLA-VAL Pressure Relief Valves Maintenance and Inspection, performed
: 10/30/09
: CLA-VAL Pressure Relief Valves Maintenance and Inspection, performed
: 11/04/09 Condition Reports (CR-)
: IP2-2009-04215
: IP2-2009-4505
: IP2-2009-5198 
===Work Orders===
: 00130508-01
: 00195301-01
: 51549691
: 51549692
: 51549733
: 51549690
: 52203978
: 00158729
: 00158730
: 00158731
: 00158732
: 51268377
: 51548550
: 52037213
: 51702237
: 52028737-01 
===Miscellaneous===
: PGI-00518, AOV Component Level Calc. for 22 Auxiliary Feed Pump Discharge Flow Control Valves to Steam Generators,
: FCV-405A,
: FCV-405B,
: FCV-405C and
: FCV-405D, Rev. 1 Drawings A227551, Fire Protection System Diagram, Rev. 63 Section 1 R22: Surveillance Testing Procedures
: 2-PT-Q013-DS085, Valve
: FCV-405A 1ST Data Sheet, Rev. 20
: CR-IP2-2009-00666, Root Cause Analysis Report, CST Underground Recirculation Line Leak, Rev. 0
: EN-DC-325, Component Performance Monitoring, Rev. 4
: EN-DC-332, Inservice Testing, Rev. 0 2-PT-Q001C, 23 Battery Surveillance and Charging,
: 2/14/09 2-PTOQ028A, 21 Residual Heat Removal Pump, performed
: 11/19/09 2-PT-Q029A, 21 Safety Injection Pump, performed
: 10/22/09 Miscellaneous
: IP2-AFW DBD, Auxiliary Feedwater System, Rev. 1 PG/-00497.
: Auxiliary Feedwater System Air Operated Valve Functional and Maximum Expected Differential Pressure Calculation, Rev. 1 IEEE 450, IEEE Recommended Practice for Maintenance, Testing and Replacement, dated 1995 Attachment
===Drawings===
: 251783. Flow Diagram Auxiliary Coolant System Residual Heat Removal Pumps, Rev. 0 Section 1 EP2: Alert and Notification System Evaluation 
===Procedures===
: Alert and Notification System for the Indian Point Energy Center Entergy Nuclear, Rev. 4 Indian Point Energy Center Emergency Preparedness Plan, Rev. 8
: IP-EP-AD30, lPEC ATI Siren System Administration, Rev. 2
: IP-EP-AD31, IPEC ATI Siren System Maintenance Administration, Rev. 0 Alert Notification System Test Failure Root Cause Evaluation Report, Rev. 1
: IP-EP-AD35, IPEC ATI Siren Site Annual Preventive Maintenance, Rev. 2 IPEC ATI Siren Annual Preventive Maintenance Test Records, February 10, 2009 ANS related Condition Reports, December 2007 -December 2009


==Section 1EP2: Alert and Notification System (ANS) Evaluation Procedures Alert and Notification System for the Indian Point Energy Center Entergy Nuclear, Rev. 4 Indian Point Energy Center Emergency Preparedness Plan, Rev. 8==
==Section 1EP3: Emergency==
: IP-EP-AD30, IPEC ATI Siren System Administration, Rev. 2
: Preparedness Organization Staffing and Augmentation System Procedures
: IP-EP-AD31, IPEC ATI Siren System Maintenance Administration, Rev. 0 Alert Notification System Test Failure Root Cause Evaluation Report, Rev. 1
: IP-EP-AD9, Notification Systems Testing and Maintenance, Rev. 7 IPEC ERO Roster Indian Point Energy Center Emergency Response Training Program Curriculum, Rev. 24 October 27, 2009, Entergy Nuclear Northeast, Indian Point Energy Center Emergency Preparedness Unit 3 Off-Hours Mobilization StaffingITralning Drill Performance Report. Drill Number 2009-5 September  
: IP-EP-AD35, IPEC ATI Siren Site Annual Preventive Maintenance, Rev. 2 IPEC ATI Siren Annual Preventive Maintenance Test Records, February 10, 2009 ANS related Condition Reports, December 2007 -December 2009 Section 1 EP3: Emergency Response Organization (ERO) Staffing and Augmentation S,)&#xa3;stem Procedures
: 17. 2009. Indian Point Energy Center Emergency Response Organization Off-hours Notification Test 3Q09 Section 1 EP4: Emergency Action Level and Emergency Plan Changes Procedures  
: IP-EP-AD9, Notification Systems Testing and Maintenance, Rev. 7 Indian Point Energy Center Emergency Response Training Program Curriculum, Rev. 24 October 27,2009, Entergy Nuclear Northeast, Indian Point Energy Center Emergency Preparedness Unit 3 Off-Hours Mobilization StaffjnglTraining Drill Performance Report, Drill Number 2009-5 September 17, 2009, Indian Point Energy Center Emergency Response Organization Off-hours Notification Test 3Q09 Section 1 EP4: Emergency Action Level (EAL) and Emergency Plan Changes Procedures
: EN-EP-305.
: EN-EP-305, Emergency Planning 10CFR50.54 (q) Review Program, Rev. 1 10
: Emergency Planning 10CFR50.54 (q) Review Program, Rev. 1 10
: CFR 50.54(q) screenings and evaluations from December 2008 to November 2009 Section 1 EP5: Correction of Emergency Preparedness Weaknesses Procedures
: CFR 50.54(q) screenings and evaluations from December 2008 to November 2009 Section 1 EP5: Correction of Emergency Preparedness Weaknesses
: EN-U-102, Corrective Action Process, Rev. 13
===Procedures===
: QA-07 -2008-IP-1, Quality Assurance Audit Report
: EN-Ll-102, Corrective Action Process, Rev. 13
: QA-07-2008-IP-1, Quality Assurance Audit Report
: QA-07-2009-IP-1, Quality Assurance Audit Report
: QA-07-2009-IP-1, Quality Assurance Audit Report
: QS-2008-IP-16, IPEC QA Follow-up of AFI from Emergency Plan Surveillance
: QS-200B-IP-16, IPEC QA Follow-up of AFI from Emergency Plan Surveillance
: QS-2008-IP-16
: QS-2008-IP-16  
: QS-2008-IP-02, QA Evaluation of the IPEC 2/6/08 Training Drill Attachment
: QS-2008-IP-02.
: LO-IP3LO-2007-00185, IPEC Snapshot Self-Assessment Report, ANS Siren System Performance
: QA Evaluation of the IPEC 2/6/08 Training Drill LO-IP3LO-2007-00185.
: IP3-LO-2009-00092. IPEC Focused Self-Assessment Report, EP INPO Based Focus Self Assessment October 29.2008, Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 3 Training Drill Performance Report, Drill Number 2008-5 November 19, 2008, Entergy Nuclear Northeast. Indian Point Energy Center, Emergency Preparedness Unit 3 Training Drill Performance Report, Drill Number 2008-6 December 3, 2008, Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 3 Training Drill Performance Report, Drill Number 2008-7 May 13, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 2 Training Drill Performance Report, Dnll Number 2009-2 September 9, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Unit 2 Training Drill Performance Report, Dnll Number 2009-3 Sections 2051/2052: Access Control to Radiologically Significant Areas/ALARA Planning and Controls Procedures
: IPEC Snapshot Self-Assessment Report. ANS Siren System Performance  
: EN-Ll-114, Performance Indicator Process
: IP3-LO-2009-00092.
: EN-RP-100, Radworker Expectations
: IPEG Focused Self-Assessment Report. EP INPO Based Focus Seif Assessment October 29.2008, Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 3 Training Drill Performance Report. Drill Number 2008-5 November 19. 2008. Entergy Nuclear Northeast, Indian Point Energy Center. Emergency Preparedness Unit 3 Training Drill Performance Report, Drill Number 2008-6 December 3, 2008, Entergy Nuclear Northeast, Indian Point Energy Center. Emergency Preparedness Unit 3 Training Drill Performance Report. Drill Number 2008-7 Attachment May 13, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 2 Training Drill Performance Report, Drill Number 2009-2 September  
: EN-RP-101, Access Control for Radiologically Controlled Areas
: 9, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 2 Training Drill Performance Report, Drill Number 2009-3 Sections 2051/2052:
: EN-RP-102, Radiological Control
: Access Control to Radiologically Significant Areas/ALARA  
: EN-RP-105, Radiation Work Permits
: Planning and Controls Procedures
: EN-RP-108, Radiation Protection Posting
: EN-U-114, Performance Indicator Process, Rev. 4
: EN-RP-110, ALARA Program
: EN-RP-100, Radworker Expectations, Rev. 3
: EN-RP-121, Radioactive Material Control
: EN-RP-101, Access Control for Radiologically Controlled Areas, Rev. 4
: EN-RP-122, Alpha Monitoring
: EN-RP-102, Radiological Control, Rev. 2 EN-RP-105.
: EN-RP-131, Air Sampling
: Radiation Work Permits, Rev. 7
: EN-RP-141, Job Coverage
: EN-RP-108, Radiation Protection Posting, Rev. 7
: EN-RP-151, Radiologica I Diving
: EN-RP-110, ALARA Program, Rev. 6
: EN-RP-202, Personnel Monitoring
: EN-RP-121, Radioactive Material Contro', Rev. 4
: EN-RP-203, Dose Assessment
: EN-RP-122, Alpha Monitoring, Rev. 3
: EN-RP-204, Special Monitoring Requirements
: EN-RP-131, Air Sampling, Rev. 7
: EN-RP-205, Prenatal Monitoring
: EN-RP-141, Job Coverage, Rev. 4
: EN-RP-208, Whole Body Counting and In-Vitro Bioassay O-RP-RWP-411, Discrete Radioactive Particle Controls O-RP-RWM-901, Interim Radwaste Storage Facility and Outside Radioactive Material Storage Area
: EN-RP-151, Radiological Diving, Rev.2
: RP-M-460, Controls for High Radiation and Locked High Radiation Areas Condition Reports (CR-)
: EN-RP-202, Personnel Monitoring, Rev. 7
: EN-RP-203, Dose Assessment, Rev. 3
: EN-RP-204, Special Monitoring Requirements, Rev. 3
: EN-RP-205, Prenatal Monitoring, Rev. 4
: EN-RP-208, Whole Body Counting and In-Vitro Bioassay, Rev. 3 O-RP-RWP-411, Discrete Radioactive Particle Controls, Rev. 0 O-RP-RWM-901, Interim Radwaste Storage Facility and Outside Radioactive Material Storage Area, Rev. 2 Condition Reports {CR-}
: IP2-2009-02184
: IP2-2009-02184
: IP2-2009-02217
: IP2-2009-02217
Line 494: Line 1,058:
: IP2-2009-02484
: IP2-2009-02484
: IP2-2009-02505
: IP2-2009-02505
: IP2-2009-03335 I P2-2009-03524
: IP2-2009-03335  
: IP2-2009-03578
: I P2 -2009-03524  
: I P2-2009-03578
: IP2-2009-03674
: IP2-2009-03674
: IP2-2009-03699
: IP2-2009-03699
: IP2-2009-03978
: IP2-2009-03978
: IP3-2009-00709
: IP3-2009-00709
: IP3-2009*01348
: IP3-2009-01348
: IP3-2009-01439
: IP3-2009-01439
: IP3-2009-01527
: IP3-2009-01527  
: IP3-2009-01769
: I P3-2009-0
: 1769
: IP3-2009-01879
: IP3-2009-01879
: IP3-2009-01981
: IP3-2009-01981  
: IP3-2009-01984
: I P3-2009-01984
: IP3-2009-02198
: IP3-2009-02198
: IP3-2009-02200
: IP3-2009-02200
: IP3-2009-02201
: IP3-2009-02201  
: IP3-2009-02619
: I P3-2009-02619
: IP3-2009-03110
: IP3-2009-03110
: IP3-2009-03721'
: IP3-2009-03721
: IP3-2009-03778
: IP3-2009-03778  
: IP3-2009-03973 Attachment
: I P3-2009-03973
===Miscellaneous===
===Miscellaneous===
: ALARA Committee Reviews IPEC 5 Year ALARA Plan 2009-2013
: ALARA Committee Reviews IPEC 5 Year ALARA Plan 2009*2013
: IP3-LO-2009-00074, (PEC Snapshot Self-Assessment Report -HRA & LHRA Controls Oversight Observation Checklists: 02C-IPEC-2009-0202, 0205,0223,0224,0241,0266,0279, 0281,0368,0496,0520,0531, Radiation Protection Attention Logs (Electronic Dosimeter Alanns) Monthly Effluent Release Reports
: IP3-LO-2009-00074, IPEC Snapshot Self-Assessment Report -HRA & LHRA Controls Oversight Observation Checklists:  
: 2C-IPEC-2009-0202, 0205, 0223, 0224,0241,0266,0279, 0281,0368,0496,0520,0531  
: Radiation Protection Attention Logs (Electronic Dosimeter Alarms) Monthly Effluent Release Reports


==Section 40A1: Performance Indicator Verification Procedures==
==Section 40A1: Performance==
: O-SOP-LEAKRA
: Indicator Verification
: TE-001, RCS Leakrate Surveillance, Evaluation, and Leak Identification, Rev. 1
===Procedures===
: NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 6
: NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 2 Attachment 
: O-SOP-LEAKRA  
: TE"001, RCS Leakrate Surveillance, Evaluation and Leak Identification, Rev. 1 EN"LI" 114, Performance Indicator Process, dated 10/08/09
: EN-U-114, Performance Indicator Process, dated 04/09/09 EN-U-114.
: Performance Indicator Process, dated 10/09/08
: EN-U-114, Performance Indicator dated 01/14/09
: EN-U-114, Performance Indicator Process, dated 10/08/09
: EN-EP-201, Performance Indicators, Rev. 9
: EN-EP-201, Performance Indicators, Rev. 9
: IP-EP-AD5, Emergency Preparedness Performance Indicator Program, Rev. 3 Perfonnance Indicator Data, 4th quarter 2008 -3rd quarter 2009 Other Indian Point Unit 3 Operating Logs
: IP-EP-AD5, Emergency Preparedness Perfonnance Indicator Program, Rev. 3 Performance Indicator Data, 4th quarter 2008 -3 rd quarter 2009 Condition Reports (CR-) IP2-2009-05032


==Section 40A2: Identification and Resolution of Problems Condition Reports (CR-IP3-) 1996-01880 2008-01287 2009-04450 1998-02235 2008-01287 2009-04262 1999-01165 2008-01235 2002-00362 2008-01108 2004-02896==
==Section 40A2: Identification==
: 2008-00656 2004-02704 2008-00226 Preventive Maintenance Change Requests 21444 68795 68806 Procedures
 
: OAP-045, Operator Burden. Program, Rev. 1
and Resolution of Problems Procedures
: EN-Ll-102, Corrective Action Process, Rev. 13 Miscellaneous Daily Plant Status Report for Monday, December 7, 2009 Operator Aggregate Impact Index IP3, November 2009. Westinghouse Technical Bulletin 04-13, Replacement Solutions for Obsolete Classic Molded Case Circuit Breakers, UL Testing Issues, Breaker Design Life and Trip Band Adjustment, 07/16/2004 Westinghouse Technical Bulletin 06-2, Aging Issues and Subsequent Operating Issues for Breakers That are at Their 20 Year Design/Qualified Lives; UL CertificationlTesting Issues Update, 03/10/2006 Work Orders
: OAP-045, Operator Burden Program, Rev. 1 Condition Reports (CR-)
: 00151859 Attachment 
: IP2-2009-4860 
: ABFP ADAMS ALARA AMP ANS ASME CCW CEDE CFR CR  
===Work Orders===
: CRG DEP DMB DRS EAL EDG ENTERGY EP 'EPZ ERO FIN GL HRA HX IMC IN IP2 IP3 IP IPEC IR ISFSI 1ST LDE LER MBFP MCCB MIC MOV NCV NDE NEI NOV NRC NYPA PI PI&R PM QA A-11
: 00205770
: 00179027
: 00177347 Section 40A3; Event Follow-up 
===Procedures===
: IP-SMM-IS-104, Electrical Work Practices and Procedure Handbook, Rev. 1
: EN-HU-102, Human Performance Tools, Rev. 5
: EN-HU-105, Human Performance
-Managed Defenses, Rev. 6 2-0AP-IB-1.
: Loss of Power to an Instrument Bus, Rev. 7 2-S0P-27.3.1.3, 23 Emergency Diesel Manual Operation, Rev. 19
: EN-WM-105, Planning, Rev. 5 Condition Reports (CR-) I P2 -2009-4860 
===Work Orders===
: 00163807
 
==Section 40A5: Other Activities==
 
===Procedures===
: 2-DCS-006, Vertical Cask Transporter Use 2-DCS-031
: GEN, Fuel Selection for Dry Cask Storage, Rev. 0 10
: CFR 72.212, Evaluation Report, Site Specific Appendix F, iP2 Specific Infonnation
: EN-DC-147, IPEC ISFSI Fire Hazards Analysis Roadmap, dated 12/20/2007, Rev. 2
: EN-DC-161, Control of Combustibles Rev. 3 Holtec International
: HI-STORM FSAR, Report
: HI-2002444, Rev. 4 Transient Combustible Evaluation No. 09-015, dated 10/19/2009 
===Condition Reports===
(CR-) IP2..;2009-03910*
: IP2-2009-05228*
*CR Initiated as a result of this inspection.
: ADAMS ALARA AMP ANS CCW CEDE CFR CR OMB ORP EAL EDG ENTERGY EP EPZ ERO FCU FCV FSAR FZ HRA IMC IPEC ISFSI 1ST LBLOCA LOE NCV NEI NOV NRC PI PM RHR ROP RSPS RWP SDE SDP
: SSC SW TCE TEDE TS UFSAR UHS WO
==LIST OF ACRONYMS==
==LIST OF ACRONYMS==
Auxiliary Boiler Feedwater Pump Agency Wide Document Management System As Low as is Reasonably Achievable Amplifier Alert and Notification System American Society of Mechanical Engineers Component Cooling Water Committed Effective Dose Equivalent Code of Federal Regulations Condition Report Condition Review Group Drill and Exercise Performance Digital Message Board Division of Reactor Safety Emergency Action Level Emergency Diesel Generator Entergy Nuclear Northeast Emergency Preparedness Emergency Planning Zone Emergency Response Organization Finding
 
: [[NRC]] [[Generic Letter High Radiation Area Heat Exchanger Inspection Manual Chapter Information Notice Indian Point 2 Indian Point 3 Inspection Procedure Indian Point Energy Center Inspectton Report Independent Spent Fuel Storage Installation In-Service Test Low Dose Equivalent Licensee Event Report Main Boiler Feedwater Pump Molded Case Circuit Breaker Microbiologically Influenced Corrosion Motor Operated Valve Non-Cited Violation Non-Destructive Examination Nuclear Energy Institute Notice of Violation Nuclear Regulatory Commission New York Power Authority Performance Indicator Problem Identification and Resolution Preventive Maintenance Quality Assurance Attachment]]
Document and Management
: [[ROP]] [[]]
System As Low As is Reasonably
: [[RSPS]] [[]]
Achievable
: [[RT]] [[]]
Amplifier  
: [[RWP]] [[]]
 
: [[SAT]] [[]]
Alert and Notification
: [[SDE]] [[]]
System Component
: [[SDP]] [[]]
Cooling Water Committed
: [[SG]] [[]]
Effective
: [[SGDT]] [[]]
Dose Equivalent
: [[SGWL]] [[]]
Code of Federal Regulations
: [[SI]] [[]]
Condition
: [[SRI]] [[]]
Report Digital Message Board Division of Reactor Projects Emergency
: [[SSC]] [[]]
Action Level Emergency
: [[SW]] [[]]
Diesel Generator
: [[SWS]] [[]]
Entergy Nuclear Northeast
: [[TB]] [[]]
Emergency
: [[TEDE]] [[]]
Preparedness
: [[TI]] [[]]
Emergency
: [[TS]] [[]]
Planning Zone Emergency
: [[UAT]] [[]]
Response Organization
: [[UFSAR]] [[]]
Fan Cooler Unit Flow Control Valve Final Safety Analysis Report Fire Zone High Radiation
: [[UHS]] [[]]
Area Inspection
: [[UT]] [[]]
Manual Chapter Indian Point Energy Center Independent
: [[VC]] [[WO Reactor Oversight Process Risk Significant Planning Standard Radiographic Test Radiation Work Permit Station Auxiliary Transformer Shallow Dose Equivalent Significance Determination Process Steam Generator Small Gas Decay Tank Steam Generator Water Level Safety Injection Senior Resident Inspector Structures, Systems, and Components Service Water Service Water System Westinghouse Technical Bulletin Total Effective Dose Equivalent Temporary Instruction Technical Specifications Unit Auxiliary Transformer Updated Final Safety Analysis Report Ultimate Heat Sink Ultrasonic Test Vapor Containment Work Order Attachment]]
Spent Fuel Storage Installation
: [[UNITED]] [[]]
In-Service
: [[NUCLEA]] [[R]]
Test Large Break Loss-of-Coolant
: [[REGULA]] [[]]
Low Dose Equivalent
: [[TORY]] [[]]
Non-cited
: [[REGION]] [[475]]
Violation
: [[ALLEND]] [[ALE]]
Nuclear Energy Institute
: [[KING]] [[]]
Notice of Violation
: [[OF]] [[]]
Nuclear Regulatory
: [[PRUSSI]] [[A.]]
Commission
: [[PENNSY]] [[LVANIA March 3, 2010 Mr. Michael Colomb Site Vice President Entergy Nuclear Operations, Inc. Vermont Yankee Nuclear Power Station 320 Governor Hunt Road Vernon,]]
Performance
: [[VT]] [[05354]]
Indicator
: [[ANNUAL]] [[]]
Preventative
: [[ASSESS]] [[]]
Maintenance
: [[MENT]] [[]]
Residual Heat Removal Reactor Oversight
: [[LETTER]] [[-]]
Process Risk Significant
: [[VERMON]] [[T]]
Planning Standard Radiation
: [[YANKEE]] [[]]
Work Permit Shallow Dose Equivalent
: [[NUCLEA]] [[R]]
Significance
: [[POWER]] [[]]
Determination
: [[STATIO]] [[N (REPORT 05000271/2010001) Oear Mr. Colomb: On February 9,2009, the]]
Process Structures, Systems, and Components
: [[NRC]] [[staff completed its performance review of the Vermont Yankee Nuclear Power Station (Vermont Yankee). Our technical staff reviewed performance indicators (Pis) for the most recent quarter and inspection results for the period from January 1 through December 31,2009. The purpose of this letter is to inform you of our assessment of your safety performance during this period and our plans for future inspections at your facility. This performance review and enclosed inspection plan do not include security information. A separate letter designated and marked as "Official Use Only -Security Information" will include the security cornerstone review and resultant inspection plan. Overall, Vermont Yankee operated in a manner that preserved public health and safety and fully met all cornerstone objectives. Plant performance for the most recent quarter, as well as for the first three quarters of the assessment cycle, was within the licensee Response column of the]]
Service Water Transient
: [[NRC]] [['s Action Matrix, based on all inspection findings being classified as having very low safety significance (Green) and all Pis indicating performance at a level requiring no additional]]
Combustible
: [[NRC]] [[oversight (Green). Therefore, we plan to conduct reactor oversight process (]]
Evaluation
: [[ROP]] [[) baseline inspections at your facility. On February 24, 2010, Entergy verbally informed the]]
Total Effective
: [[NRC]] [[of actions that Entergy has taken regarding certain employees as a result of its independent internal investigation into the alleged contradictory or misleading information provided to the State of Vermont that was not corrected. While the]]
Dose Equivalent
: [[NRC]] [[does not have jurisdiction over the communications between Entergy and the State of Vermont, the]]
Technical
: [[NRC]] [[is aware that some of these individuals have responsibilities that involve decision-making communications that are material to the]]
Specifications
: [[NRC]] [[andlor involve regulated activities. In light of this information, the]]
Updated Final Safety Evaluation
: [[NRC]] [[issued a Demand For Information (]]
Report Ultimate Heat Sink Work Order Attachment
: [[OF]] [[t) on March 1, 2010. The letter transmitting the]]
: [[OFI]] [[and the details of the]]
: [[OFI]] [[can be found in the]]
: [[NRC]] [['s document system (]]
: [[ADAMS]] [[) under accession number ML 100570237.]]
: [[M.]] [[Colomb 2 As part of our ongoing inspections, the]]
: [[NRC]] [[will continue to review Entergy's implementation of the industry voluntary ground water protection initiative using Temporary Instruction (TI) 173 and follow Entergy's activities to address the tritium in the groundwater at the Vermont Yankee Nuclear Power Station. The enclosed inspection plan details the inspections, less those related to physical protection, scheduled through June 30, 2011. The inspection plan is provided to allow for the resolution of any scheduling conflicts and personnel availability issues well in advance of inspector arrival onsite. Routine resident inspections are not listed due to their ongoing and continuous nature. The inspections in the last nine months of the inspection plan are tentative and may be revised at the mid-cycle review. In accordance with]]
: [[10 CFR]] [[2.390 of the]]
: [[NRC]] [['s "Rules of Practice," a copy of this leUer and its enclosure will be made available electronically for public inspection in the]]
: [[NRC]] [[Public Document Room or from the Publicly Available Records (]]
: [[PARS]] [[) component of]]
: [[NRC]] [['s document system (]]
: [[ADAMS]] [[).]]
: [[ADAMS]] [[is accessible from the]]
: [[NRC]] [[Web site at rm/adams.html (the Public Electronic Reading Room). If circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact me at 610-337-5306 with any questions you may have regarding this letter or the inspection plan. Sincerely, Projects Branch 5 Division of Reactor Projects Docket Nos. 50-271 License No. DPR-28 Enclosure: Vermont Yankee Inspection/Activity Plan cc w/encl: Distribution via ListServ]]
: [[M.]] [[2 As part of our ongoing inspections, the]]
: [[NRC]] [[will continue to review Entergy's implementation of the industry vOluntary ground water protection initiative using Temporary Instruction (TI) 173 and follow Entergy's activities to address the tritium in the groundwater at the Vermont Yankee Nuclear Power Station. The enclosed inspection plan details the inspections, less those related to physical protection, scheduled through June 30, 2011. The inspection plan is provided to allow for the resolution of any scheduling conflicts and personnel availability issues well in advance of inspector arrival onsite. Routine resident inspections are not listed due to their ongoing and continuous nature. The inspections in the last nine months of the inspection plan are tentative and may be revised at the mid-cycle review. In accordance with]]
: [[10 CFR]] [[2.390 of the]]
: [[NRC]] [['s "Rules of Practice," a copy of this letter and its enclosure will be made available electronically for public inspection in the]]
: [[NRC]] [[Public Document Room or from the Publicly Available Records (]]
: [[PARS]] [[) component of]]
: [[NRC]] [['s document system (]]
: [[ADAMS]] [[).]]
: [[ADAMS]] [[is accessible from the]]
: [[NRC]] [[Web site at rm/adams.html (the Public Electronic Reading Room). If circumstances arise which cause us to change this inspection plan, we will contact you to discuss the change as soon as possible. Please contact me at 610-337-5306 with any questions you may have regarding this letter or the inspection plan. Sincerely, /RA! Donald]]
: [[E.]] [[Jackson, Chief Projects Branch 5 Division of Reactor Projects Docket Nos. License No. Enclosure: Vermont Yankee Inspection/Activity S. Collins,]]
: [[RA]] [[(R10RAMAIL]]
: [[RESOUR]] [[]]
: [[CE]] [[)]]
: [[H.]] [[Jones,]]
: [[ORP]] [[,]]
: [[RI]] [[M. Oapas,]]
: [[ORA]] [[(R10RAMAIL]]
: [[RESOUR]] [[]]
: [[CE]] [[)]]
: [[A.]] [[Rancourt,]]
: [[DRP]] [[,]]
: [[OA]] [[O. Lew,]]
: [[ORP]] [[(R1DRPMAIL]]
: [[RESOUR]] [[]]
: [[CE]] [[) RidsNrrPMVennonlYankee Resource]]
: [[J.]] [[Clifford,]]
: [[DRP]] [[(R1DRPMAIL]]
: [[RESOUR]] [[]]
: [[CE]] [[) RldsNrrDorlLl1-1 Resource]]
: [[D.]] [[Roberts,]]
: [[DRS]] [[(R1DRSMall]]
: [[ROP]] [[regortsResource@nrc.gov P. Wilson,]]
: [[DRS]] [[(R1DRSMall]]
: [[DRS]] [[Branch Chiefs (6) L. Trocine, Rl L. Scholl D. Jackson, D. Silk,]]
: [[DRS]] [[]]
: [[T.]] [[Setzer, J. Trapp,]]
: [[DRS]] [[]]
: [[J.]] [[Heinly, D. Screnci,]]
: [[PAO]] [[]]
: [[B.]] [[Sienei, N. Sheenan,]]
: [[PAO]] [[]]
: [[D.]] [[Spindler,]]
: [[DRP]] [[,]]
: [[R.]] [[Barkley,]]
: [[ORA]] [[]]
: [[SUNS]] [[I Review Complete: .... (Revlewer's Initials) Ml No 100621409]]
: [[CS]] [[:.-___]]
: [[DOCUME]] [[]]
: [[NT]] [[]]
: [[NAME]] [[: S:\]]
: [[ROP]] [[*10]]
: [[EOC]] [[RevieW\Branch 5Wy\2009 Annual Assessment Letter]]
: [[VY.]] [[doc Aftar declaring this document "An Official Agency Record" It wilVwiIi not be released to the To receive a copy of this document, Indicate in the box: "C" =Copy without attachment/enclosure "EO =Copy attahc ment/ I "N" enc osure N0 copy]]
: [[OFFICE]] [[]]
: [[NAME]] [[]]
: [[RIIDRP]] [[]]
: [[TS]] [[etzerrrcs I]]
: [[RIIDRP]] [[I]]
: [[DJ]] [[ackson/ALB for]]
: [[RIIDRP]] [[]]
: [[DL]] [[ew/DCL I I]]
: [[DATE]] [[3/2110 312/10 3/3/10]]
: [[OFFIC]] [[iAl]]
: [[RECORD]] [[]]
COPY
__ Page 1 of2 Vermont Yankee 03/02/2010 15:46:26 Inspection / Activity Plan Report 22 01/01/2010 -06/30/2011 Planned Dates Start End Title 01/01/2010 01/25/2010 05/0312010 05/03/2010 05/0312010 05/03/2010 0512412010 08116/2010 07/1212010 07/1212010 11/0112010 11/29/2010 08115/2010 08115/2010 08/15/2010 08/15/2010 08/1512010 08/15/2010
08/15/2010 09/13/2010
09/27/2010 09/1312010 0911312010 12131/2010 03/1212010 05/07/2010 05/07/2010 0510712010 05/0712010 0512812010* 08/20/2010 07/16/2010 07/16/2010 11/05/2010 1211012010 0812012010 0812012010 0812012010 0812012010 08120/2010 08/2012010 0812012010 09/1712010 10/01/2010 09117/2010 09/17/2010
: [[ISFSI]] [[-]]
: [[ANNUAL]] [[]]
: [[SAMPLE]] [[.]]
: [[ISFSII]] [[NSPECTION]]
: [[1 IP]] [[60855 Operation Of An]]
: [[ISFSI]] [[]]
: [[TI]] [[-173 -]]
: [[IMPLEM]] [[ETATION]]
: [[OF]] [[]]
: [[INDUST]] [[RY]]
: [[GROUND]] [[]]
: [[WATER]] [[lP 2515/173 Review of the I mplem entation of the lnd ustry Ground Water Protection Voluntary Initiative]]
: [[711110BG]] [[-]]
: [[INSERV]] [[ICE]]
: [[INSPEC]] [[]]
: [[TION]] [[]]
: [[IP]] [[7111108G Inservice Inspection Activities -]]
: [[BWR]] [[71124 -HP]]
: [[OUTAGE]] [[]]
: [[INSPEC]] [[TION]]
: [[ACCESS]] [[/AiR/]]
: [[ALARA]] [[]]
: [[IP]] [[71124.01 Radiological Hazard Assessment and Exposure Controls]]
: [[IP]] [[71124.02 Occupational]]
: [[ALARA]] [[Planning and Controls]]
: [[IP.]] [[71124.03 In-Plant Airborne Radioactivity Control and Mitigation 71124 -HP]]
: [[INSTRU]] [[]]
: [[MENTS]] [[]]
: [[1 IP]] [[71124.05 Radlat10n Monitoring Instrumentation 7111111B -]]
: [[REQUAL]] [[]]
: [[INSP]] [[]]
: [[WITH]] [[P/F]]
: [[RESULT]] [[S 3]]
: [[IP]] [[7111111 B Licensed Operator Requalification Program 71124 -EFFLUENTS]]
: [[INSPEC]] [[]]
: [[TION]] [[]]
: [[IP]] [[71124.06 Radioactive Gaseous and Uquid Effluent Treatment]]
: [[TI]] [[*177 *]]
: [[MANAGI]] [[]]
: [[NG]] [[]]
: [[GAS]] [[]]
: [[ACCUMU]] [[LATION]]
: [[IN]] [[]]
: [[ECCS]] [[2 lP 2515/177 Managing Gas Accumulation In Emergency Core Cooling. Decay Heat Removal & Containment Spray System 11/29EXM -INITIAL]]
: [[OPERAT]] [[]]
: [[OR]] [[]]
: [[LICENS]] [[]]
: [[ING]] [[]]
: [[EXAM]] [[5 U01791]]
: [[FY]] [[11-VERMONT]]
: [[YANKEE]] [[]]
: [[INITIA]] [[L]]
: [[OPERAT]] [[]]
: [[OR]] [[]]
: [[LICENS]] [[]]
: [[ING]] [[]]
: [[EXAM]] [[U01791]]
: [[FY]] [[11-VERMONT]]
: [[YANKEE]] [[.]]
: [[INITIA]] [[L]]
: [[OPERAT]] [[]]
: [[OR]] [[]]
: [[LICENS]] [[]]
: [[ING]] [[]]
: [[EXAM]] [[]]
: [[EP]] [[]]
: [[PROGR]] [[-]]
: [[EP]] [[]]
: [[PROGRA]] [[M]]
: [[INSPEC]] [[TION]]
: [[IP]] [[7111402 Radiological Environmental Monitoring Program lP 7111403 Emergency Response Organization Augmentation Tesllng]]
: [[IP]] [[7111404 Emergency Action Level and Emergency Plan Changes]]
: [[IP]] [[7111405 Correction of Emergency Preparedness Weaknesses and Deficiencies]]
: [[IP]] [[71151-EP01 DrilVExercise Performance]]
: [[IP]] [[71151*]]
: [[EP]] [[02]]
: [[ERO]] [[Drill Participation]]
: [[IP]] [[71151*EP03 Alert & Notification System]]
: [[TRI]] [[]]
: [[FIRE]] [[*]]
: [[TRIENN]] [[]]
: [[IAL]] [[]]
: [[FIRE]] [[]]
: [[PROTEC]] [[TION]]
: [[INSPEC]] [[]]
: [[TION]] [[]]
: [[3 IP]] [[7111105T Fire Protection [TrienniaD]]
: [[IP]] [[7111105T Fire Protection 71124 -HP]]
: [[INSPEC]] [[]]
: [[TION]] [[-ACCESS &]]
: [[ALARA]] [[]]
: [[IP]] [[71124.01 Radiological Hazard Assessment and Exposure Controls]]
: [[IP]] [[71124.02 Occupational]]
: [[ALARA]] [[Planning and Controls *hlsreport does not Include]]
: [[INPO]] [[and]]
OUTAGE This report shows only on-site and announced inspection
Page 2 of2 Vennont Yankee 03/0212010 15:46:26 Inspection I Activity Plan Report 22 01/01/2010 -06/30/2011 Planned Dates Start End Title 10/18/2010 10/18/2010 10/1812010 10/1812010 04/04/2011 04/1812011 0510212011 05/02/2011 0510212011 05/0212011 0510212011 05/23/2011 05/23/2011 05/23/2011 06/0812011 06/20/2011 0612712011 10/2212010 1012212010 10/2212010 10/2212010 0410812011 0412212011 0510612011 05/0612011 05106/2011 05/06/2011 05/0612011 05/2712011
05/2712011 0512712011 06/10/2011 06/24/2011 07/01/2011 71124 -RW
: [[TRANSP]] [[]]
: [[ORTATI]] [[ON]]
: [[IISFSI]] [[&]]
: [[PI]] [[]]
: [[IP]] [[60855 Operation Of An]]
: [[ISFSI]] [[]]
: [[IP]] [[71124.08 Radioactive Solid Wasta Processing and Radioactive Material Handling. Storage. and Transportation]]
: [[IP]] [[71151-OR01 Occupational Exposure Control Effectiveness]]
: [[IP]] [[71151-]]
: [[PR]] [[01]]
: [[RETS]] [[/]]
: [[ODCM]] [[Radiological Effluent 71152B -PI&R]]
: [[4 IP]] [[711528 Identification and Resolution of]]
: [[IP]] [[711528 Identification and Resolution of]]
: [[EP]] [[]]
: [[EX]] [[-EP]]
: [[EXERCI]] [[]]
: [[SE]] [[]]
: [[EVALUA]] [[]]
: [[TION]] [[]]
: [[4 IP]] [[7111401 Exercise Evaluation]]
: [[IP]] [[7111404 Emergency Action Level and Emergency Plan Changes]]
: [[IP]] [[71151-]]
: [[EP]] [[01 Drili/Exerclse Performance rp 71151-EP02]]
: [[ERO]] [[Drill Participation]]
: [[IP]] [[71151-EP03 Alert & Notification System 71124 -HP-OOSE I]]
: [[REMP]] [[]]
: [[IISFSI]] [[]]
: [[IP]] [[60!l55 Operation Of An]]
: [[ISFSI]] [[]]
: [[IP]] [[71124.04 Occupational Dose Assessment lP 71124.07 Publ1c Radiation Safety 7111121 -]]
: [[COB]] [[l]]
: [[6 IP]] [[7111121 Component DeSign Bases Inspection]]
: [[IP]] [[7111121 Component Design Bases Inspection]]
: [[IP]] [[7111121 Component Design Bases Inspection his report does not include r]]
: [[NPO]] [[and]]
: [[OUTA]] [[]]
: [[GE]] [[This report shows only on-site and announced inspection]]
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Revision as of 03:18, 24 August 2018

IR 05000247-09-005, on 10/01/2009 - 12/31/2009, Indian Point Nuclear Generating (Indian Point) Unit 2; Integrated Inspection Report and Notice of Violation
ML100400177
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 02/09/2010
From: Gray M K
Reactor Projects Branch 2
To: Pollock J E
Entergy Nuclear Operations
References
EA-09-296 IR-09-005
Download: ML100400177 (46)
v • d • e


Text

UNITED NUCLEAR REGULATORY REGION 475 ALLENDALE KING OF PRUSSIA, PENNSYLVANIA February 9, 2010 EA-09-296 Mr. Joseph Site Vice President Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB Buchanan, NY 10511-0249 INDIAN POINT NUCLEAR GENERATING UNIT 2 -NRC INTEGRATED INSPECTION REPORT 05000247/2009005 and NOTICE OF VIOLATION (EA-09-296)

Dear Mr. Pollock:

On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report documents the inspection results, which were discussed on January 21, 2010 with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's 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 NRC Enforcement Policy included on the NRC's Web site at www.nrc.gov; select About NRC, How We Regulate, Enforcement, and 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. During the inspection, the NRC identified a violation involving Entergy's submittal of inaccurate information to the NRC related to the medical qualifications of licensed operators.

Letters to the NRC certified that the operators had been medically examined and had met all medical qualifications, when, in fact, one test (namely, a tactile test) had not been performed.

A tactile test is required to ensure that operators can distinguish among various shapes of control knobs and handles by touch. The test was not performed because your Medical Review Officer {MRO) was unaware that such a test was required.

Further, the MRO considered his review of the operators'

medical history records for neurological conditions to be sufficient to verify the operators'

ability to feel, manipulate, and distinguish plant components when needed. Violations involving the provision of inaccurate or incomplete information are of particular concern to the NRC, and may be considered for escalated enforcement under certain circumstances.

However, in this case, the NRC has classified this violation at Severity Level IV, after considering the guidance set forth in Section IV.A.3 of the Enforcement Policy because the inaccurate information did not invalidate the NRC licensing since all of the operators subsequently passed a tactile test when Entergy administered it shortly after the NRC identified the violation.

Further, the actual and potential safety significance of this violation was very low in that the Medical Review Officer had conducted a neurological evaluation, albeit not a tactile test. and the operators had been observed successfully manipulating control knobs and handles by Entergy and NRC personnel in the conduct of their licensed duties. Nonetheless, this violation demonstrates the importance of taking all of the necessary steps and conducting all of the necessary reviews to assure that information submitted to the NRC is complete and accurate in all material respects.

Although this Violation has been placed in your corrective action program. a Notice of Violation is being issued and a response is being required to better understand:

1) what actions were taken in 2004 in response to NRC Information Notice (IN) 2004-20, "Recent Issues Associated with NRC Medical Requirements for licensed Operators,fl which. in part, reminded facility licensees that licensed operators and the personnel who perform and interpret their medical examinations need to be familiar with the regulatory requirements and guidelines (it should be noted that this IN specifically described an instance in which a facility licensee had not conducted some tests required in the ANSI standard for any of its licensed operators);

2) why appropriate action was not taken in response to IN 2004-20 to identify appropriate tactile testing was being conducted; and 3) the corrective actions taken and planned at this time to assure all information submitted to the NRC is complete and accurate in all material respects.

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.

Based on the results of this inspection, this report also documents three additional findings of very low safety significance.

All of these findings were determined to be violations of NRC requirements.

However, because of their very low safety significance, and because the findings were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report. you should provide a written response within 30 days of the date of this inspection report. with the basis for your denial. to the Nuclear Regulatory CommiSSion, AnN.: Document Control Desk, Washington D.C.

with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory CommiSSion, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2. In addition, if you disagree with the characterization of any finding, you should provide a response within 30 days of the date of this inspection report, with the basis for your dlsagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point Nuclear Generating Unit 2. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS). accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html.

To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction. ! * ! I

Sincerely.

Projects Branch 2 Division of Reactor Projects Docket No.

License No. DPR-26 Enclosure 1: Notice of Violation Enclosure 2: Inspection Report No. 05000247/2009005

w/Attachment:

Supplemental Information cc w/enc!: Distribution via ListServ I nATF I OFFICE NAME DATE J. Poliack 3 In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html.

To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.

Distribution w/encl: (via Ewmail) S. Collins, RA (R10RAMAIL RESOURCE)

M. Oapas, ORA (R10RAMAIL RESOURCE)

O.lew,ORP (R1DRPMAIL RESOURCE)

J. Clifford, DRP (R1DRPMAIL RESOURCE)

D. Roberts, DRS (R1DRSMaii Resource)

P. Wilson, DRS (R1DRSMaii Resource)

L. Trocine, RI OEDO M. Gray, DRP B. Bickett, DRP S. McCarver, DRP SUNSI Review Complete:

__

__

Sincerely,IRA! Mel Gray, Chief Projects Branch 2 Division of Reactor Projects M. Osborn, DRP E. Keighley, ORP K. Mangan, DRP, Acting SRI A. Ayegbusi, DRP t RI D. Hochmuth, DRP RidsNrrPMlndianPoint Resource RidsNrrDorlLpl1-1 Resource ROPreport Resource@nrc.gov (Reviewer'S Initials)

ML 100400177 DOCUMENT NAME: G:\DRP\BRANCH2\a -Indian Point 2\lnspection Reports\IP2 IR2009-005\IP2 2009 005rev3.doc After declaring this document "An Official Agency Record" it will be released to the Public. To recelvQ a copy of thii document, indicate In the box: 'C'" Copy without attachmentlenclosure

'E'" Copy with attachmentlenclosure

'N':; No copy RI/DRP RI/DRS I RI/DRP I RllEnf I I RI/RC I KManQanlKM BBicketU BB MMcLaughlinJ MMM MGray/MG KFarrar/KF 021 en 110 02/03 f10 02/04 110 0210R 110 02104/10 RIIDRP I I I I I DLew/DL 02/08/10 1 NOTICE OF VIOL ATION Entergy Nuclear Operations, Inc. Indian Point Unit 2 and Unit 3 Delcket No. 50-247 &50-286 License Nos. DPR-26 and DPR-64 EA-09-296 During an NRC inspection conducted from October 19 through October 22, 2009, a violation of NRC requirements was identified.

In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR 50.9 requires, in part, that information provided to the Commission by an applicant for a license or by a licensee or information r l 9quired by statute or by the Commission's regulations, Orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

10 CFR 55.21 requires, in part, that an applicant for a license shall have a medical examination by a physician and the licensee shall have a medical examination by a physician every two years. The physician shall determine that the applicant or lic.:msee meets requirements of Section 55.33(a)(1}.

10 CFR 55.33(a)(1}

requires, in part, that an applicant's medical condition and general health will not adversely affect the performance of aSSigned operator job duties or cause operational errors endangering public health and safety. 10 CFR 55.23 requires, in part, that to certify the medical fitness of the applicant, an authorized representative of the facility licensee shall complete and sign NRC Form-396, "Certification of Medica! Examination by Facility Licensee." NRC Form*396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant and that the guidance contained in American National Standards Institute/American Nuclear Society (ANSIIANS)

Standard 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants" was followed in conducting the examination and making the determination of medical qualification.

ANSIIANS 3.4-1983, Section 5.4 provides specific minimum capacities required for medical qualifications.

Section 5.14 requires, "Tactile discrimination sufficient to distinguish among various shapes of control knobs and handles by touch." Contrary to the above, from October 20,2004 through October 22, 2009, Entergy Nuclear Operations, Inc. (Entergy)

provided information to the NRC that was not complete and accurate in all material respects.

Specifically.

Entergy had not completed medical examinations of licensed operators in accordance with ANSI/ANS 3.4-1983.

The licensee submitted numerous NRC Form-396s for renewal of senior reactor operator and reactor operator licenses and for initial license applicants that certified that the applicants met the medical requirements of ANSI/ANS 3.4-1983 when, in fact" tactile testing had not been conducted.

This is a Severity Level IV violation (Supplement VII) . . Pursuant to the provisions of 10 CFR 2.201. Entergy Nuclear Operations.

Inc. is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional EnClosure 1

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; EA-09-296" 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, 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 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 NRC's document system (ADAMS). accessible from the NRC Web site at http://www.nrc.gov/reading-rmiadams.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 protectetd and a redacted copy of your response that deletes such information, If you request withholding of such material, you !!llJ.§! 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 9 th day of February 2010. Enclosure 1

Docket No.: License No.: Report No.: Licensee:

Facility:

Location:

Dates: Inspectors:

Approved By: 1 U.S. NUCLEAR REGULATORY COMMISSION REGION I 50-247 DPR-26 05000247/2009005 Entergy Nuclear Northeast (Entergy)

Indian Point Nuclear Generating Unit 2 450 Broadway, GSB Buchanan, NY 10511-0249 October 1, 2009 through December 31, 2009 G. Malone, Senior Resident Inspector

-Indian Point 2 O. Ayegbusi, Resident Inspector

-Indian Point 2 P. Cataldo, Senior Resident Inspector

-Indian Point 3 J. D'Antonio, Senior Operations Engineer S. Barr. Senior Emergency Prep Inspector J. Commiskey.

Health Physicist C. Crisden, Emergency Preparedness Specialist T. Fish. Senior Operations Engineer J. Lilliendahl, Reactor linspector K. Mangan, Senior Reactor Inspector J. Nicholson, Health Physicist J. Schoppy, Senior Reactor Inspector Mel Gray, Chief Projects Branch 2 Division of Reactor Projects Enclosure 2

2

SUMMARY OF FINDINGS

IR 05000247/2009005; 10/01/2009

-12131/2009;

Indian Point Nuclear Unit 2; Licensed Operator Requalification Program; Alert and Notification System (ANS) Evaluation;

Event Follow-Up; and Other Activities.

This report covered a three-month period of inspection by resident and region based inspectors.

Four finding of very low significance (Green) were identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process." The cross-cutting aspects for the findings were determined using IMC 0305, "Operating Reactor Assessment Program." Findings for which the significance determination process (SOP) does not apply may be Green, or be assigned a severity level (SL) after NRC management review. The NRC's program for overseeing safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating

Systems SL IV. An NRC-identified Severity Level IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators.

The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by tOUCh) from the required licensed operator medical examinations.

The inspectors determined that Entergy's medical physician did not adequately test all licensed operators (both initial and renewal licensees)in accordance with 10 CFR 55.21 and 10 CFR 55.33 with respect to ANSlIANS-3.41983.

However, Entergy had submitted medical information, as required by 10 CFR 55 for licensed operators and applicants that stated the testing had been performed satisfactorily.

Following identification of the issue, Entergy entered the issue into the corrective action program (CR-IP3-2009-04487)and completed corrective actions to develop and administer an appropriate test. The inspectors noted that all licensed operators passed this new test and no new license conditions were required.

Entergy's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9. Because this violation of 10 CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dis positioned using the traditional enforcement process. The finding was more than minor because documents which provided the information to the NRC were signed under oath by the company medical physician and the Site Vice President.

The applicability of cross-cutting aspects related to the performance deficiency of this finding will be determined after NRC review of Entergy's to the Notice of Violation. (Section 1 R11.2)

Green.

A self-revealing non-cited violation (NCV) of very low safety significance of 10 CFR 50, Appendix B Criterion V "Instructions, Procedures, and Drawings," was identified because Entergy personnel did not perform work regarding replacement of a control room digital recorder.

As a result, during performance of the work, personnel inadvertently shorted a live wire resulting in a partial loss of control room indications and alarms related to the safety relief valve acoustic monitor flow indications, low range steam and feed flow indications, and inadvertent control rod movement.

Entergy personnel reset the breakers to restore control room indications and entered this issue into the corrective action program as 04860. Personnel subsequently replaced the digital recorder with the circuit breaker opened to eliminate the electrical hazard. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiatino events to prevent undesirable consequences.

Specifically, the grounded recorder power supply resulted in a loss of control room indications and alarms that could have impacted operations response to an event. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,>>

and determined it to be of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance related to work practices.

Specifically, Entergy personnel did not follow procedures during the replacement of a control room digital recorder.

H.4(b) per IMC 0305] (Section 40A3.2)

Cornerstone: Emergency

Preparedness

Green.

A self-revealing NCV of very low safety significance of 10 CFR 50.47(b}(5}

was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be degraded for 56 days. The siren technicians did not use a detailed written procedure or work instruction to perform siren file updates, but instead relied on performing the task from memory. As a result, on September 16, 2009, Entergy conducted a full volume siren test durinu which a total of 18 sirens indicated a failure to function.

Entergy entered the siren failures into their corrective action process for resolution and performed a root cause of the event to determine the short and long term corrective actions. The finding was more than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of facilities and equipment, and impacted the cornerstone objective of ensuring that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.

This finding was evaluated using IMC 0609 Appendix B, "Emergency Preparedness Significance Determination Process (SOP)" and was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by siren technicians

H.4(c) per IMC 0305J (Section 1EP2)

Other Findings

SL IV, An NRC-identified Severity Level IV, NCV of 10 CFR 72.212(b)(2)(ii), was identified because Entergy personnel did not evaluate a change to the written evaluation described in its Holtec Updated Final Safety Analysis Report (UFSAR) prior to implementing the change. Specifically, inspectors identified that Entergy personnel were storing combustible material on the Independent Spent Fuel Storage Installation (ISFSI) pad which was contrary to the Holtec UFSAR and the Entergy 72.212 Evaluation Report which stated that transient combustibles will not be stored on the ISFSI pad. Following the inspectors'

questions, Entergy personnel determined the required evaluation in accordance with the requirements of 10 CFR 72.48(c) was not performed.

Entergy personnel entered the issue into their corrective action program and verified that all combustibles had been removed from the pad. The Reactor Oversight Process (ROP) was not used for this finding because inspections of ISFSI activities are covered under NRC Manual Chapter 2690 and are not incorporated in the reactor safety cornerstones in the ROP's Significance Determination Process (SOP). It was determined that the failure to evaluate a change to the written evaluation required by 10 CFR 72.212 using the requirements of 10 CFR 72.48{c) was a performance deficiency that was reasonably within Entergy's ability to foresee and prevent The finding was determined to be a Severity Level IV violation based on Supplement VI, Example 0.2 of the NRC Enforcement Policy. A cross-cutting aspect was not assigned since the performance deficiency was applicable to evaluation in accordance with the ROP. (Section

REPORT DETAILS

Summary of Plant Status Indian Point Unit 2 began the inspection period operating at full reactor power (100%). On November 2, Unit 2 shutdown due to an automatic reactor trip due to a turbine-generator protective trip resulting from a loss of the generator exciter power supply. On November "1, operators returned the plant to 100% power. Unit 2 remained at or near full power during the remainder of the inspection period.

REACTOR SAFETY

Cornerstones:

Initiating Events, Mitigating Systems, and Barrier Integrity 1 R01 Adverse Weather Protection (71111.01 -1 sample)

.1 Station Readiness

for Extreme Cold Conditions Inspection Scope The inspectors reviewed the readiness of risk-significant systems for winter cold weather conditions.

The inspectors reviewed Entergy's adverse weather procedures, operating experience.

corrective action program, UFSAR, Technical Specifications (TS). operating procedures.

and applicable plant documents to determine the types of adverse weather challenges to which the site is susceptible.

The inspectors also checked local area temperatures.

as well as the operability of ventilation and heating systems. to ensure the plant was prepared for cold weather conditions.

In addition, the following risk-significant systems that were required to be protected from adverse weather conditions were selected and collectively represented one inspection sample: Motor driven and turbine driven auxiliary feedwater system; Diesel generator fire pump; and 21, 22 and 23 emergency diesel generators (EDGs).

b. Findings

No findings of significance were identified.

1R04 Eguipment

Alignment (71111.04Q -3 samples)

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns to verify the operability of redundant or diverse trains and components during periods of system train unavailability or following periods of maintenance.

The inspectors referenced system procedures.

UFSAR. and system drawings to verify the alignment of the available train suppolted its required safety functions.

The inspectors also reviewed applicable condition reports Enclosure

.1 (CRs) and work orders to ensure Entergy personnel

identified and properly addressed equipment discrepancies that could potentially impair the capability of the available train, as required by 10 CFR 50, Appendix B, Criterion XVI, Action." The documents reviewed during these inspections are fisted in the Attachment.

The inspectors performed a partial walkdown on the following systems, which represented three inspection samples: 22 EDG after planned outage; 22 residual heat removal (RHR) train when 21 RHR pump was out of service; and EDG fuel oil system following testing.

b. Findings

No findings of significance were identified . . Full System Walkdown (71111.04S

-1 sample)

a. Inspection Scope

The inspectors performed a complete system walkdown of accessible portions of the component cooling water (CCW) system to identify discrepancies between the existing equipment lineup and the required lineup. The inspectors reviewed operating procedures, surveillance tests, piping and instrumentation drawings, equipment lineup check-off lists, and the UFSAR to verify the system was aligned to perform its required safety functions.

The inspectors reviewed a sample of eRs written to address deficiencies associated with the system to ensure they were appropriately evaluated and resolved.

The documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified. Fire Protection (71111.050 -5 samples) Resident Inspector Quarterly Walkdowns

a. Inspection Scope

The inspectors conducted tours of several fire areas to assess the material condition and operational status of fire protection features.

The inspectors verified, consistent with the applicable administrative procedures, that: combustibles and ignition sources were adequately controlled; passive fire barriers, manual firE!.fighting equipment, and suppression and detection equipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program. The inspectors evaluated the fire protection program for conformance with the requirements of License Condition 2.K. The documents reviewed during this inspection are listed in the Enclosure

.

This inspection represented five inspection samples for fire protection tours, and was conducted in the following areas: ISFSI pad area; Fire Zone (FZ) 25. 23 battery room; FZ 15 control room; FZ 90A, 91A spent fuel pool area; and FZ 252 cable spreading room.

b. Findings

No findings of significance were identified.

1 Heat Sink Performance (71111.07T -3 Samples)

a. Inspection Scope

Based on a plant specific risk assessment, past inspection results, and resident inspector input, the inspectors selected the following heat exchanger samples: 22 CCW heat exchanger; 23 EDG jacket water and lube oil heat exchangers; and Ultimate heat sink (UHS), which included operation of the service water system and UHS. The inspectors reviewed whether potential common cause heat sink performance problems were identified and corrected by the licensee.

The inspectors also reviewed potential macro fouling (silt, debris, etc.) issues and biotic fouling issues to verify the issues were closely examined by Entergy personnel.

In response to Generic Letter 13, "Service Water System Problems Affecting Safety-Related Equipment," Entergy committed to performing frequent periodic cleaning of essential service water heat exchangers in lieu of testing for degraded performanCE!.

To ensure adequate implementation of Generic Letter 89-13 commitments, the inspectors reviewed Entergy's inspection, cleaning, and eddy current testing methods and frequency with the responsible system engineer.

The inspectors compared surveillance test and inspection data, including as found conditions and eddy current summary sheets, to the established acceptance criteria to verify that the results were acceptable and that system heat exchanger operation was consistent with design. The inspectors reviewed heat exchanger design basis values and assumptions, plugging limit calculations, and vendor information.

to verify whether Entergy personnel incorporated the information into the heat exchanger inspection and maintenance procedures.

The inspectors walked down the intake area, portions of the service water system, including the service water pump and strainer pits, CCW heat exchangers, and EDG heat exchangers, to assess the material condition and operational functioning of these systems and components.

The inspectors reviewed a sample of condition reports related to the service water system to ensure that station personnel were appropriately identifying, characterizing, and correcting problems related to these systems and components.

The documents reviewed during this inspection are listed in the Attachment.

.1 9

b. Findings

No findings of significance were identified.

1Licensed Operator Requalmcation Program (71111.11 Q -1 sample) Quarterly Review

a. Inspection Scope

On October 6, the inspectors observed licensed operator simulator training.

which included simulated steam generator instrumentation failures and a large break coolant-accident (LBLOCA) coincident with the failure of several plant systems to automatically respond to adverse conditions, to verify operator performance was adequate and evaluators were identifying and documenting crew performance problems.

The inspectors evaluated the performance of risk-significant operator actions including the use of emergency operating procedures.

The assessed the clarity and effectiveness of communications, implementation of actions in response to alarms. performance of timely control board operation and manipulation, and the oversight and direction provided by the control room supervisor.

The inspectors also assessed simulator fidelity with respect to the actual plant. The inspectors evaluated licensed operator training for conformance with the requirements of 10 CFR 55, "Operator Licenses." The documents reviewed during this inspection are listed in the Attachment.

This observation of operator simUlator training represented one inspection sample.

b. Findings

No findings of significance were identified . . Licensed Operator Requalification (71111.11 B-1 sample)

a. Inspection Scope

On December 15, the inspectors reviewed results of the 2009 comprehensive written and annual operating tests to determine whether pass/fail rates were consistent with the guidance of NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process (SOP)." Inspectors verified the following: Crew failure rate on the dynamic simulator was less than (Failure rate was Individual failure rate on the dynamiC simulator test was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the walkthrough test (job performance measures)was less than or equal to 20%. (Failure rate was 0.0%); Individual failure rate on the 2009 comprehensive written exam was less than or equal to 20%. (Failure rate was 0.0%): and More than 75% of the individuals passed all portions of the exam (100% of the individuals passed all portions of the exam). Enclosure

b. Findings

Introduction:

An Severity Level IV Violation of 10 CFR 50.9, "Completeness and accuracy of information" was identified because Entergy submitted inaccurate medical information for licensed operators.

The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by touch) from the required licensed operator medical examinations.

Description:

The NRC's requirements related to the conduct and documentation of medical examinations for operators are contained in Subpart C, Medical Requirements, of 10 CFR 55, Operators' Licenses.

Specifically, 10 CFR 55.21, Medical Examination, requires every operator be examined by a physician when he or she first applies for a license and every two years, thereafter, once the license is received.

The medical examination is performed in order for the physiCian to determine whether the operator meets the requirements of 10 CFR 55.33(a)(1).

The physician is to verify that the operator's medical condition and general health will not adversely affect the performance of assigned operator duties or cause operational errors that endanger public health and safety. The facility licensee (Entergy)must also certify which industry standard (Le., the 1983 or 1996 version of ANSI/ANS-3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, or other NRC-approved method) was used in making the fitness determination.

For the medical examination performed for licensed operators at Indian Point Units 2 and 3, the inspectors determined that Entergy had stated on NRC Form 396 that the 1983 industry standard was used for the completion of the medical examination.

The inspectors noted that ANSI-3.4 1983, Paragraph 5.4.14 "Neurological," requires licensed operators to have "Tactile discrimination (Stereognosis)sufficient to distinguish among various shapes of control knobs and handles by touch." Additionally, the inspectors identified that the Form 396 was signed by both the medical review officer and Site Vice President, under oatl'i, verifying the examination had been performed.

During the medical records review, the inspectors determined that Entergypersonnel had not been conducting tactile testing of its licensed operators.

This omission had the potential for being significant since, during a transient aggravated by limited visibility, operators may be required to perform actions relying on their ability to distinguish, by touch, between different shapes of operating switches and knobs. Following identification of the issue Entergy personnel completed corrective actions to develop and administer an appropriate test. The inspectors noted that aI/licensed operators passed this new test, and no new license conditions were required.

Analysis:

The inspectors determined that a long-standing deficiency had existed at the Indian Point Units 2 and 3 in that the licensee's medical phYSician was not adequately testing all licensed operators (both initial and renewal licensees ) in accordance with 10 CFR 55.21 and 55.33 with respect to ANSI/ANS-3.4 1983. 10 CFR 55.23 requires that an authorized representative of the facility licensee shall certify the medical fitness of an applicant by completing and signing an NRC Form 396. NRC Form 396, when signed by an authorized representative of the facility licensee, certifies that a physician conducted a medical examination of the applicant as required in 10 CFR 55.21, and that Enclosure the guidance contained in ANSI/ANS-3.4 1983 was followed in conducting the examination and making the determination of medical qualification.

The licensee's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action by the NRC and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9 and the issue was within the licensee's ability to foresee and prevent. Because a violation of 1 (I CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dispositioned using the traditional enforCElment process. The finding was more than minor because the document which provided the information was provided to the NRC signed under oath by the company medical doctor and the site vice president.

Because there was no evidence that operators mis-operated equipment due to omitted tactile tests, the finding was determined to be of very low safety significance (SL IV). The applicability of cross-cutting aspects related to trie performance deficiency of this finding will be determined after NRC review of Entergy's response to the Notice of Violation.

Enforcement:

10 CFR 50.9 states, in part, "Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects." Contrary to this, from October 20,2004 through October 22,2009, Entergy submitted inaccurate information to the NRC on NRC Form 396 regarding the medical certification and testing of its licensed operators and initial applicants, This information was material to the NRC because the NRC relied on this certification to determine whether the applicant met the requirements to operate the controls of a nuclear power plant pursuant to 10 CFR 55. This issue has been entered into the facility corrective action program 04487) and is of very low safety significance.

The licensee implemented immediate corrective action and satisfactorily performed the required test. )"he inspectors verified the adequacy and promptness of the licensee's corrective actions for the medical issue. These corrective actions included the development of a tactile test which required operators to identify by touch various control knobs and switch shapes within a bag. The new tests were administered to all licensed operators and senior licensed operators.

All operators passed the test and no new deficiencies were identified, This violation is being treated consistent with other licensed operator medical examination findings and the NRC Enforcement Policy. (NOV 05000247/2009005*01, Incomplete Licensed Operator Medical Examinations)1 R12 Maintenance Effectiveness (71111.12Q -1 sample) a. Inspection" Scope The inspectors reviewed performance-based problems that involved structures, systems, and components (SSCs) to assess the effectiveness of maintenance activities.

When applicable, the reviews focused on: Enclosure Proper maintenance rule scoping in accordance with 10 CFR 50.65; Characterization of reliability issues; Changing system and component unavailability; 10 CFR 50.65(a)(1)and (a)(2) classification; Identifying and addressing common cause failures; Trending of system flow and temperature values; and Appropriateness of performance criteria for SSGs classified (a)(2). The inspectors also reviewed the system health report, maintenance backlogs, and maintenance rule basis document.

The inspectors evaluated maintenance effectiveness and monitoring activities against the requirements of 10 CFR 50.65. The documents reviewed during this inspection are listed in the Attachment.

The following component was reviewed and represented one inspection sample: Appendix R diesel generator coolant compatibility.

b. Findings

No findings of significance were identified. 1 R Maintenance Risk Assessments and Emergent Work Control (71111.13 -4 samples)

a. Inspection Scope

The inspectors reviewed scheduled and emergent maintenance activities to verify that the appropriate risk assessments were performed prior to removing equipment from service for maintenance or repair. The inspectors reviewed selected risk assessments to verify assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete.

When emergent work was performed, the inspectors reviewed the plant risk to ensure risk was promptly reassessed and managed. Documents reviewed during this inspection are listed in the Attachment.

The following activities represented four inspection samples: Emergent maintenance associated with the Appendix R diesel generator concurrent with maintenance on 138kV line 33332 L&M. power range nuclear instrumentation recalibrations, and preventative maintenance on the 22 containment spray pump on October 2; Planned maintenance associated with the 23 CCW pump and preventative maintenance on the 21 and 22 safety injection (SI) and RHR pump motor breakers on October 20; Planned maintenance associated with the 33332 L&M line, 23 CCW pump and 22 SW pump following an inadvertent trip of the 22 EDG output breaker on October 26; and Unplanned maintenance outage associated with the 22 EDG on NovembEtr 9 and 10.

b. Findings

No findings of significance were identified.

.1 1 R15 Operability

Evaluations (71111.15 -2 samples) Resident Quarterly Review

a. Inspection Scope

The inspectors reviewed operability evaluations to assess the acceptability of the evaluations, the use and control of compensatory measures, when applicable, and compliance with Technical Specifications (TS). The inspectors' reviews included verification that operability determinations were performed in accordance with procedure 04, "Operability Determinations." The inspectors assessed the technical adequacy of the evaluations to ensure consistency with the TS, UFSAR, and associated design basis documents.

The documents reviewed are listed in the Attachment.

The following operability evaluations were reviewed and represented two inspecti<m samples:

  • 21 EDG day tank level Indication; and
  • 22 EDG jacket water heater breaker failure.

b. Findings

No findings of significance were identified. \ 1 R19 Post-Maintenance Testing (71111.19 -8 samples)

a. Inspection Scope

The inspectors reviewed post-maintenance test procedures and associated testing activities for selected risk-significant mitigating systems, and assessed whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel.

The Inspectors verified that: test acceptance criteria were clear and the test demonstrated operational readiness consistent with design basis documentation; test instrumentation had current calibrations with the appropriate range and accuracy for the application; and the tests were performed as written, with applicable prerequisites satisfied.

Upon completion of the tests, the inspectors reviewed whether equipment was returned to the proper alignment necessary to perform its safety function.

Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix B, Criterion XI, "Test Control." The documents reviewed are listed in the Attachment.

The following post-maintenance testing activities were reviewed and represented eight inspection samples:

  • 23 CCW pump overhaul;
  • Motor operated valve (MOV) SI-18526 motor and actuator overhaul;
  • Starting air system maintenance and output breaker inspection on the 22 EDG;
  • Cable pull and repair splicing of the L&M 33332 line;
  • Internal inspection of EDG 23 heat exchangers; and
  • Replacement of diesel fire pump reHef valve. Enclosure

b. Findings

No findings of significance were identified.

1 Surveillance Testing (71111.22

-6 samples)

a. Inspection Scope

The inspectors observed performance of portions of surveillance tests and/or reviewed test data for selected risk-significant structures, systems, and components (SSCs) to assess whether tests satisfied TS, UFSAR, Technical Requirements Manual, and Entergy procedure requirements.

The inspectors verified that: test acceptance criteria were clear, demonstrated operational readiness, and were consistent with design basis documentation; test instrumentation had accurate calibration, and appropriate range and accuracy for the application; and tests were performed as written, with applicable prerequisites satisfied.'

Following the tests, the inspectors verified that the equipment was capable of performing the required safety functions.

The inspectors evaluated the surveillance tests against the requirements in TS. The documents reviewed during this inspection are listed in the Attachment.

The following surveillance tests were reviewed and represented six inspection samples: Feedwater valves FCV-405 A-D test (1ST); 23 EDG load test; 21 SI pump 1ST; 21 RHR pump 1ST; Condensate storage tank guided wave evaluations of underground portions of the condensate and SW piping; and 23 station battery quarterly surveillance.

b. Findings

No findings of significance were identified.

Cornerstone:

Emergency Preparedness (EP) Alert and Notification System fANS) Evaluation (71114.02 -1 sample)

a. Inspection Scope

An onsite review was conducted to assess the maintenance and testing of Indian Point Energy Center's (lPEC) current ANS. During the inspection, the Inspector interviewed the Entergy staff responsible for overseeing the ANS testing and maintenance of the system. The inspector reviewed ANS procedures and the ANS design report to ensure Entergy's compliance with design report commitments for system maintenance and testing. The inspector reviewed CRs pertaining to the ANS for causes, trends, and corrective actions. The inspector also reviewed Entergy's root cause report related to siren test results conducted in September 2009. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment

2. Planning Standard, 10 CFR 50.47(b)(5)

and the related requirements of 10 CFR 50, Appendix E, were used Enclosure as reference criteria.

The documents reviewed during this inspection are listed in the Attachment.

b. Findings

Introduction:

A self-revealing NCVof very low safety significance (Green) of 10 CFR 50.47(b)(5)was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be in a degraded condition for 56 days.

Description:

The new ANS siren system is comprised of 172 sirens located throughout the four counties within the 10 mile Emergency Planning Zone (EPZ). Of the 172 sirens, 13 are capable of voice reproduction.

The voice enabled sirens are located in areas, such as Harriman State and Croton Point Parks, where the population may not have access to media that would transmit Emergency Alert Messages.

The inspector's review of Entergy's root cause evaluations determined that, in July 2009, Entergy received new voice chips along with two data files (one for voice and one for non-voice sirens) along with instructions for installation of the chips and data files from the siren system vendor. The new voice chips and software provided an upgrade to the previous voice message. On July 15, 2009, Entergy personnel discussed the task of installing new voice chips on the digital message boards (OMS) for the 13 voice enabled sirens and installing the updated voice data file for each siren. The first voice chip installation and data file update was performed on July 20, 2009. Although the siren system vendor provided the installation instructions for the data file, the instructions were not included in the Entergy work instructions nor were they provided to the technician performing the upgrade. On July 22, 2009, technicians continued to update all voice sirens with the new voice chip and the new data file. While updating a single voice siren data file, the UPDATE ALL command was inadvertently invoked three times within a short period of time. The technician recognized the error and proceeded to abort the process all three times. A similar data file update error had previously occurred on July 20, 2009. While actions were taken to recover from the error, a CR was not and no actions were taken to prevent reoccurrence.

Between July 22 and July 29, 2009, the technicians continued to update the remaining voice sirens with the new voice chips and data file with no additional instances of the UPDATE ALL command being invoked. The installation of voice chips and the voice data files was completed on July 29, 2009. All voice sirens were updated and verified with the voice chips and the new data file. The post maintenance testing for this activity would not have identified the latent error with the non-voice enabled sirens because it was not intended to have modified these sirens during this work activity.

As a result of the data file update error on July 22, 2009. 14 non-voice sirens werH configured as voice sirens. After the technician made the file update error on July 22,2009, the technician did not verify that the correct data files were installed for all non-voice sirens (three non-voice sirens were verified as having the correct files after the July 20, 2009 data update error). This error caused 14 non-voice sirens to be left in Enclosure 16 a condition where the sirens would function (annunciate);

however, the indication at the siren activation points would indicate that the sirens had failed (red-dots versus dot for successful activation).

In August 2009, routine polling, silent tests and annual Preventive Maintenance (PM) were conducted by Entergy. The annual PM procedure requires verification if the individual siren's data file is correct for the type of siren (voice or non-voice).

During the PMs, several siren data files were found to be incorrect and were corrected during the PM. The last four PMs conducted on non-voice sirens in the August/September timeframe each began with a non-voice siren verification failure. This failure was :an indication that the non-voice siren was configured with a voice siren data file. The Entergy Root Cause report determined that the failure should have been identified by the technician and indicated that there was a more significant problem with the siren data files. This problem was neither documented in a CR nor was it reported to management.

The silent tests that were conducted would not have identified voice data file configuration errors. On September 16, 2009, Entergy conducted a full volume test of the siren system. Of the 172 sirens activated during the test, 18 siren failures were observed (red dots on displays indicating siren failures).

Of the 18 failures observed, four were reported as amplifier (AMP) failures and 14 were reported as DMS errors. The inspector did not identify a performance deficiency associated with the four AMP siren failures.

The 14 DMS errors were due to an incorrect data file being installed for the siren. The sirens indicating an error were non-voice sirens that were installed with the voice data file. According to procedure IP-EP-AD30, IPEG ATI Siren System Administration, maintenance on the siren system will be performed using procedures, step lists, and checklists per IP-EP-AD31, IPEC Siren System Maintenance Administration Procedure.

IP-EP-AD31 states checklist and procedures will be used if the work is beyond the skill of the craft or the vendor tech manuals. Contrary to IP-EP-AD30, the inspectors determined the technician did not use detailed written procedures nor work instructions to perform the siren updates. Instead the technician relied on performing the from memory. As a result, on September 16, 2009, 14 DMB failures occurred due to an incorrect data file being installed for the sirens. Troubleshooting testing conducted following the September 16, 2009. full volume test, demonstrated that while the 14 sirens indicated that they had failed to function, the sirens most likely sounded based on this subsequent testing. In the case of a siren indicating failure during an actual event, Entergy would use an installed reverse calling system to notify the affected public. Following the siren test failures, Entergy diagnosed the data file error, installed the correct data file, and had all 14 sirens returned to an operable status on the day of the test. On October 22, 2009, a subsequent full volume test demonstrated 100 percent successful siren activation.

Analysis:

The inspector determined that Entergy's failure to use procedures, step lists or checklists while performing maintenance on the siren system was a performance deficiency resulting in approximately 8% of the system to be degraded for 56 days. The finding is greater than minor because it is associated with the emergency preparedness (EP) cornerstone attribute of Facilities and Equipment (Maintenance of Equipment)and affected the EP cornerstone objective of ensuring the capability to implement adequate measures to protect the health and safety of the public in the event of a radiological emergency.

This finding was evaluated using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Sheet 1, "Failure to Comply." The finding is associated with the failure to meet or implement a regulatory requirement (planning standard).

The finding is not more than Green because it did not result in a Risk Significant Planning Standard (RSPS) function being lost or degraded.

The SOP defines degradation of this RSPS to be, "the public alert and notification system (e.g., sirens, other supporting primary notification methods) has design flaws or deficiencies in the test program, maintenance program, or procedures that degrade a portion of the system for a significant period from the time of discovery (e.g., 100% over 25 days, greater than 48% over 45 days, greater than 24% over'90 days, greater than 12% over 6 months)." In this case, approximately 8% of sirens were degraded for over 45 days; therefore, it was concluded that the RSPS was not dewaded (as defined by the SOP) and the finding was determined to be of very low safety significance (Green). This finding has a aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by station personnel and siren technicians (H,4(c>>).

Enforcement:

1 0 CFR 50.54(q) states in part that the facility licensee shall follow and maintain in effect emergency plans which meet the standards in 50,47(b) and the requirements in Appendix E of this part. Planning Standard 10 CFR 50,47 (b)(5) requires, in part, that a means to provide early notification and clear instruction to the populace within the plume exposure pathway EPZ have been established.

Contrary to the above, from July 22, 2009 until September 16, 2009, a means to provide early notification and clear instruction to the populace within the plume exposure pathway EPZ had not been established in the areas adjacent to the 14 non-functional sirens. A contributing cause for this violation was the failure to use procedures, step lists or checklists during a siren maintenance actiVity conducted on July 22, 2009. Because this violation was of very low safety significance and it was entered into Entergy's corrective action program (CR-IP2-2009-3701);

this violation is being treated as an NCV, consistent with Section VLA.1 of the NRC Enforcement Policy. (NCV 05000247/2009005*02, Siren Test Failure) 1 Emergency Response Organization (ERO) Staffing and Augmentation System (71114.03 -1 sample) Inspection Scol2e The inspector conducted a review of IPEC's ERO augmentation staffing requirements and the process for notifying and augmenting the ERO. This was performed to ensure the readiness of key licensee staff to respond to an emergency event and to ensure Entergy's ability to activate their emergency facilities in a timely manner. The inspector reviewed the IPEC ERO roster, sampling of training records, and CRs related to the ERO staffing augmentation system. The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment

3. Planning Standard, 10 CFR 50,47(b)(2)

and related requirements of 10 CFR 50, Appendix E, were used as reference criteria.

The documents reviewed during this inspection are listed in the Attachment. Findings No findings of significance were identified.

1EP4 Emergency Action Level (EALl and Emergency Plan Changes (71114.04

-1 sample)

a. Inspection Scope

Since the last NRC inspection of this program area, Entergy implemented various changes to different sections of their emergency plan. Entergy had determined that, in accordance with 10 CFR 50.54(q), any change made to the emergency plan, and its lower-tier implementing procedures, had not resulted in any decrease in effectiveness of the plan, and that the revised plan continued to meet tl1e standards in 50.47(b) and the requirements of 10 CFR 50 Appendix E. The inspector reviewed all emergency plan changes, including the changes to lower-tier emergency plan implementing procedures, to evaluate for any potential decreases in effectiveness of the emergency plan. However, this review by the inspector was not documented in an NRC Safety Evaluation Report and does not constitute formal NRC approval of the changes. Therefore, these changes remain subject to future NRC inspection in entirety.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment

4. The requirements

in 10 CFR 50.54(q) were used as reference criteria.

The documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of Significance were identified.

1 EP5 Correction of Emergency Preparedness Weaknesses (71114.05 -1 sample)

a. Inspection Scope

The inspectors reviewed a sampling of self-assessment procedures and reports to assess Entergy's ability to evaluate their EP performance and programs.

The inspectors reviewed a sampling of CRs from December 2007 through November 2009, initiated by Entergyat IPEC from drills and audits. Additionally, the inspectors reviewed 10 CFR 50.54(t) audits; and self*assessment reports. This inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment

5. Planning Standard, 10 CFR 50.47{b)(14)

and the related requirements of 10 CFR 50. Appendix E, were used as reference criteria.

The documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

2. RADIATION

SAFETY Cornerstone:

Occupational Radiation Safety (OS) 20S1 Access Contro! to Radiologically Significant Areas (71121.01 15 samples)

a. Inspection Scope

During September 28 through October 2, 2009, the inspectors conducted activities to verify that Entergy staff at IPEC were properly implementing physical, engineering, and Enclosure 19 administrative controls for access to high radiation areas (HRAs), and other radiologically controlled areas, and that workers were adhering to these controls when working in these areas. Implementation of the access Gontrol program was reviewed against the criteria contained in 10 CFR 20, site technical speCifications, and Entergy's procedures required by the Technical Specifications as criteria for determining compliance.

During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The documents reviewed during this inspection are listed in the Attachment.

The inspectors performed independent radiation dose rate measurements and reviewed the following items: Plant Walk Downs and RWP Reviews The inspectors reviewed exposure significant work areas within radiation areas, HRAs, and airborne areas in the plant to assess licensee controls and surveys for adequacy.

Work reviewed included 3R15 Refueling Outage and On-Une work activities: U2 Rep Platform Entry (Oil AddiUon) U2 Vapor Containment, Replace 21 CRD Fan Motor radiation work permit (RWP) 2009-2033 U2 Fuel Moves, RWP 2009-2043 U2 Dry Cask Storage & Associated Work, RWP 2009-2029 Radiation protection support for locked HRA (LHRA) Entries, RWP 3501 Maintenance Support, RWP 2009-3506 Waste Management, RWP 2009-3504 Scaffolding, RWP 2009-3518 Outage Valve Work, RWP 2009*3520 Reactor Disassembly

& Reassembly, RWP 2009-3521 Split Pin Repair & Associated Work, RWP 2009-3530 RCP Pump & Motor Work, RWP 2009-3534 With a survey instrument and assistance from a Health Physics qualified individual, the inspectors walked down various areas to determine:

whether the RWP, procedure, and engineering controls were in place and whether surveys and postings were adequate.

The inspectors reviewed RWPs that provide access to exposure-significant areas of the plant. Specified electronic personal dosimeter alarm set points were reviewed by inspectors with respect to current radiological condition appllcability and workers were queried to verify their understanding of plant procedures governing alarm response and knowledge of radiological conditions in their work area. The inspectors determined there were no RWPs for airbome radioactivity areas with the potential for individual worker internal exposures of >50 millirem (mrem) committed effective dose equivalent (CEDE). Additionally, the inspectors determined there were no internal dose assessments that resulted in actual internal exposures greater than 50 mrem CEDE. Enclosure Problem Identrfication and Resolution The inspectors reviewed access control-related eRs generated since the last inspection in this area was conducted.

Staff members were interviewed and documents reviewed to determine that follow-up activities are being conducted in an effective and timely manner, commensurate with their safety and risk. For repetitive deficiencies or significant individual deficiencies in problem identification and resolution, the inspectors determined jf the licensee's assessment activities addressed the repetitive aspects. The inspectors reviewed events to determined whether there existed performance indicator occurrences that involved dose rates greater than 25 Rem/hour at 30 em, dose rates greater than 500 Rem/hour at 1 meter, unintended exposures greater than 100 mrem total effective dose equivalent (TED E), greater than 5 Rem shallow dose equivalent (SDE), or greater than 1.5 Rem lens dose equivalent (LDE). Job-in-Progress Reviews The inspectors observed aspects of various on-going activities to confirm that radiological controls, such as required surveys, area postings, job coverage, and job site preparations were conducted.

The inspectors verified that personnel dosimetry was properly worn and that workers were knowledgeable of work area conditions.

The inspectors attended briefing meetings for U2 Badger Testing and ISFSI related activities.

High Risk Significant.

High Dose Rate High Radiation Areas and Very HRA (VHR82 Controls . Key control associated with LHRA and VHRAs were reviewed by inspectors to assess Entergy's controls and inventory and to verify accessible LHRAs were properly seicured and posted during plant tours. The inspectors discussed with radiation protection supervision the adequacy of high dose rate HRA and \lHRA controls and procedures and verified that no programmatic or procedural changes have occurred that reduce the effectiveness and level of worker protection, Radiation Worker Performance During observation of the work activities listed above, the inspectors evaluated radiation worker performance with respect to the specific radiation protection work requirements and their knowledge of the radiological conditions in their work areas. The inspectors reviewed eRs related to radiation worker performance to determine if an observable pattern traceable to a similar cause was evident Radiation Protection Technician Proficienc),:

During observation of the work activities listed above, inspectors evaluated radiation protection technician work performance with respect to' their knowledge of the radiological conditions, the specific radiation protection work requirements and radiation protection procedures.

The inspectors reviewed eRs related to radiation protection technician performance to determine if an observable pattern traceable to a similar cause was evident. Enclosure

b. Findings

No findings of significance were identified 20S2 AlARA Planning and Controls (71121.02

-10 samples)

a. Inspection Scope

During September 28 through October 2, 2009, the inspectors conducted the foUowing activities to verify that Entergy staff were properly maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the AlARA program was reviewed for conformance with the criteria contained in 10 CFR 20, applicable industry standards, and Entergy's procedures.

The documents reviewed during this inspection are listed in the Attachment.

Inspection Planning The inspectors reviewed pertinent infonmation regarding cumulative exposure history, current exposure trends, and on-going activities to assess current perfonmance and outage exposure challenges.

The inspectors determined the site's 3-year rolling collective average exposure.

The inspectors reviewed work performed during the inspection period, the associated ALARA plans, RWPs, AlARA Committee Reviews, exposure estimates, actual exposures and post job reviews. Jobs reviewed included those listed earlier in this report in Section 20S1. The inspectors reviewed implementing procedures associated with maintaining occupational exposures AL.ARA. This included a review of the processes used to estimate and track work activity exposures.

Radiological Work Planning With respect to the work activities reviewed, the inspectors reviewed dose summary reports, related post-job ALARA reviews, related RWPS, exposure estimates and actual exposures, and ALARA Committee meeting paperwork.

The inspectors reviewed ALARA work activity evaluations, exposure estimates, and exposure mitigating requirements were reviewed for work packages.

The inspectors' review was to verify whether the licensee has established procedures and work controls, based on sound radiation protection principles.

The inspectors compared the results aChieved with the intended dose that was established in the planning of the work. The inspectors evaluated the basis for inconsistencies between the intended and actual work activity doses and station management awareness and involvement.

Job Site Inspections and ALARA Controls The inspectors reviewed work activities that present the highest radiological risk to workers. The inspectors evaluated the licensee's use of engineering controls to achieve dose reductions and to verify that procedures and controls are consistent with ALARA reviews. Associated ALARA Plans and RWPS were reviewed by inspectors to determine if appropriate exposure and contamination controls were being employed.

Radiation Worker Performance Through observations and interviews, the inspectors reviewed whether workers and technicians were found to be knowledgeable of the work area radiological conditions and low dose waiting areas. Declared Pregnant Workers The inspectors reviewed information associated with declared pregnant workers (juring the assessment period and whether appropriate monitoring and controls were being utilized to ensure compliance with 10 CFR 20. Problem Identification and Resolution The inspectors reviewed elements of the licensee's corrective action program related to implementing radiological controls to determine if problems are being entered into the program for timely resolution.

No findings of significance were

OTHER ACTIVITIES

40A 1 Performance Indicator Verification (71151 -8 samples) Inspection Scol2e The inspectors reviewed performance indicator (PI) data for the cornerstones listed below and used Nuclear Energy Institute

'Regulatory Assessment Performance Indicator Guideline," Revision 6, to verify individual PI accuracy and completeness, The inspectors reviewed the PI data and supporting documentation from the fourth quarter of 2008 through the third quarter of 2009 to verify the accuracy of the reported data, The documents reviewed during this inspection are listed in the Attachment.

Barrier Integritv Cornerstone

Encfosure

.1 Public Radiation

Safety Cornerstone Radiological Effluent Technical Specifications (RETS)fOffsite Dose Calculation Manual (ODCM) Radiological Effluent Occurrences.

Emergency Preparedness Cornerstone Drill and Exercise Performance (DEP); ERO Drill Participation; and ANS Reliability.

b. Findings

No findings of significance were identified. Identification and Resolution of Problems (71152 -2 sam pies) Resident Inspector Daily Review of Conditions Reports

a. Inspection Scope

As required by Inspection Procedure 71152, "Identification and Resolution of Problems," and to identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for CRs and attending CR group screening meetings.

In accordance with the baseline inspection modules, the inspectors selected corrective action program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for further follow-up and review. The inspectors assessed Entergy personnel's threshold for problem identification, adequacy of the causal analysis, extent of condition reviews, and operability determinations.

and timeliness of the associated corrective actions.

b. Findings

No findings of significance were identified . . Semi-Annual Trend Review (71152 -1 sample)

a. Inspection Scope

The inspectors performed a semi-annual review of Unit 2 issues, to identify trends that might indicate the existence of more significant safety issues. The inspectors included in this review, repetitive or closely-related issues that may have been documented by Entergy outside of the corrective action program, such as trend reports, performance indicators, major equipment problem lists. maintenance rule assessments, and maintenance or corrective action program backlogs.

The inspectors also reviewed Entergy's corrective action program database for the tl1ird and fourth quarters of 2009, to assess CRs written in various subject areas (eqUipment problems, human performance

.3 issues, etc.), as well as individual

issues identified during the NRC's daily CR review. The inspectors reviewed Entergy's quarterly trend report for the third quarter of 2009, and specific inputs from the Engineering Department that were included in the site trend report, to verify the existence or absence of, identified trends and the adequacy of existing corrective actions to address these trends. The inspectors also reviewed 121, "Entergy Trending Process," to verify that Entergy was appropriately evaluating and trending adverse conditions in accordance with applicable procedures.

The documents reviewed during this inspection are listed in the Attachment.

b. Assessment and Observations No findings of Significance were identified.

The inspectors identified several issues and events that occurred over the course of the past year, and more specifically, the third and fourth quarters of 2009, which could objectively be considered adverse trends. The inspectors verified that these issues were either addressed within the scope of the corrective action program, or through department review and documentation in the quarterly trend report for overall assessment.

For example, the inspectors reviewed the following issues: IP2-2009-04306

-Root Cause Evaluation:

Adverse Trend* Centrifugal Pump Rework; and IP2-2009*02629

  • Recent events involving weaknesses in supplemental personnel work practices; No adverse trends were identified by the inspectors that were not previously addressed by Entergy personnel.

Aggregate Impact of Operator Workarounds (71152 -1 sample)

a. Inspection Scope

The inspectors conducted a review of the aggregate impact of operator burdens and workarounds.

The inspectors reviewed Entergy's implementation of procedures 45, "Operator Burden Program." The inspectors conducted control room walkdowns and interviewed plant operators to determine the impact of defiCiencies on operator response to plant events. Additionally.

the inspectors reviewed operator logs, CRs and performed system walkdowns to verify that there were no risk significant operator actions that had not been evaluated by Entergy personnel.

b. Findings and Observations

No findings of significance were identified.

The inspectors verified that operator workarounds and burdens were entered into the corrective action program at an appropriate threshold and that corrective actions were planned or taken commensurate with their safety significance.

.1 40A3 Event Follow-Up

(71153 -2 samples) Reactor Trip on November 2, 2009. Due to a Turbine-Generator Exciter Protective Trip

a. Inspection Scope

The inspectors responded to the control room on November 2,2009, following an automatic reactor trip due to a turbine-generator protective trip resulting from a loss of the Generrex power supply. The inspectors observed Entergy's post-trip response in the control room to determine if plant equipment responded as expected, and to ensure that operating procedures were being appropriately implemented.

The inspectors attended post-trip review and forced outage meetings, and discussed the event, plant response and corrective actions with plant management.

The purpose of the reviews was to confirm that Entergy had taken appropriate actions during and foHowing the event, and had taken appropriate corrective actions for the trip prior to commencing restart activities.

The documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified . . 2 Partial Loss of Control Room Indication During NI-41 Recorder Replacement

a. Inspection Scope

The inspectors responded to the control room on November 23,2009, following notification by the shift manager that there had been a partial loss of control room annunciators and alarms associated with safety relief valve acoustic monitor indication, low range steam and feedwater flow indication, and inadvertent control rod movement Entergy personnel determined that the partial loss of c(mtrol room indications and control rod movement was due to inadvertent grounding of a live feed wire during the replacement of a control room digital recorder.

The grounding caused the recorder's associated breaker to open and the 21 instrument bus to from its normal source (static inverter)to its alternate source (transformer).

The inspectors verified that Entergy operations and maintenance personnel had taken appropriate actions following the inadvertent grounding of the wire and resultant control room indications.

The inspectors' review included verification that applicable TS limiting conditions of operation (LCO) were entered by operations personnel for the eqUipment made inoperable by the partial loss of control room indications/alarms.

Finally, the inspectors performed system walkdowns, interviewed personnel, reviewed applicable CR's, work packages, plant procedures, operating experience and corrective actions associated with the apparent cause evaluation performed by Entergy personnel to independently assess the causes of the partial loss of control room annunciators.

The documents reviewed during this inspection are listed In the Attachment.

b. Findings

Introduction:

A self-revealing NCV of very low safety significance (Green) of 10 CFR 50, Appendix B Criterion V "Instructions, Procedures, and Drawings," was identified because Entergy personnel did not perform work in accordance with instructions associated with the replacement of a control room digital recorder.

As a result, during performance of Enclosure the work, Entergy personnel shorted a live wire which resulted in a partial loss of control room indications and alarms, and inadvertent control rod movement.

Description:

On November 23, 2009, during the replacement of control room safety related digital recorder NR-41, electrical maintenance personnel inadvertently grounded the recorder's live power lead on the bracket of the recorder.

NR-41 provides control room operators indication for reactor power from the power range upper detectors.

Entergy personnel determined this resulted in the NR-41 circuit breaker opening and the power supply for the 21 instrumentation bus auto transferring from normal (static inverter)power to the alternate (transformer)power supply requiring entry into TS limiting condition of operation (LCO) 3.8.7. The opening of the circuit breaker caused a partial loss of control room annunciators retated to acoustic monitors for safety relief valves, and low range steam and feed flow indication.

In addition, operations personnel observed control rods (control bank '0') move in half a step. Entergy personnel determined that the control rod movement occurred because of the power transient associated with the 21 instrument bus transferring from its static inverter to an alternate power supply. The inspectors determined that Work Order (WO) 163807 provided instructions for replacing NR-41 and required the performance of a pre-job brief. WO 163807 idEmtified that working on live circuits was a "safety hazard" and an "error likely situation." The WO instructed maintenance personnel to "tape all areas where feed wires present. if applicable." The inspectors determined that EN-HU-102, "Human Performance Tools," requires an acceptable defense against an error likely situation and taping of all areas was identified in the WO as the human performance tool to address the error likely situation.

Entergy determined that the maintenance personnel did not apply the electrical barriers to prevent the inadvertent ground of the live power supply prior to performing the work. Following the event. inspectors observed that Entergy personnel replaced the digital recorder with the circuit breaker opened to eliminate the electrical hazard. Entergy entered the issue into the corrective action program (CR-IP2-2009-04860)and implemented corrective actions which included supplemental training for station personnel regarding the station's requirements to follow procedural direction.

Analysis:

The inspectors determined that a performance deficiency associated with this finding was that Entergy maintenance personnel did not follow instructions provided in the WO to install electrical protective barriers when working on live equipment.

The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the grounded recorder power supply resulted in a loss of control indication and alarms that would impact operations' response to an event. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A.

Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green) because it did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events. Enclosure

.1 The inspectors

determined that this finding had a cross"cutting aspect in the area of human performance related to work practices because Entergy personnel did not follow procedures during the replacement of a control room digital recorder. (H.4(b))

Enforcement:

10 CFR 50, Appendix'B Criterion V "Instructions, Procedures, and Drawings" in part, requires that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions.

Contrary to the above, on November 24,2009, maintenance personnel did not follow the instructions provided in the WO during replacement of the safety-related digital recorder NR-41. Because this issue was of very low safety significance and was entered into Entergy's correctivH action program (IP2-2009-04860).

this violation -is being treated as an NeV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000247/2009005*03, Partial Loss of Control Room Indication during NI-41 Recorder Replacement)40A5 Other Activities Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with site security procedures and regulatory requirements relating to nuclear plant security, These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.

b. Findings

No findings of significance were identified . . 2 Independent Spent Fuel Storage Installation

{60855 and 60855.1)

a. Inspection Scope

On December 14, 2009, Entergy personnel completed its dry cask loading campaign for Unit 2. The inspectors reviewed documents and records associated with the operation of the Indian Point Energy Center (IPEG) Independent Spent Fuel Storage Installation (ISFSI), including training records for personnel involved with loading dry cask canisters.

The inspectors met with reactor engineering personnel to review the fuel selection process and associated documentation.

The video recording of the fuel bundles placed into the last loaded canister was reviewed by inspectors to verify proper bundle location.

The inspectors review also included verification of the annual inventory and the location of each dry cask canisters on the ISFSI pad. The inspectors interviewed radiation protection personnel to review total dose per canister, AlARA goals, and neutron dose determinations.

The inspectors also interviewed fire protection personnel to determine the follow up to assess the adequacy of the evaluation Entergy performed for all transient combustibles to be stored on the ISFSI pad. The documents reviewed during this inspection are listed in the Attachment.

b. Findings

Introduction:

An NRC-identified NCV of very low safety significance (SLlV) of 10 CFR 72.212(b)(2)(ii), was identified because Entergy personnel did not evaluate a change to the written evaluation required by this paragraph using the requirements of 10 CFR 72.48(c), prior to storing transient combustibles on the ISFSI pad. The Holtec Final Safety Analysis Report (FSAR) and the Entergy 72.212 Evaluation Report, state that transient combustibles will not be stored on the ISFSI pad.

Description:

On September 28, 2009, inspectors questioned whether a mobile lift was allowed per procedures to be stored on the ISFSI pad adjacent to unloaded STORMs. Entergy personnel issued condition report O. Station personnel removed the mobile lift off the pad but other transient combustibles.

such as plywood and pallets, remained on the pad. During follow-up inspection related to the condition report, the inspectors determined that Entergy personnel were operatinfl under an incorrect assumption that there were active and non-active portions of the ISFsr pad, and that it was acceptable to store transient combustibles and fueled vehicles on the ISFSI pad as long as they were kept at a minimum of 30 feet from loaded casks. The inspectors determined that there was no description of active and non-active portions of the ISFSI pad in Entergy procedures relating to dry cask storage. Entergy uses the Holtee dry cask system under the Certificate of Compliance number 1014 issued to Holtec. The inspectors identified that the Holtec FSAR and the Entergy 72.212 Evaluation Report stated that there will be no combustibles stored on the ISFSI pad. The Holtee FSAR also provided design information that included a worst case fire analysis which concluded that a 50 gallon fuel tank fire (from the vertical cask transporter fuel tank) would result in only minor impact on the HI-STORM.

The inspectors reviewed Entergy's control of combustibles corporate procedure which identified the Indian Point ISFSI pad as a Level 1 combustible control zone. The procedure defines Level 1 as a fire sensitive area where transient combustible loading is prohibited unless evaluated and approved via this procedure.

In accordance with the Entergy corporate procedure, a transient combustible evaluation (TCE) should be performed prior to the combustibles being stored on the ISFSI pad. A TCE was performed by Entergy on October 19,2009, after the inspectors identified and informed Entergy personnel that combustibles were being stored on the ISFSI pad. The TCE determined that the fire hazard from the combustibles stored on the pad presented less of a fire hazard than the scenario analyzed in the Holtec FSAR. The inspectors questioned whether the TCE was appropriate considering the licensing basis documentation in the Holtec FSAR and 72.212 Evaluation Report allowed no combustibles on the pad. The inspectors determined an evaluation in accordance with the requirements of 10 CFR 72.212(b}(2}{ii}

was required to store combustibles on the ISFSI pad. Subsequent to inspector questions, Entergy personnel entered the issue into the corrective action program and all transient combustibles were removed from the pad.

Analysis:

The Reactor Oversight Process (ROP) was not used in this case because inspections of ISFSI activities are covered under NRC Manual Chapter 2690 and are not considered applicable to evaluation under a reactor safety cornerstone in the ROP's Significance Determination Process (SOP). It was determined that the failure to evaluate a change to the written evaluation required by 10 CFR 72.212 in accordance with requirements of 10 CFR 72.48(c} is a Enclosure

.3 29 performance

deficiency that was reasonably within Entergy's ability to foresee and prevent. The finding was determined to be a Severity Level IV violation based on Supplement VI, Example D.2 of the NRC Enforcement Policy.

Enforcement:

10 CFR 72.212(b )(2)(H) requires in part that a licensee shall evaluate any changes to the written evaluations required by this paragraph using the requirements of 10 CFR 72.48{ c}. Contrary to the above, prior to September 28, 2009, Entergy personnel did not evaluate changes to the written evaluation required by this paragraph.

Specifically, the Entergy's 10 CFR 72.212 evaluation report determined that a fire suppression system is not used at the IPEC ISFSI pad because there are no combustible materials stored on the ISFSI. However, between September 28, 2009 and December 17, 2009, combustibles were stored on the ISFSI pad and the licensee did not evaluate this change using the requirements of 10 CFR 72.48(c).

Because this violation was of very low safeW significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as an NCV consistent with the NRC Enforcement Policy. (NCV 05000247/2009*005*04, Transient Combustibles Stored on the ISFSI Pad) Temporary Instruction 2515/175:

Emergency Response Organization, Drill/Exercise Performance Indicator.

Program Review The inspectors performed NRC Temporary Instruction (TI) 2515/175, ensured the completeness of the licensee's completed 1 from the TI, and forwarded that data to NRC Headquarters.

40A6 Meetings.

including Exit On January 21, 2010, the inspectors presented the inspection results to Mr. Joseph Pollock and other Entergy managers and staff, who acknowledged the inspection results. Entergy staff identified documents which were to be considered proprietary and handled as such. ATTACHMENT:

=SUPPLEMENTAL

INFORMATION=

KEY POINTS OF CONTACT

Site Vice President

General Manager, Plant Operations

Assistant

General Manager, Plant Operations

Director of Nuclear Safety Assurance

Director, Engineering

Emergency

Planning Manager Site Operations

Manager System Engineering

Manager Licensing

Specialist

Training Manager Fire Protection

Licensing

Specialist

Assistant

Operations

Manager Training Instructor

Supervisor, Radiation

Protection

Support Training Instructor

Manager, Radiation

Protection

Acting Director, Nuclear Safety Assurance

LOR Program Administrator

Senior Lead Engineer Security Supervisor

Acting Manager, Corrective

Actions & Assessment

Dry Cask Superintendent

Specialist, Effluent & Environmental

Monitoring

Licensed Operator Requalification

Training Supervisor

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

05000247/2009-005-01

NOV Incomplete

Licensed Operator Medical Examinations

Opened and Closed

05000247/2009-005-02

NCV Siren Test

05000247/2009-005-03

NCV Partial Loss of Control Room Indication

during Nl 41 Recorder Replacement

05000247/2009-005-04

NCV Transient

Combustibles

Stored on the ISFSI pad Attachment

LIST OF DOCUMENTS

REVIEWED Common Documents Used Indian Point Unit 2 Updated Final Safety Analysis Report, Rev. 21 Indian Point Unit 2 Individual Plant Examination of External Events, August 1992 Indian Point Unit 2 Technical Specifications and Bases, Amendment

2 Indian Point Unit 2 Technical Requirements Manual, Rev. 9 Indian Point Unit 2 Control Room Narrative Logs Indian Point Unit 2 Plan of the Day Section 1 R01: Adverse Weather Protection

Procedures

2-S0P-30.1.
Electric Heat Tracing, Rev. 25 2-COL-11.5, Space Heating and Winterization, Rev. 28
COL 30.1, Electric Heat Tracing, Rev. 25 2-S0P-11.5, Space Heating and Winterization, Rev. 32 Condition Reports (CR-) IP2*2006-01308
IP2-2006-04980
IP2-2007 -00883
IP2-2009-00729
IP2-2007 -03368 Section 1 R04: Equipment Alignment

Procedures

2-PT-Q024C.
EDG Fuel Oil Transfer Pump, Rev. 9 2-COL-27.3.1, Diesel Generators, Rev. 25 2-COL-4.2.1.
Residual Heat Removal System, Rev. 26 2-COL-4.1.1, Component Cooling System, Rev. 22 2-S0P-4.1.2, Component Cooling System Operation.
Rev. 34 Condition Reports (CR-)
IP2-2009-05261
IP2-2009-02977
IP2-2008-02406
IP2-2008-01705
IP2-2009-03666
IP2-2008-02091
IP2*2008-02037
IP2-2008-02054

Drawings

9321-LL-3133-20, Diesel Generator Compressor Fuel, Oil Pump and Jacket Water and Lube, Oil Heaters, Sheet 4 IP2--S-000284-14.
Schematic for Diesel Generator
9321-F-2030-39, Fuel Oil to Diesel Generators
21-F-3220-23, Wiring Diagram. Diesel Generators
21-22-23.
Sheet 4 9321-F-3217-06, Wiring Diagram, Diesel Generators
21-22-23*.
Sheet 1 A227781-82, Flow Diagram Auxiliary Coolant System. Sheet 1 Miscellaneous
2-ARP-003, Diesel Generator, Low Fuel Level, Rev. 6 2-ARP-SHF, CCR Electrical, Rev. 29 2-IC-PC-I-L-1207S, Diesel Generator Fuel Oil Day Tank No. 21 Level. Rev. 3 Maintenance Rule BasiS Document Component Cooling Water (CCW). Rev. 02 Attachment Section 1 R05: Fire Protection

Procedures

2-S0P-27.1.6, Instrument Bus DC Distribution System, Rev. 39 Condition Reports (CR-)
IP2-2009-04233
IP2-2009-05007*

Miscellaneous

IP2-RPT-03-00015, IP2 Fire Hazards Analysis, Rev. 2
PFP-253, (PEC Pre-Fire Plans, Rev. 0
PGI-00433, Combustible Loading Calculation, Rev. 4 Pre Fire Plan
PFP-252, Cable Spreading Room -Control Building, Rev. 0 Section 1 R07: Heat Sink Performance

Calculations

FFX-00361-00, Minimum Wall Thickness

Calculations

for Tubes of Jacket Water Cooler and Lube Oil Cooler for EDG, Rev. 0 .

FMX-00295-00, Tube Plugging Limits for EDG LO and JW Cocllers, Rev. 0
PGI-00087-00, EDG LO Cooler Sizing, Rev. 0
PGI-00387-00, Testing of the CCW Heat Exchangers at Power Operation, Rev. 0 Vendor Manual 755C, Instruction Book for ConEd Component Cooling HXS, Rev. 0 Test Results 2-HTX-004-CCW, Component Cooling Water Heat Exchanger Maintenance, dated 2/23/05. 12112/06, and 2/17/09 O-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, dated 1/8107. 7/16/08, 1/4109, and
1217109 Eddy Current I nspection Reports for 22 CCW HX, dated 11/15/02,
2115105, 12/13/06, and
2110109 Eddy Current Inspection Reports for 23 EDG Jacket Water and Lube Oil HXs, dated 11/6/02, and 12/7109 Modifications
EC10675, Timed Operation of the Zurn Strainer Circuitry, Rev. 4 EC12566, Material Upgrade of Service Water Strainer Slowdown System, Rev. 5 Conditions Reports (CR-)
IP2-2004-05064
IP2-2006-03916
IP2-2006-03917
IP2-2006-03929
IP2-2006-03941
IP2-2006-03962
IP2-2006-03964
IP2-2006-03965
IP2-2006-03974
IP2-2006-07009
IP2-2009-03355
I P2-2009-02085
I P3-2009-04739*
System Health Reports Unit 2 Service Water System, First Quarter 2009, Second Quarter 2009, Third Quarter 2009 Attachment

Drawings

9321-F-2033, Service and Cooling Water, Rev. 80 9321-F-2722, Nuclear Service Water System (sheet 1), Rev. 117
Nuclear Service Water System (sheet 2), Rev. 69 A234191, Flow Diagram Closed Cooling Water System. Rev. 45 Procedures
2-AOP-CCW-1, Loss of Component Cooling Water, Rev. 1 2-AOP-SW-1, Service Water Malfunction.
Rev. 3 2-AOP-UCCW-1, Leakage into CCW System. Rev. 3 2-S0P-24.1, Service Water System Operation.
Rev. 57 2-S0P-27.3.1.3, 23 EDG Manual Operation, Rev. 19 2-S0P-4.1.2, Component Cooling System Operation.
Rev. 34 Program Documents
EN-DC-150, Condition Monitoring of Maintenance Rule Structures, Rev. 0
IP3-RPT-UNSPEC-03499, Indian Point Units 2&3 Eddy Current Program, Rev. 1
SEP-SW-001, NRC Generic Letter 89-13
Service Water Program, Rev. 2
EN-DC-340, Microbiologically Influenced Corrosion Monitoring Program, Rev. 0 Miscellaneous ConEd Letter, S. Bram to NRC, dated 2/2/90, Response to
GL 98*13 ConEd Letter, S. Bram to NRC, dated 7/19/91, Implementation Status of
GL 98-13 ConEd Letter, S. Bram to NRC. dated 2111/92, Updated Implementation Status of
GL 98-13 ConEd Letter, S. Bram to NRC, dated 9/7/94, Service Water System Operational Performance Inspection

lO-IP3-2009-00019, IPEC Heat Sink Performance, dated 6117/09 Section 1 R11: Licensed Operator Regualification Program Procedures

2-E-0, Reactor Trip or Safety Injection, Rev. 2 2-E-1, Loss of Reactor or Secondary Coolant, Rev. 0 2-ES-1.3, Transfer to Cold Leg Recirculation, Rev. 2 2-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, Rev. 0 2-AOP-LEAK-1, Sudden Increase in Reactor Coolant System Leakage, Rev, 7 2-AOP-INST

-1, Instrument/Controller Failures.

Rev. 5 Miscellaneous
LRQ-SES-22, SG Pressure Channel Failure, RCS Leak, LBLOCA, Transition to
ES-1.3 with Equipment Failures, Rev. 2 Radiological Emergency Data Form Drill, Notification
  1. 2, 10/6109 at 8:02 Radiological Emergency Data Form Drill, Notification
  1. 3,10/6109

at 8:28

Section 1R12: Maintenance

Effectiveness

Procedures

2-PT-M110, Appendix R DG Functional Test, Rev. 2 2-PT-M110, Appendix R DG Functional Test, Rev. 1, performed
06/12/08 Attachment

Condition Reports

(CR-)

IP2-2009-03053
IP2-2009-00721
IP2-2009-00199
IP2-2009-04021
I P2-2009-04038
I P2-2009-4042
I P2-2009-04259
IP2-2009-04
744 I P2 -2009-04806

Drawings

501424, Station Blackout & Appendix R Diesel Generator Set PY800 Manual Double Wall U/L Listed -Fuel Oil Day Tank Mechanical, Rev. 0
501425, Station Blackout & Appendix R Diesel Generator Set Wiring Digram Fuel Oil Day Tank Electrical, Rev. 0 Miscellaneous Operators Risk Report, dated 10/2/09 Section 1 R13: Maintenance Risk Assessments and Emergent Work Control Procedures
IP-SMM-WM-101, On-Line Risk Assessment, Rev. 2
IP-SMM*WM-103, Control of Maintenance Activities Under Allowable Outage Time (AOT) Action Statements, Rev. 1 Condition Reports (CR-)
IP2-2009-04420

Miscellaneous

Operator Risk Report for October 5, 2009

Section 1R15: Operability

Evaluations

Procedures

EN-OP-111, Operational Decision-Making Issue (ODMI) Process, Rev. 3
OAP-005, Narrative Logs. Rev. 2
OAP-017, Plant Surveillance and Operator Rounds, Rev. 6 Condition Reports (CR-)
IP2-2009-05300
IP2-2009-3527*
IP2-2009-2469
IP2-2009-3564
IP2-2008-4212

Drawings

A208088-43, 480 Vac. Switchgears

& 22, Bus 2A, 3A, 5A & 6A 9321-F-2030, Flow Diagram Fuel Oil Diesel Generators, Rev. 39 Miscellaneous

3.8 Electrical Power, Technical Requirements Manual (TRM), Rev. 1
EC 5000033794, IP2 Station Blackout and Appendix R Diesel Generator Set, Rev. 1 Calculations
IP-CALC-06-00299, EDG Fuel Oil Day Tank Low Level Analytical Limit, Rev. 0 Work Orders
00199669 Attachment

Section 1R19: Post-Maintenance

Testing Procedures
2-PMP-008-CCW, Inspection/Repair of the Component Coofing Pump, Rev. 2
MSl-B-007-A, Chesterton Seals (Series 123), Rev. 7
CUP-B-002-A, Falk Type T10ff20 Steefflex Couplfng, Rev. 8 2-PT-Q034B,
PCV-1310A

and

PCV-1310B
Nitrogen Supply, Rev. 6 2-PT-Q034, 22 Auxiliary Feed Pump, Rev. 26
Fire Protection System Operation, Rev. 22 O-Vl V-413-MOV, Motor Operated Valve Minor Preventive Maintenance, Rev. 4
PT-M54, Fan Cooler Units Operation, performed
11/16/09 2-PT -M021 B, Emergency Diesel Generator load Test, performed
11/10/09 2-PT-2Y018F, Transfer Switch
EDD-6 (22 EDG) Test, performed
11110109 O-Vl V-465-VSR,
CLA-VAL Pressure Relief Valves Maintenance and Inspection, performed
10/30/09
CLA-VAL Pressure Relief Valves Maintenance and Inspection, performed
11/04/09 Condition Reports (CR-)
IP2-2009-04215
IP2-2009-4505
IP2-2009-5198

Work Orders

00130508-01
00195301-01
51549691
51549692
51549733
51549690
52203978
00158729
00158730
00158731
00158732
51268377
51548550
52037213
51702237
52028737-01

Miscellaneous

PGI-00518, AOV Component Level Calc. for 22 Auxiliary Feed Pump Discharge Flow Control Valves to Steam Generators,
FCV-405A,
FCV-405B,
FCV-405C and
FCV-405D, Rev. 1 Drawings A227551, Fire Protection System Diagram, Rev. 63 Section 1 R22: Surveillance Testing Procedures
2-PT-Q013-DS085, Valve
FCV-405A 1ST Data Sheet, Rev. 20
CR-IP2-2009-00666, Root Cause Analysis Report, CST Underground Recirculation Line Leak, Rev. 0
EN-DC-325, Component Performance Monitoring, Rev. 4
EN-DC-332, Inservice Testing, Rev. 0 2-PT-Q001C, 23 Battery Surveillance and Charging,
2/14/09 2-PTOQ028A, 21 Residual Heat Removal Pump, performed
11/19/09 2-PT-Q029A, 21 Safety Injection Pump, performed
10/22/09 Miscellaneous
IP2-AFW DBD, Auxiliary Feedwater System, Rev. 1 PG/-00497.
Auxiliary Feedwater System Air Operated Valve Functional and Maximum Expected Differential Pressure Calculation, Rev. 1 IEEE 450, IEEE Recommended Practice for Maintenance, Testing and Replacement, dated 1995 Attachment

Drawings

251783. Flow Diagram Auxiliary Coolant System Residual Heat Removal Pumps, Rev. 0 Section 1 EP2: Alert and Notification System Evaluation

Procedures

Alert and Notification System for the Indian Point Energy Center Entergy Nuclear, Rev. 4 Indian Point Energy Center Emergency Preparedness Plan, Rev. 8
IP-EP-AD30, lPEC ATI Siren System Administration, Rev. 2
IP-EP-AD31, IPEC ATI Siren System Maintenance Administration, Rev. 0 Alert Notification System Test Failure Root Cause Evaluation Report, Rev. 1
IP-EP-AD35, IPEC ATI Siren Site Annual Preventive Maintenance, Rev. 2 IPEC ATI Siren Annual Preventive Maintenance Test Records, February 10, 2009 ANS related Condition Reports, December 2007 -December 2009

Section 1EP3: Emergency

Preparedness Organization Staffing and Augmentation System Procedures
IP-EP-AD9, Notification Systems Testing and Maintenance, Rev. 7 IPEC ERO Roster Indian Point Energy Center Emergency Response Training Program Curriculum, Rev. 24 October 27, 2009, Entergy Nuclear Northeast, Indian Point Energy Center Emergency Preparedness Unit 3 Off-Hours Mobilization StaffingITralning Drill Performance Report. Drill Number 2009-5 September
17. 2009. Indian Point Energy Center Emergency Response Organization Off-hours Notification Test 3Q09 Section 1 EP4: Emergency Action Level and Emergency Plan Changes Procedures
EN-EP-305.
Emergency Planning 10CFR50.54 (q) Review Program, Rev. 1 10
CFR 50.54(q) screenings and evaluations from December 2008 to November 2009 Section 1 EP5: Correction of Emergency Preparedness Weaknesses

Procedures

EN-Ll-102, Corrective Action Process, Rev. 13
QA-07-2008-IP-1, Quality Assurance Audit Report
QA-07-2009-IP-1, Quality Assurance Audit Report
QS-200B-IP-16, IPEC QA Follow-up of AFI from Emergency Plan Surveillance
QS-2008-IP-16
QS-2008-IP-02.
QA Evaluation of the IPEC 2/6/08 Training Drill LO-IP3LO-2007-00185.
IPEC Snapshot Self-Assessment Report. ANS Siren System Performance
IP3-LO-2009-00092.
IPEG Focused Self-Assessment Report. EP INPO Based Focus Seif Assessment October 29.2008, Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 3 Training Drill Performance Report. Drill Number 2008-5 November 19. 2008. Entergy Nuclear Northeast, Indian Point Energy Center. Emergency Preparedness Unit 3 Training Drill Performance Report, Drill Number 2008-6 December 3, 2008, Entergy Nuclear Northeast, Indian Point Energy Center. Emergency Preparedness Unit 3 Training Drill Performance Report. Drill Number 2008-7 Attachment May 13, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 2 Training Drill Performance Report, Drill Number 2009-2 September
9, 2009 Entergy Nuclear Northeast, Indian Point Energy Center, Emergency Preparedness Unit 2 Training Drill Performance Report, Drill Number 2009-3 Sections 2051/2052:
Access Control to Radiologically Significant Areas/ALARA
Planning and Controls Procedures
EN-U-114, Performance Indicator Process, Rev. 4
EN-RP-100, Radworker Expectations, Rev. 3
EN-RP-101, Access Control for Radiologically Controlled Areas, Rev. 4
EN-RP-102, Radiological Control, Rev. 2 EN-RP-105.
Radiation Work Permits, Rev. 7
EN-RP-108, Radiation Protection Posting, Rev. 7
EN-RP-110, ALARA Program, Rev. 6
EN-RP-121, Radioactive Material Contro', Rev. 4
EN-RP-122, Alpha Monitoring, Rev. 3
EN-RP-131, Air Sampling, Rev. 7
EN-RP-141, Job Coverage, Rev. 4
EN-RP-151, Radiological Diving, Rev.2
EN-RP-202, Personnel Monitoring, Rev. 7
EN-RP-203, Dose Assessment, Rev. 3
EN-RP-204, Special Monitoring Requirements, Rev. 3
EN-RP-205, Prenatal Monitoring, Rev. 4
EN-RP-208, Whole Body Counting and In-Vitro Bioassay, Rev. 3 O-RP-RWP-411, Discrete Radioactive Particle Controls, Rev. 0 O-RP-RWM-901, Interim Radwaste Storage Facility and Outside Radioactive Material Storage Area, Rev. 2 Condition Reports {CR-}
IP2-2009-02184
IP2-2009-02217
IP2-2009-02425
IP2-2009-02484
IP2-2009-02505
IP2-2009-03335
I P2 -2009-03524
I P2-2009-03578
IP2-2009-03674
IP2-2009-03699
IP2-2009-03978
IP3-2009-00709
IP3-2009-01348
IP3-2009-01439
IP3-2009-01527
I P3-2009-0
1769
IP3-2009-01879
IP3-2009-01981
I P3-2009-01984
IP3-2009-02198
IP3-2009-02200
IP3-2009-02201
I P3-2009-02619
IP3-2009-03110
IP3-2009-03721
IP3-2009-03778
I P3-2009-03973

Miscellaneous

ALARA Committee Reviews IPEC 5 Year ALARA Plan 2009*2013
IP3-LO-2009-00074, IPEC Snapshot Self-Assessment Report -HRA & LHRA Controls Oversight Observation Checklists:
2C-IPEC-2009-0202, 0205, 0223, 0224,0241,0266,0279, 0281,0368,0496,0520,0531
Radiation Protection Attention Logs (Electronic Dosimeter Alarms) Monthly Effluent Release Reports

Section 40A1: Performance

Indicator Verification

Procedures

NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 2 Attachment
O-SOP-LEAKRA
TE"001, RCS Leakrate Surveillance, Evaluation and Leak Identification, Rev. 1 EN"LI" 114, Performance Indicator Process, dated 10/08/09
EN-U-114, Performance Indicator Process, dated 04/09/09 EN-U-114.
Performance Indicator Process, dated 10/09/08
EN-U-114, Performance Indicator dated 01/14/09
EN-U-114, Performance Indicator Process, dated 10/08/09
EN-EP-201, Performance Indicators, Rev. 9
IP-EP-AD5, Emergency Preparedness Perfonnance Indicator Program, Rev. 3 Performance Indicator Data, 4th quarter 2008 -3 rd quarter 2009 Condition Reports (CR-) IP2-2009-05032

Section 40A2: Identification

and Resolution of Problems Procedures

OAP-045, Operator Burden Program, Rev. 1 Condition Reports (CR-)
IP2-2009-4860

Work Orders

00205770
00179027
00177347 Section 40A3; Event Follow-up

Procedures

IP-SMM-IS-104, Electrical Work Practices and Procedure Handbook, Rev. 1
EN-HU-102, Human Performance Tools, Rev. 5
EN-HU-105, Human Performance

-Managed Defenses, Rev. 6 2-0AP-IB-1.

Loss of Power to an Instrument Bus, Rev. 7 2-S0P-27.3.1.3, 23 Emergency Diesel Manual Operation, Rev. 19
EN-WM-105, Planning, Rev. 5 Condition Reports (CR-) I P2 -2009-4860

Work Orders

00163807

Section 40A5: Other Activities

Procedures

2-DCS-006, Vertical Cask Transporter Use 2-DCS-031
GEN, Fuel Selection for Dry Cask Storage, Rev. 0 10
CFR 72.212, Evaluation Report, Site Specific Appendix F, iP2 Specific Infonnation
EN-DC-147, IPEC ISFSI Fire Hazards Analysis Roadmap, dated 12/20/2007, Rev. 2
EN-DC-161, Control of Combustibles Rev. 3 Holtec International
HI-STORM FSAR, Report
HI-2002444, Rev. 4 Transient Combustible Evaluation No.09-015, dated 10/19/2009

Condition Reports

(CR-) IP2..;2009-03910*

IP2-2009-05228*
  • CR Initiated as a result of this inspection.
ADAMS ALARA AMP ANS CCW CEDE CFR CR OMB ORP EAL EDG ENTERGY EP EPZ ERO FCU FCV FSAR FZ HRA IMC IPEC ISFSI 1ST LBLOCA LOE NCV NEI NOV NRC PI PM RHR ROP RSPS RWP SDE SDP
SSC SW TCE TEDE TS UFSAR UHS WO

LIST OF ACRONYMS

Document and Management

System As Low As is Reasonably

Achievable

Amplifier

Alert and Notification

System Component

Cooling Water Committed

Effective

Dose Equivalent

Code of Federal Regulations

Condition

Report Digital Message Board Division of Reactor Projects Emergency

Action Level Emergency

Diesel Generator

Entergy Nuclear Northeast

Emergency

Preparedness

Emergency

Planning Zone Emergency

Response Organization

Fan Cooler Unit Flow Control Valve Final Safety Analysis Report Fire Zone High Radiation

Area Inspection

Manual Chapter Indian Point Energy Center Independent

Spent Fuel Storage Installation

In-Service

Test Large Break Loss-of-Coolant

Low Dose Equivalent

Non-cited

Violation

Nuclear Energy Institute

Notice of Violation

Nuclear Regulatory

Commission

Performance

Indicator

Preventative

Maintenance

Residual Heat Removal Reactor Oversight

Process Risk Significant

Planning Standard Radiation

Work Permit Shallow Dose Equivalent

Significance

Determination

Process Structures, Systems, and Components

Service Water Transient

Combustible

Evaluation

Total Effective

Dose Equivalent

Technical

Specifications

Updated Final Safety Evaluation

Report Ultimate Heat Sink Work Order Attachment

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