IR 05000289/2009005

IR 05000289-09-005 on 10-01-09 - 12-31-09 for Three Mile Island, Unit 1, Routine Integrated Inspection

ML100330674
Person / Time
Site:	Three Mile Island
Issue date:	02/02/2010
From:	Bellamy R R NRC/RGN-I/DRP/PB6
To:	Pardee C G Exelon Generation Co, Exelon Nuclear
BELLAMY, RR
References
FOIA/PA-2010-0209 IR-09-005
Download: ML100330674 (44)

Text

I I UNITED STATES I NUCLEAR REGULATORY COMMISSION REGiOilii I *i 475 ALLENDALE ROAD KING OF PRUSSIA. PENNSYLVANIA 19406*1415 I February 2010 I I I I Mr. Charles Senior Vice President, Exelon Generation Company, LLC I President and Chief Nuclear Officer (CNO), Exelon Nuclear 4300 Winfield Road Warrenvme, IL 60555 THREE MILE ISLAND STATION. UNIT 1 -NRC INSPECTION REPORT I !

Dear Mr. Pardee:

I!. On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Three Mile Island, Unit 1 (TMI) facility.

The enclosed inspection I report documents the inspection results, which were discussed on January 15, 2010, with Mr. William Noll 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.

This report documents one NRC-identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements.

Additionally, a licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance and because it is entered into your corrective action program, the NRC is treating the finding as a violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administration, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspectors at the Three Mile Island facility.

In addition, if you disagree with the characterization of the cross-cutting aspect of any finding in this report, 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, Region I and the NRC Senior Resident Inspectors at the Three Mile Island facility.

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 enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of I 1 I* 2 NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room). We appreciate your cooperation.

Please contact me at 610-337-5200 if you have any questions . regarding this letter.

Sincerely.

Ronald R. Bellamy, Ph.D., Chief Projects Branch 6 Division of Reactor Projects Docket 50-289 License DPR-50 Inspection Report

w/Attachment:

Supplemental cc w/encl: Distribution via ListServ I NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.htmf (the Public Electronic Reading Room). I We apprec:iate your cooperation.

Please contact me at 610-337-5200 if you have any questions regarding this letter. I

Sincerely.

IRA! I* Ronald R. Bellamy. Ph.D., Chief Projects Branch 6 Division of Reactor Projects Distribution w/enel: S. Collins, RA (R10RAMAIL Resource)

M. Dapas, ORA (R10RAMAIL Resource)

D. Lew, DRP (R1DRPMAIL Resource)

J. Clifford, DRP (R1DRPMAIL Resource)

D. Roberts, DRS (R1DRSMail Resource)

P. Wilson, DRS (R1DRSMaii Resource)

R. Bellamy, DRP . S. Barber, DRP C. Newport, DRP J. Greives, DRP E. Gray, DRS P. Kaufman, DRS E. Burket, DRS D. Kern, DRP, SRI J. Brand, DRP, RI C. LaRegina, DRP, OA L. Trocine, RI OEDO RidsNrrPMThreeMilelsland Resource RidsNrrDorl Lp11-2 Resource ROPReportsResource@nrc.gov SUNSI Review Complete:

RRB (Reviewer's Initials)

ML 100330674 DOCUMENT NAME: G:\ORP\BRANCH6\+++THREE MILE ISLAND\TIVIlINSPECTION REPORTS\

TMI 09-005.DOC After declaring this document "An Official Agency Record" it will be released to the Public. TO receive a copy of this document, indicate In the box: "c"", Copy without altachmenVenclosure "E":::Copy wi1h attachment/enclosure, "N"=No copy OFFICE HI/DRP I RIIORP J NAME OKern/DK RBellamyl RRB DATE 01/15/10 02/01/10 OFFICIAL RECORD COPY Docket No: License No: Report No: Licensee:

Facility: . Location:

Dates: Inspectors:

Approved by: U.S. NUCLEAR REGULATORY REGION DPR-50 05000289/2009005 Exelon Generation Company Three Mile Island Station, Unit 1 Middletown, PA 17057 October 1 throug h December 31, 2009 D. Kern, Senior Resident Inspector J. Brand, Resident Inspector C. Newport, Project Engineer R. Nimitz, Senior Health Physicist H. Gray, Senior Reactor Inspector P. Kaufman, Senior Reactor Inspector J. Lilliendahl, Reactor Inspector E. Burket, Reactor Inspector D. Spindler, Resident Inspector D. Everhart, Physical Security Inspector R. Bellamy, Ph.D., Chief Projects Branch 6 Division of Reactor Projects (DRP) Enclosure

SUMMARY

OF

IR 05000289/2009005; 10/1/2009-12131/2009;

Exelon Generation Company, LLC; Three Mile Island, Unit 1, Routine integrated report. The report covered a three-month period of baseline inspection conducted by resident inspectors and announced inspections by regional inspectors.

One Green finding was identified, which was a non-cited violation (NCV). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,

Determination Process (SOP)." Findings for which the SOP does not apply may be Green or be assigned a severity level after NRC management review. Cross-cutting aspects associated with findings are determined using IMC 0305, "Operating Reactor Assessment Program," dated December 2009. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight," Rev. 4, dated December 2006.

Cornerstone: Mitigating

Systems

Green.

The inspectors identified a Green, non-cited violation of the Three Mile Island operating license for not adequately conSidering the effects of carbon dioxide (C02) toxicity.

Specifically, for a fire in the relay room which causes a C02 initiation and a control room evacuation, C02 would migrate into adjacent areas. Because operators must enter these adjacent areas to perform time critical.

safe shutdown actions, the potential existed to delay or incapacrtate the operators which would negatively impact the ability to safely shutdown the plant. Exelon made procedural and training changes to ensure that operators immediately don self-contained breathing apparatus in the event of a control room evacuation after a C02 initiation in the relay room. The finding was more than minor because it was associated with the external factors (fire) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

This issue was found to be of very low safety significance (Green) due to the low degradation rating resulting from the minimal impact on the fire protection program. This finding did not have a cross-cutting aspect because the most significant contributor of the performance deficiency was not reflective of current licensee performance. (Section 40A5)

Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors.

Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. This violation and corrective actions are listed in Section 40A7 of this report.

4

REPORT DETAILS

Summary of Plant Status Three Mile Island, Unit 1 (TMI) began the inspection period at 100 percent rated thermal power and gradually reduced power due to end-of-cycle fuel depletion.

Operators reduced power from 83 to 73 percent on October 23 and to 60 percent on October 24, as specified by Technical Specifications (TS) due to several main steam safety valves (MSSVs) not lifting within the specified setpoints during periodic surveillance tests (see section 1 R22). On October 25, operators began a plant shutdown and the turbine output breakers were opened on October 26, beginning the 18th refueling outage (T1 R18). On November 21, a small, amount of radioactive material inadvertently became airborne within the reactor building containment during once through steam generator (OTSG) replacement activities.

Consequently, several workers became contaminated and/or ingested small amounts of radioactive material.

This issue is the subject of a separate NRC inspection activity and will be documented in NRC Inspection Report 5000289/2010007.

Major work accomplished during this refueling outage included replacement of the 'A' and '8' OTSGs, alloy 600 dissimilar weld inspection and mitigation, replacement of 'A', '8', and '0' 120 volt vital bus inverters, cooling tower modification, repair of numerous cooling water system leaks due to microbiological induced corrosion, and reactor core refueling.

The reactor remained defueled at the close of the inspection period.

REACTOR SAFETY

Cornerstones:

Initiating Events, Mitigating Systems, Barrier Integrity 1 R01 Adverse Weather Protection (71111.01

-1 sample)

a. Inspection Scope

(Cold Weather) The inspectors walked down risk significant plant areas between December 15 and 29 to . assess Exelon's protection for cold weather conditions.

The inspectors evaluated outside instrument line conditions and the status of the heat trace system. The walkdown included the condensate storage tanks, the emergency diesel rooms, the emergency feedwater pump rooms and safety-related river water system components located within the heat exchanger vault. The inspectors also reviewed implementation of procedure WC-AA-10I.

Seasonal Readiness, Rev. 7 and OP-AA-108-111-1001, Severe Weather and Natural Disaster Guidelines, Rev, 4 for cold weather conditions. Findings No findings of significance were identified.

1 R04 ,!;mJipment Alignment (71111.04) Inspection Seoge Partial System Walkdowns (71111.04Q

-3 samples) The inspectors performed three partial system walkdown samples on the following systems and components:

On October 26, the inspectors walked down the 'B' low pressure injection (LPI) while the 'A' LPI train was running for shutdown On October 29, the inspectors walked down the 'A' LPI train and reactor system (RCS) in preparation for, and during RCS draindown to reactor vessel level for removal of the reactor vessel On November 2. the inspectors walked down portions of the'S' emergency 4KV, emergency 480v, and 'B' engineered safeguards actuation systems (ESASs) preparation for performance of the Engineered Safeguards Train 'A' Sequence and Power Transfer The partial system walkdowns were conducted to ensure redundant trains and standby equipment relied on to remain operable for accident mitigation were properly aligned. Complete System Walkdown (71111.04S

-1 sample) From November 13 thru 19, the inspectors performed one complete system sample on the spent fuel pool cooling system while the plant was in shutdown mode refueling outage T1 R18, with all fuel in the spent fuel pool (defueled).

This was performed while the 'B' spent fuel pool pump (SF-P-'I 8) was running with temporary power supply per OP-TM-731-525, De-energizing 1S ES 480V Rev. 3. The inspectors conducted a detailed review of the alignment and condition the system using piping and information diagrams and evaluated open corrective program reports for impact on system operation.

In addition, the inspectors reviewed associated protected equipment log. and interviewed the system engineer and room operators.

Additional documents reviewed are listed in the I

b. Findings

No findings of significance were identified.

1 R05 Fire I

.1 Quarterly

Defense-in-Depth Walkdown of Plant Areas (71111.05Q -4 samples)

a. Inspection Scope

The inspectors conducted fire protection inspections for several plant fire based on the presence of equipment important to safety within boundaries.

The inspectors conducted plant walkdowns and verified the areas were as described in the TMI Fire Hazard Analysis Report, and that fire protection features were properly controlled per surveillance procedure 1038. Administrative Protection Program, Rev. 74. The plant walkdowns were conducted throughout inspection period and included assessment of transient combustible material control, fire detection and suppression equipment operability.

and compensatory measures established for degraded fire protection equipment in accordance with procedure Fire Protection System impairment Control. Rev. 6. In addition, the inspectors verified that applicable clearances between fire doors and floors met the Enclosure criteria of Attachment 1 of Engineering Technical Evaluation CC-AA-309-101, Engineering Technical Evaluations, Rev. 10. Fire zones and areas inspected included:

• Fire Zone DG-FA"1, Diesel Generator Building, Diesel Generator A;

• Fire Zone AB-FZ-6A, Auxiliary Building Elevation 305', Motor Control Center B;

• Fire Zone CB-FA-2D, Control Building Elevation 322', East Inverter Room; and

• Fire Zone CB-FA-2E, Control Building Elevation 322', West Inverter Room.

b. Findings

No findings of significance were identified . . 2 Fire Protection (Triennial)

Unresolved Item (URI) Follow-up

a. Inspection Scope

(Closed) URI 05000289/2008009-02 Potential C02 Migration Outside the Relay Room This unresolved item was opened pending NRC review of the licensee's calculation of the concentration, pressurization, and migration for the C02 extinguishing system in the Three Mile Island relay room. Specifically, the potential existed for C02 to migrate outside the relay room and affect theabiHty of operators to perform alternative shutdown activities.

Likewise, with a C02 migration potential outside the relay room, the potential existed for less than adequate C02 concentration in the relay room for fire suppression.

Finally, the potential existed for C02 overpressurization in the relay room. The inspectors reviewed the Three Mile Island Fire Hazards Analysis and the National Fire Protection Association 12, Carbon Dioxide Extinguishing Systems, to determine the requirements for the relay room C02 system. The inspectors reviewed the new design analyses for the relay room fire scenario, TM-09-00527 and TM-08-00963, to evaluate the ability of the relay room C02 system to perform its design function, and to evaluate any potential C02 migration issues. Finally, the inspectors interviewed system engineers,design engineers, and operations personnel to assess the impact of the C02 system on the fire safe shutdown procedure.

The inspectors agreed with Exelon's evaluation that the relay room would reach and maintain an adequate concentration of C02, and there is reasonable assurance that the relay room will not become overpressurized.

b. FindIngs

Introduction.

The inspectors identified a Green, non-cited violation of the Three Mile Island operating license for not adequately considering the effects of C02 toxicity.

Specifically, for a fire in the relay room which causes a C02 initiation and a control room evacuation, C02 would migrate into adjacent areas. Because operators must enter these adjacent areas to perform time critical, safe shutdown actions, the potential existed to delay or incapacitate the operators which would negatively impact the ability to safely shutdown the plant

Description.

The Three Mile Island relay room is an alternate shutdown area. As such, the potential exists that a control room evacuation may be required for a large fire in the Enclosure relay room. The relay room is equipped with a C02 total flooding extinguishing system. The ESAS room and the 1 E 4160 volts alternating current (VAC) switchgear room are adjacent to the relay room. As part of a fire induced control room evacuation, time critical actions to transfer control power are required in the ESAS room and the 1 E 4160VAC switchgear rooms. The inspectors observed gaps under the doors between the relay room and the adjacent rooms, which would provide a path for C02 migration.

In response to NRC questions about C02 concentration in rooms adjacent to the relay room, Exelon performed an analysis in TM w 08-00963, "C02 Discharge into Control Building Relay Room." This analysis indicated that although C02 concentration was not adversely affected in the relay room, worst case C02 concentrations could reach 18% in the ESAS room and 10% in the 1 E 4160VAC switchgear room. The National Institute for Occupational Safety and Health documents that a C02 concentration above 4% is "Immediately Dangerous to Life or Health." Concentrations of C02 above 10% can cause confusion, rapid breathing, and unconsciousness.

The impact of the potential C02 migration is mitigated by the use of a wintergreen odorizer in the C02 system which will release a wintergreen scent in the event of a C02 initiation.

The scent is intended to alert personnel of the presence of C02. In summary, for a fire in the relay room which causes a C02 initiation and a control room evacuation, C02 would migrate into the adjacent rooms (ESAS room and 1E 4160VAC switchgear room). As part of the control room evacuation, an operator would enter the adjacent rooms to align the remote shutdown panel. Because of the C02 in the adjacent rooms, the potential existed that the operator would be delayed or incapacitated while performing the necessary alignment.

Exelon made procedural and training changes to ensure that operators immediately don self-contained breathing apparatus in the event of a control room evacuation with a C02 Initiation in the relay room. Exelon also initiated assignment

1. 17 in IR 817422 to consider modifications to prevent the C02 migration.

The performance deficiency associated with this issue is the failure to adequately consider the toxicity effects of C02.

Analysis:

The inspectors determined that the failure to consider the toxicity effects of C02 was a performance deficiency that was reasonably within Exelon's ability to foresee and prevent. The finding was more than minor because it was associated with the external factors (fire) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRC's regulatory function, and was not the result of any willful violation of NRC requirements.

In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors concluded that this finding was of very low safety sJgnificance (Green) due to the low degradation rating resulting from the minimal impact on the fire protection program. Specifically, because of the presence of the Wintergreen scent and the low probability of needing to conduct a control room evacuation following automatic C02 initiation in the relay room, the inspectors determined that the finding represented a minimal impact to the reliability of the safe shutdown procedure.

In accordance with Enclosure I , i '. 8 steps 1.2 and 1.3 of IMC 0609, Appendix F, this finding was assigned a low rating, and screened to Green. This finding did not have a cross-cutting aspect the most significant contributor of the performance deficiency was not reflective I current licensee performance.

I

Enforcement.

Three Mile Island Nuclear Station Operating License Condition 2.c.( requires that Exelon shall implement and maintain in effect all provisions of the Protection Program as described in the Updated Final Safety Analysis Report for TMI-1. The UFSAR section 9.9.2 states that the Fire Hazards Analysis Report considered to be part of the Fire ProteCtion Program. Section 5.0.E.5 of the Hazards Analysis Report states that consideration has been given to the toxicity of Contrary to the above, from the time of plant operation in 1979 until February 3, Exelon did not adequately consider the toxicity of C02. Specifically, C02 was to migrate from the relay room to adjacent areas where operators performing time actions could become delayed or unconscious due to the effects of C02. Because finding is of very low safety significance and has been entered into Exelon's action program (IR 817422), this violation is being treated as a non-cited consistent with Section VI.A 1 of the NRC Enforcement policy. 05000289/2009005 N 01, Potential C02 Migration Outside the Relay Room Fire 1 R07 Heat Sink Performance (71111.07

-2 samples)

a. Inspection Scope

The inspectors performed two inspection samples. The inspectors reviewed the removal capability of the 'A' safety-related decay heat service closed cooling cooler (DG-C-2A)per DC-C2A Heat Transfer Test, Rev. 5, in November 2009 during the cooldown for T1R18 refueling outage. This component a water-to-water heat exchanger with river water from the decay heat river water on the tube side and decay heat closed cooling on the shell side. The reviewed Issue Report (IR) 987930 and engineering change request (ECR)000, T1 R18 DC-C-2A Heat Transfer Performance (GL 89-13 Test), which evaluated heat load mismatch identified during the heat balance test, between the heat from the decay closed system and the heat added into the decay river Engineers determined the heat load mismatch was primarily due to flow accuracy issues during testing. Engineers factored the overall impact of the instrument inaccuracy into the flow balance calculation and concluded the DC-C-2A the minimum cleanliness requirement of greater than 42 percent, the heater was able support operation for the upcoming operating cycle, and it would be able to perform design safety function if required.

This heat exchanger is scheduled to be inspected, and cleaned during the next refueling outage T1 R19 (scheduled for fall 2011). The inspectors performed field walk downs and interviewed the system design engineers to verify the inspection results were appropriately categorized pre-established acceptance criteria.

the frequency of inspection was sufficient, and various bio-fouling treatment processes were used to ensure continued satisfactory exchanger The inspectors also verified the heat transfer capability of the'S' safety-related heat service closed cooling water cooler (DC-C-2S).

This component is a heat exchanger with river water from the decay heat river water system on the tube and decay heat closed cooling on the shell side. The inspectors reviewed the I /'. 9 I inspections completed on November 20 per ER-TM-340-1002, Inspection of NS-C1A (B){C)(D), IC-C1A (B) and SR-C1A (B)(C)(D), Rev. 1. The inspectors I performed an independent internal visual inspection of the heat exchanger to assess current material condition, and interviewed the field technician, system engineer, other key personnel responsible for oversight of the heat exchangers to assess adequacy of performance I

b. Findings

I No findings of significance were identified.

I 1 ROB Inservice Inspection (ISI}(71111 08 -1 Sample)

a. Inspection Scope

I The inspectors observed selected samples of in-process nondestructive (NDE) activities and reviewed documentation of completed NDE and activities.

The sample selection was based on the inspection procedure objectives risk priority of those components and systems where degradation would result in r significant increase in risk of core damage. The observations and documentation review I were performed to determine whether the activities were performed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements.

This 151 inspection was conducted concurrently with the inspection of the Generator Replacement (SGR) project. The SGR included an array of tasks including welding, post weld heat treatment, nondestructive testing of welds radiography and visual test methods, and hydrostatic pressure testing. As preservice testing including ultrasonic testing (UT) and eddy current testing was to finished welds. The inspectors observed samples of these tests in progress reviewed both the results and the applicable procedures.

For 151, the observed the performance of three NDE activities in process and documentation and examination reports for an additional three NDE activities. inspectors reviewed five samples of welding activities on a pressure boundary reviewed ASME repair packages for repair/replacements performed during this cycle. The inspection included a walk down of a replacement weld in the nuclear water piping The inspectors observed manual UT, eddy current testing, and magnetic particle (MT) and reviewed inspection documentation of liquid penetrant testing (PT) and testing (VT) activities to evaluate effectiveness of the process, examiner, and in identifying degradation of risk significant systems, structures and components and evaluate the activities for compliance with the requirements of ASME Section XI of Boiler and Pressure Vessel Code. The inspectors reviewed the NDE report of integral support FW112 and its related inspection The inspectors reviewed the procedures used to perform visual examinations for indications of boric acid leaks from pressure retaining components.

The inspectors reviewed a sample of issue reports and action requests initiated as a result of the inspections performed in accordance with the licensee's boric acid control program. The inspectors selected IRs that identified evidence of both active and inactive leak locations i 10 which could result in degradation of safety significant components.

The reviewed five IRs shown in the Attachment which identified active and I leaks identified through plant walkdowns performed during the plant shutdown.

The inspectors reviewed operability evaluations and corrective actions provided in the IR and I determined that the actions specified were consistent with the requirements of the ASME j-Code and 10 CFR 50, Appendix B, Criterion XVI. I The inspectors performed a visual evaluation of selected portions of the TMI building containment liner coating and moisture barrier at the concrete floor slab containment vertical wall intersection.

The evaluation was made to compliance with the requirements of ASME Section XI, IWE (requirements for Class and Metallic Liners of Class CC Tlie inspectors observed the corrective actions being taken to address minor, identified corrosion activity at the containment liner plate/concrete floor interface. licensee removed the moisture barrier, caulking material, and some concrete in locations and cleaned and corrected identified problems prior to performing a inspection (VT-3) of the moisture barrier area. The visual inspection was to assess condition of the containment liner plate, determine the presence of moisture, and the extent of corrosion activity.

The inspectors reviewed the wall determinations and observed the restorative actions including local areas of welding final wall thickness determinations by UT thickness measurements.

The drawings indicated the liner was originally fabricated using carbon steel plate of 3/8" nominal thickness (walls and dome) and 3/4" nominal thickness (below concrete floor slab). The inspectors compared the as final wall thicknesses in areas with the design minimum wall thickness requirement and found the material to within design

b. Findings

No findings of significance were identified.

1R12 Maintenance

Effectiveness (71111.12Q

-1 sample)

a. Inspection Scope

The inspectors evaluated the listed sample for Maintenance Rule (MR) implementation by ensuring:

appropriate MR scoping; characterization of failed structures, systems, and components (5SCs); MR risk categorization of 5SCs; SSC performance criteria or goals; and appropriateness of corrective actions. Additionally, extent-of-condition follow-up, operability.

and functional failure determinations were reviewed to verify they were appropriate.

The inspectors verified that the issues were addressed as required by 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants; Nuclear Management and Resources Council 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Rev. 2; and Exelon procedure ER-AA-310, Implementation of the Maintenance Rule, Rev. 6. The inspectors verified that appropriate corrective actions were initiated and documented in IRs, and that engineers properly categorized failures as maintenance rule functional I failures and maintenance preventable functional failures, when applicable.

l. _Enclosure On October 29, while the plant was shutdown for refueling outage T1R18, operators identified one of the four main steam supply valves to the turbine driven emergency feed water pump failed to open during performance of OP-TM-424-241.

1ST of MS-V-10A and MS-V-10B.

Rev. 0 (IR 985703). This valve was replaced during the refueling outage to address unrelated seat leakage per AR-A2187690.

On December 26, operators identified the valve failed to open once more during performance of post maintenance testing after valve replacement per OP-241 (IR 1009479).

Corrective actions included initiation of a troubleshooting plan per AR-A2240580.

The inspectors performed field walk downs, interviewed the system engineer, and verified this issue would not have prevented the emergency feedwater system from performing its intended safety function.

In addition, the inspectors verified these failures were not maintenance rule functional failures, since the capability to supply steam to the turbine driven emergency feedwater pump was not affected. . 1 R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 -2 samples) a. I nspection Scope The inspectors reviewed the scheduling, control, and equipment restoration during the following maintenance activities to evaluate their effect on plant risk. This review was against criteria contained in Exelon Administrative Procedure 1082.1, TMI Risk Management Program, Rev. 8 and WC-M-101, On-Line Work Control Process, Rev. 16A. Additional documents reviewed are listed in the attachment. On October 29, station personnel detensioned the reactor vessel head and drained the RCS to mid-loop level (between 20 and 24 inches above the cold leg pipe centerline).

The Res cold legs and steam generator primary sides were also drained. The RCS time to boil in the event of a loss of the decay heat removal system function was 14 minutes and the associated shutdown safety risk profile was Yellow. The inspectors attended pre-job briefings, performed field walkdowns to verify protected eqUipment status, interviewed operators, and observed the draindown evolution from the control room and the reactor building. On November 13-19, electrical technicians removed the 'B' spentfuel pool cooling pump normal power supply and installed a temporary power supply per work order (WO) R2044279.

This temporary modification was done while the 'B' auxiliary transformer and 'A' vital bus were out of service for scheduled maintenance.

This condition elevated the associated shutdown safety risk profile to Yellow.

b. Findings

No findings of significance were identified.

1 R15 Operability Evaluations (71111.15

-2 samples)

a. Inspection Scope

The inspectors verified the selected degraded conditions were properly characterized, operability of the affected systems was properly evaluated in relation to TS requirements, applicable extent-of-condition reviews were performed, and no Enclosure unrecognized increase in plant risk resulted from the equipment issues. The inspectors referenced NRC IMC Part 9900, Operability Determinations

& Functionality Assessments for Resolutions of Degraded or Nonconforming Conditions Adverse to Quality or Safety and Exelon procedure OP-M-108-115, Operability Determinations, Rev. 9, to determine acceptability of the operability evaluations.

The inspectors reviewed operability evaluations for the following degraded equipment issues: On November 17, the inspectors reviewed IRs 996295,994874,995903, and 995908 which evaluated degraded floor penetration fire seals (#'s 898,899, 900, 905, and 906) for five safety related 120 Volt vital inverters.

Engineers determined the damage to the seals occurred during inverter replacement activities performed in November 2007 eC' and 'E' inverters)and in November 2009 ('A', 'B', and '0' inverters).

The evaluation also determined the 'c' and 'E' inverter seals (# 898 and 899) would not have been able to perform their 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated function.

However, engineers determined the degraded fire seals would have provided at least a one hour fire rating and due to defense-in-depth, the degraded fire seals would not have affected operability of the vital inverters in the event of a fire. Corrective actions included extent-of-condition walkdowns and repair of the fire seals. See Section 40A7.

• On December 15, the inspectors reviewed I R 000325 which documented elevated handwheel torque of NR-V-1A during the performance of a periodic surveillance test. Engineers determined the increased handwheel torque did not effect operability and decreased the surveillance test interval as a precautionary measure. The inspectors interviewed the system engineer.

verified that the increased handwheel torque did not affect operability.

and verified appropriate mitigating actions had been taken.

b. Findings

No findings of significance were identified. 1 R 19 Post Maintenance Testing (PMT) (71111.19

-11 samples)

a. Inspection Scope

The inspectors reviewed and/or observed the following post-maintenance test activities to ensure:

(1) the post-maintenance test was appropriate for the scope of the maintenance work completed;

(2) the acceptance criteria were clear and demonstrated operability of the component; and

(3) the post-maintenance test was performed in accordance with procedures. On October 23-25, technicians adjusted the spring load compression on several MSSVs and successfully retested the lift setpoints using procedure 1303-11.3.

Seven of 18 MSSVs initially had not lifted within their specified setpoints and required adjustment.

Long term actions to evaluate the cause, consequence, and recommended maintenance were initiated using IR 983712. On November 17-18, operators and electrical technicians performed procedures 26.1, Vital Power Inverter Maintenance, Rev. 3A, and 1107-2B. 120 Volts Vital Enclosure Electrical System, Rev. 24A (Interim Change IC-28349)after sCheduled

'A' inverter (EE-INV-1A)replacement activities; On November 24-28, operators performed PMT of the 'A' emergency diesel generator (EDG) per procedures 1405-3.2, Diesel Engine Maintenance, Rev. 40, and 1301-8.2, Diesel Generator Major Inspection (Mechanical), Rev. 86, after a scheduled major overhaul; On December 2-3, operators and electrical technicians performed procedures 26.1, Vital Power Inverter Maintenance, Rev. 3A, and 1107-2B, 120 Volts Vital Electrical System, Rev. 24A (Interim Change IC-28349)after scheduled

'D' inverter (EE-INV-1 D) replacement activities; On December 8, technicians performed diagnostic testing on MS-V-4A in accordance with WO C2020072 following valve and actuator replacement; On December 8, technicians performed diagnostic testing on MS-V-4B in accordance with WO C2020073 following valve and actuator replacement; On December 8-9, operators and electrical technicians performed procedures 26.1, Vital Power Inverter Maintenance, Rev. 3A, and 1107 -2B, 120 Volts Vital Electrical System, Rev. 24A (Interim Change IC-28349)after scheduled

'B' inverter (EE-INV-1B)replacement activities; On December 9-11, technicians performed PMT on the 'A' EDG room ventilation fan (AH-E-29A)in accordance with WO C2019922, following maintenance to address . increased fan noise level and apparent degradation; On December 10-13, operators performed PMT of the 'B' EDG per procedures 1405-3.2, Diesel Engine Maintenance, Rev. 40,1303-4.16 Emergency Power System, Rev. 121, and 1301-8.2, Diesel Generator Major Inspection (Mechanical).

after a scheduled major overhaul; On December 14. operators performed post maintenance testing of spent fuel cooling pump SF-P-1 B per OP-TM-731-525, Rev. 3, Attachment 1, after removal of temporary power installed to support scheduled outage activities; On December 25, operators partially performed OP-TM-211-204, 1ST of MU-V-36 and MU-V-37, Rev. 1. This operational test was performed as required by WO C2021371 following MU-V-37 actuator preventive maintenance and stem nut replacement.

b. Findings

No findings of significance were identified.

1 R20 Refueling and Other Outage Activities (71111.20

-1 sample)

a. Inspection Scope

Enclosure Station personnel conducted refueling outage T1 R18 from October 26 through the end of this inspection period and continuing into January 2010. The inspectors reviewed selected reactor shutdown, refueling.

and outage maintenance activities to determine whether shutdown safety functions (i.e., reactor decay heat removal, reactivity control, electrical power availability, reactor coolant inventory.

spent fuel cooling, and containment integrity)were properly maintained as required by TSs and TMI-2006-010, TMI-1 Outage Fuel Protection Criteria, Rev. 3. Specific attributes evaluated included configuration management, communications, instrumentation accuracy, and identification and resolution of problems.

The inspectors closely evaluated configuration and inventory control during periods of reduced RCS inventory due the associated increase in shutdown risk. The inspectors also performed inspections of accessible areas inside containment, interviewed applicable engineers, supervisors, and plant operators, and consulted with NRC specialists.

Additional documents reviewed during ttle inspection are listed in the Attachment.

Specific activities evaluated included: Reviewed the T1 R18 Shutdown Safety Plan to verify it properly addressed safety criteria prescribed by TMI-2006-01 0, TMI-1 Outage Fuel Protection Criteria.

Rev. 3. The inspectors noted that station management had rearranged the sequence of several outage activities in order to reduce planned shutdown risk levels. During previous refueling outages, IMI shutdown risk was typically Orange for several activities such as RCS mid loop operations.

The highest planned shutdown risk condition permitted by the T1R18 plan was Yellow; Plant shutdown and cooldown on October 25-26; Shutdown cooling (decay heat removal trains 'A' and 'B' operation)per 111, Rev. 5 and Rev. 5; Reactor vessel head removal and lift; Fuel offload; Engineered safeguards train 'A' emergency sequence and power transfer test; Reactor building emergency core cooling system (ECCS) sump as-found inspection; Reactor building walkdown to inspect for indication of RCS leakage and boric acid corrosion; ReS draindown to mid-loop operation in accordance with procedure 1103-11, Rev. 6? on October 29-31; Reactor building containment tendon detensioning and removal; Reactor building containment opening formation for removal and replacement of both OTSGs; and Fuel inspections, fuel assembly reconstitution, and fuel assembly upper endplate replacements performed in the spent fuel pool. 1 R22 Surveillance Testing (71111.22

-4 samples) Inspection Scope (1 Inservice Testing [1ST] Sample and 3 Routine The inspectors observed and/or reviewed the following operational surveillance tests to verify adequacy of the test to demonstrate operability of the required system or component safety function.

Inspection activities included review of previous surveillance history to identify previous problems and trends. observation of pre-evolution briefings, and initiation/resolution of related IRs for selected surveillances.

15 On October 20-25, 1303-11.3, Surveillance Test and Set Main Steam Safety Rev. 33. Six MSSVs did not lift at required setpoints, requiring adjustment expansion of test scope to additional valves in accordance with the 1ST (IR On November 2, operators performed procedure OP-TM-642-231 ES Train Emergency Sequence and Power Transfer Test, Rev. 2. BS-P-1A load exceedance of the allowable range was evaluated in IR 987993. In EF-P-2A load delay exceedance of the allowable range was evaluated in IR On December 14, OP-TM-541-251, Leakage Exam ofNR Underground Piping, 4, intermediate change IC-28533.

The inspectors also reviewed IR 1004914 documented minor discrepancies identified during testing; On December 15,1303-11.11, Station Battery Load Test, Rev. 338.

b. Findings

I I No findings of significance were identified.

I 2. RADIATION SAFETY Cornerstone:

Occupational Radiation Safety 20S1 Access Controls (71121.01

-19 samples)

a. Inspection Scope

The inspectors reviewed selected activities and associated documentation in the listed areas. The evaluation of Exelon's performance in these areas was against contained in 10 CFR 20, applicable TSs, and applicable Exelon Plant Walkdowns and Radiation Work Permit (RWP) Reviews (Jobs-in"progress Reviews) The inspectors walked down selected radiological controlled areas and housekeeping, material conditions, posting, barricading, and access controls radiological areas. The inspectors reviewed exposure significant work areas determine if radiological controls were acceptable and conducted selective radiation surveys with a survey instrument.

The inspectors walked down the steam generator storage The inspectors selectively reviewed radiological controls for outage work activities including refueling activities, spent fuel pool work activities, and on-going containment outage work activities that presented elevated radiological risk to workers including highest collective occupational doses. Aspects of steam generator replacement were also reviewed including primary piping cutting and welding, steam generator removal activities, and pipe end decontamination activities.

The inspectors attended selected High Radiation briefings for the decontamination activities (Le., core spray line) and observed on-going decontamination activities.

The reviews included selective evaluation of the adequacy of applied radiological controls including RWPs, procedure adherence, radiological surveys. job coverage, system breach surveys, airborne radioactivity sampling and controls, use of engineering controls, and contamination controls.

The reviews included barrier integrity and the application of englneering controls for potential airborne radioactivity areas and radioactive source term, and radiation levels present. The inspectors selectively attended and observed on..going radiological briefing activities and also attended radiological controls shift turnover meetings.

The inspectors reviewed applicable RWPs and electronic personnel dosimeter (EPD) alarm set points (both integrated dose and dose rate) to verify that the setwpoints were commensurate with ambient/expected conditions, RWPs, plant policy, and were appropriate for the conditions.

The inspectors discussed worker actions upon EPD alarms. The inspectors reviewed, and discussed TS High Radiation Areas (HRAs), including HRA controls for access to the Reactor Building.

The inspectors conducted a HRA key I audit. The inspectors reviewed control ofactivated or contaminated materials within I. water filled pools. As part of this review, the inspectors selectively reviewed fuel sipping activities.

The inspectors discussed controls for radiation dose rate gradients to verify that Exelon had applied appropriate radiological controls including use of multiple dosimeters or repositioning of dosimetry to accurately measure radiation doses. The inspectors also reviewed and discussed inter-comparison of electronic dosimeter and thermoluminescent dosimeter results to identify anomalies and licensee actions. The inspectors selectively reviewed personnel exposure investigations.

The inspectors reviewed and discussed internal dose assessments for 2009 and the outage to identify any apparent actual occupational internal doses greater than 50 millirem committed effective dose equivalent.

The review also included the adequacy of evaluation of selected dose assessments and included selected review of the program for evaluation of potential intakes associated with hard-to-detect radionuclides (e.g., airborne transuranics).

The inspectors selectively reviewed 2009 whole body counter logs and data. Problem Identification and Resolution The inspectors selectively reviewed selfwassessments and audits since the previous inspection to determine if identified problems were entered into the corrective action program for resolution.

The inspectors evaluated the database for repetitive deficiencies or significant individual deficiencies to determine jf self-assessment activities were identifying and addressing the deficiencies.

The review also included evaluation of data to determine if any problems involved Performance Indicator (PI) events with dose rates greater than 25 Rlhr at 30 centimeters, greater than 500 Rlhr at 1 meter or unintended exposures greater than 100 millirem total effective dose equivalent, 5 rem shallow dose equivalent, or1.5 rem lens dose equivalent.

The inspectors also reviewed the corrective action database for non-PI radiological incidents to determine if follow-up activities were being conducted in an effective and timely manner consistent with radiological risk. The review also included problem reports since the last inspection which involved potential radiation worker or radiation protection personnel errors to determine if there Enclosure was an observable pattern traceable to a similar cause. The review included an evaluation of corrective action (see Section 40A2). High Risk Significant, High Dose Rate HRA and Very HRA Controls The inspectors discussed procedure changes for HRA access controls since the last inspection with selected supervisors to determine if the changes resulted in a reduction in the effectiveness and.level of worker protection.

The inspectors conducted a selective review of HRA controls (e.g., adequate posting and locking of entrances).

The inspectors discussed controls for HRA and very high radiation areas (VHRAs)with radiation protection technicians.

The inspectors reviewed the access key inventory for HRA and VHRA access areas and conducted a key inventory.ln.addiUon, the inspectors discussed administrative controls for keys, including key inventory.

Radiation Worker/Radiation Protection Technician Performance and Radiation Protection Technician Proficiency The inspectors evaluated radiation protection technician performance and proficiency relative to control of hazards and work activities, as applicable.

In addition, the inspectors reviewed problem reports to identify problems with worker or radiation protection technician performance.

The inspectors selectively questioned both radiation workers and radiation protection personnel regarding on-going activities and knowledge of controls and conditions.

b. Findings

No findings of significance were identified.

On November 21, 2009, Exelon experienced an airborne radioactivity event within the Unit 1 Containment Building.

The inspection associated with that matter will be separately documented in NRC Inspection Report 05000289/2010007.

20S2 As Low As is Reasonably Achievable (ALARA) Planning and Controls (71121.02

-10 samples)

a. Inspection Scope

The inspectors conducted the following activities to determine if Exelon was properly implementing operational, engineering, and administrative controls to maintain personnel occupational radiation exposure ALARA. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, applicable industry standards, and applicable Exelon procedures.

Inspection Planning, Radiological Work Planning The inspectors reviewed pertinent information since the previous inspection regarding plant collective exposure history, current exposure trends, and ongoing and planned activities in order to assess current performance and exposure challenges.

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

The inspectors evaluated site specific trends in collective exposures (using NUREG-Enclosure 0713 and plant historical data) .. The inspectors evaluated occupational exposures received for 2009 (year-to-date), relative to applicable ALARA goals. . The inspectors selected work activities likely to result in the highest personnel collective exposures and selectively reviewed the planning and preparation for those work activities.

The inspectors evaluated the level of detail associated with projected dose estimation.

The inspectors reviewed the integration and implementation of ALARA requirements into procedures and RWP documents.

Job Site Inspections and ALARA Controls The inspectors toured throughout the facility and reviewed radiological conditions, use of shielding, and informational ALARA postings.

The inspectors reviewed ALARA planning and conduct of work associated with refueling activities, spent fuel pool activities, and outage work activities, including steam generator replacement activities.

The inspectors made independent radiation measurements to validate ambient conditions.

The inspectors determined if personnel were using lowdose wait areas to maintain their doses ALARA. The inspectors evaluated the level of supervisor oversight to determine if first-line supervisors conducted work in a dose efficient manner. The inspectors attended various radiological briefings to evaluate the scope and effectiveness of briefings for purposes of dose reduction.

The inspectors reviewed exposures of individuals from selected work groups to identify significant exposure variations which may exist among workers. Verification of Dose Estimates and Exposure Tracking The inspectors reviewed Exelon's method for adjusting exposure estimates or planning work when unexpected changes in scope, radiation levels, or emergent work were encountered to determine if the adjustments were based on sound radiation protection and ALARA prinCiples.

The inspectors selectively reviewed work-in-progress evaluations to determine reasons for increased dose accumulation (scaffolding, Alloy 600 work, lSI activities).

The inspectors also reviewed the frequency of these adjustments to evaluate the original AlARA planning process. The inspectors assessed whether work activity planning included consideration of the benefits of dose rate reduction activities, such as shielding provided by water filled components/piping, job scheduling, and scaffolding installation and removal activities.

Source-Term Reduction and Control The inspectors reviewed and discussed Exelon's understanding of the Unit 1 plant source term, including knowledge of input mechanisms to reduce the source term and the source term control strategy in place. The inspectors selectively reviewed Exelon's efforts to reduce areas of elevated dose rates (e.g., HRAs). The inspectors toured containment work areas to review implementation of occupational dose reduction initiatives, shielding efforts, and efforts by workers to minimize occupational exposure.

I 19 I I I Radiation Worker/Radiation Protection Technician Performance I The inspectors selectively observed radiation worker and radiation protection technician I performance in the area of ALARA practices to identify acceptable performance in areas of greatest radiological risk to workers. Declared Pregnant Workers I The inspectors selectively reviewed the declared pregnant worker program and I exposure control. I I Problem Identification and Resolution I The inspectors selectively reviewed problem reports in this area since the last inspection j to determine if Exelon was including ALARA deficiencies and issues in its corrective action program (see Section 40A2). The review included self-assessments, audits and corrective action reports related to ALARA program since the last inspection to determine if the follow-up activities being conducted in an effective and timely manner commensurate with their to safety and

b. Findings

No findings of significance were identified.

2083 gadiation Monitoring Instrumentation and Protective Eguipment (71121.03

-5 samples)

a. Inspection Scope

The inspectors selectively reviewed radiation monitoring/measurement in the below listed areas. The review was against criteria contained in applicable and station Self-Contained Breathing Apparatus The inspectors selectively reviewed the testing and certification of three breathing apparatus (Kits 1, 3, 20) including inspection data and hydrostatic testing. inspectors observed alarm testing for Kit Verification of Instrument Calibration, Operability.

and Alarm Set Point Verification The inspectors selectively reviewed radiological survey data for primary systems to identify radionuclides present to determine adequacy of surveying and monitoring techniques utilized.

The inspectors also reviewed the radiological controls aspects covered by 7'1121.02 that will be documented in NRC Inspection Report

.1 Problem Identification

and Resolution The inspectors reviewed problem reports in this area since the last inspection to determine if Exelon was including instrument deficiencies and issues in its corrective action program. The review included self*assessments, audits and corrective action reports (see Section 40A2).

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

40A 1 Performance Indicator Verification (71151) a. Inspection Scoge Cornerstone:

Mitigating Systems (6 samples) The inspectors reviewed Exelon's assessment of safety system functional failures (SSFFs) for the period September 2008 through September 2009. The inspectors verified accuracy of the reported data through review of selected station operating logs, system health reports, SSFF databases, and Licensee Event Reports. The inspectors also reviewed Exelon's assessment of mitigating systems performance indicators (MSPls) for the period October 2008 through September 2009. Verification included the n:wiew of selected definitions, data reporting elements, calculation methods, definition of terms, use of clarifying noles, Consolidated Data Entry MSPI Derivation Reports for unavailability and unreliability, monitored component demands, demand failure data, operator logs, maintenance rule database entries, and corrective action program documents.

Reviews were performed to determine whether associated PI data had been accurately reported to the NRC in accordance with NEI 99-02. Additional documents reviewed are listed in the Attachment.

The following Pis were evaluated:

• SSFFs

• High Pressure Safety Injection System (Makeup)

• Emergency Feedwater System

• Emergency AC Power System (Emergency Diesel Generators)

• Decay Heat Removal MSPI: Cooling Water Support Systems (Decay Closed, Decay River, Nuclear Closed, Nuclear River) Cornerstone:

Barrier Integrity (2 samples) The inspectors reviewed selected station records, corrective action program documents, calculation methods, and definitions of terms to verify NRC Pis had been acct,lrately reported to the NRC as required by Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 5 and 6. The PI sample listed below was verified for the period April 2008 to September 2009.

• Reactor Coolant System Activity Enclosure

-.1 The inspectors also reviewed selected station records including operating logs, surveillance test reports, and IRs, observed associated surveillance tests, conducted interviews with operators and engineers, and performed equipment walkdowns to assess reactor coolant system identified leak rate for the period September 2008 through September 2009. This review was performed to determine whether associated PI data had been accurately reported to the NRC in accordance with NEI 99-02. Additional documents reviewed are listed in the Attachment.

• Reactor Coolant System Identified Leak Rate Cornerstone:

Occupational Radiation Safety (1 sample) Implementation of the Occupational Exposure Control Effectiveness PI Program was reviewed.

The inspectors reviewed corrective action program records for occurrences involving HRAs, VHRAs, and unplanned personnel radiation exposures for the period October 2008 through September 2009. The inspectors reviewed individual radiation exposure results and selectively reviewed exposure records and associated RWPs. The review was against the applicable criteria specified in NEI 99 w 02. The purpose of this review was to verify that occurrences that met NEI 99-02 criteria were recognized and identified as PI occurrences.

Cornerstone:

Public Radiation Safety (1 sample) Implementation of the Radiological Effluents Technical Specifications/Offsite Dose Calculation Manual PI was reviewed.

The inspectors reviewed corrective action program records and projected monthly and quarterly dose assessment results due to radioactive liquid and gaseous effluent releases for the period October 2008 through September 2009. The inspectors selectively reviewed 2007/2008 projected exposure data. The inspectors also reviewed and discussed potential abnormal releases via groundwater or effluents.

The review was against the applicable criteria specified in NEI 99-02. The purpose of this review was to verify that occurrences that met NEI 99:02 criteria were recognized and identified as PI occurrences.

b. Findings

No findings of significance were identified.

40A2 Identification and Resolution of Problems (71152) Review of Issue Reports and Cross-References to Problem Identi'fication and Resolution Issues Reviewed Elsewhere The inspectors performed a daily screening of items entered into the licensee's corrective action program. This review was accomplished by reviewing a list of daily IRs, reviewing selected IRs, attending daily screening meetings, and accessing the licensee's computerized corrective action program database . . 2 Annual Sample -Configuration Controls (1 sample)

a. Inspection Scope

Enclosure

! I 22 ! I During the Fall 2007 refueling outage (1R17), licensee and contractor personnel several equipment configuration errors while reinstalling steam generator covers in preparation for raising RCS level from mid-loop to reactor vessel flange I The causes of the errors were primarily deficient work control and work coordination I between work groups. These errors resulted in unreliable reactor vessel level indication I and declaration of an Unusual Event. Additional equipment configuration issues noted I.f during the outage included reactor startup with the steam supply valves to the driven emergency feedwater pump inadvertently isolated.

These deficiencies were previously documented as NCV 05000289/2007005-05 and 05000289/2007005-03, respectively.

The inspectors reviewed Exelon's common evaluation and corrective actions associated with these NCVs. The inspectors reviewed corrective action program documents, design changes, procedure training documents, contractor oversight procedures.

and related industry experience.

The inspectors also intervIewed" statIon and contractor personnel to reasonable corrective actions were implemented to address the causes and reasonable assurance that reactor vessel level control and indioations would reliable during the Fall 2009 refueling outage. Additional documents reviewed are in the

b. Findings and Observations

No findings of significance were identified.

The licensee determined the principle of the violations were low work standards, unclear accountability, and lack of for work activities.

Corrective actions included a design modification for steam (SG) upper head inspection port ventilation to eliminate excessive pressure significant revision to 5G inspection port and manway cover removal and procedures, direct accountability for SG ventilation exhaust fans, revision to the water level control procedure, and training for appropriate licensee and contractor Training on the procedure revisions and lessons learned briefings were conducted for applicable station personnel and contractor staff. Configuration control was as a station focus area for 2009. Additional corrective actions to address configuration control errors were also implemented.

The inspectors determined problem evaluation.

extent-of-condition review, and scope of corrective actions reasonable to address the causes of the inaccurate RCS water level and mispositioning events identified during 1

.3 Annual Sample -Instrument

Calibration (1 sample)

a. Inspection Scope

ThiS inspectors reviewed Exelon's corrective actions (IRs 523284 and 525514) to address a previously NRC identified violation (NCV 05000289/2006005-01)untimely corrective actions for unreliable borated water storage tank level In addition, the inspectors interviewed operators and engineers and reviewed procedures.

IRs, and instrument calibration results and trends for the makeup system I and building spray system. The inspectors evaluated this information to verify ! reasonable corrective actions were implemented to address adverse instrument performance trends and provide reasonable assurance that instrument indications were accurate and reliable to support safe plant operation.

I i I I I 23 I

b. Findings

No findings of significance were identified.

Station personnel write issue reports required by procedure 1001J.1, Surveillance Testing Program, Rev. 9, for instrument found outside of the acceptance criteria during the calibration test.

use the IR process to screen the instrument's recent calibration history (last calibration cycles) to determine whether instrument reliability is a concern. If deemed reliability concern per guidance in procedure ER-AA-S20, Instrument Trending, Rev. 3, engineers initiate a more detailed review of instrument suitability affect on operability.

The inspectors determined that for most building spray makeup system instruments, appropriate monitoring and actions were initiated to reliable Notwithstanding, the inspectors identified that several cafibration deficiencies for makeup tank level and pressure instrumentation were not trended and appropriate corrective actions were not implemented in a timely Operators used these instruments to help support operability of several emergency cooling injection system: Instruments MU17-PT and MU-LT-778 were repeatedly outside of the procedural operability guideline settings over the last 8 Adjustments to as-found settings and reduction of the surveillance interval were effective in maintaining the instruments within their procedural acceptance Additionally, the inspectors determined that these calibration surveillance failures not properly coded and therefore were not captured in the ER-AA-520 biennial Monitoring Trend Report. The inspectors reviewed design basis calculations, specifications, and operator daily functional instrument comparison checks. functional checks verify close agreement between several makeup ta'nk level pressure instruments.

The inspectors determined that despite degraded performance, operators maintained makeup tank pressure and level within unrestricted operating region as specified in TS Figure 3.3-1. Therefore the of the degraded instrumentation was minor. Engineers entered the inspectors' into the corrective action program (IR I.4 Semi-Annual Review to Identify Trends (1 sample)

a. Inspection Scope

I I-The inspectors performed a semi-annual review of common cause issues in order idenUfy any unusual trends that might indicate the existence of a more significant issue. This review included an evaluation of repetitive issues identified via the I action program, self-revealing issues, and issues evaluated using supplemental to the formal corrective ac:tion program, such as the maintenance program and corrective maintenance program. The results of the trending review were compared with the results ofnormar baseline inspections.

b. Findings

No findings of significance were identified.

The inspectors determined that, overall, corrective actions to address performance deficiencies from the previous refueling outage were effective.

F(Jel handling was safely performed and coordinated.

For the OTSG replacement, the reactor building containment concretelliner barrier opening was, safely created and restored.

Industry operating experience was integrated in a timely Enclosure I I ! 24 I manner. Work hours and fatigue assessment were effectively managed. Very few incidents involving degraded nuclear or personnel safety occurred . . Identification and Resolution of Problems (71121.01, 71121.02, 71121.03,7112201, I 7112202,7112203,71151)1 Inspection Scope, I The inspectors selectively reviewed problem reports, self-assessments, and Nuclear Oversight (NOS Observations)to determine if identified problems were into the corrective action program for resolution.

The inspectors selectively reviewed reports to evaluate Exelon's threshold for identifying, evaluating, and resolving I The review included a check of possible repetitive issues, such as worker or errors (IRs 985046,985193,986383,987987,986289,991735, 988615, 992260,991757,991755,993726, I 587980,975430, 1003634,998566, 1001022).

This review was against criteria contained in 10 CFR20, Technical Specifications, and the station II . Findings No findings of significance were identified.

I, Identification and Resolution of Problems for Steam Generator Replacement Activities I (71152) I !

a. Inspection Scope

I The inspectors reviewed IRs associated with the SGR to ensure that Exelon was i identifying, evaluating, and correcting problems associated with these areas and that the I*I'planned or completed corrective actions for the issues were appropriate and completed r in a timely manner. The inspectors also reviewed a sample of Exelon's assessments related to 10 CFR 50.59 Safety Evaluations and plant modification activities at TMI. The listing of the IRs and self-assessments reviewed is provided in I Additionally, the inspectors reviewed those IRs initiated during the project issues related to the SGR These issues covered all phases of the project I prolect engineering.

steam generator fabrication and shop testing. steam i I transportation and receiving, temporary storage, steam generator lifting steam generator removal operations, piping insulation removal, pipe cutting, control procedures, and physical security.

The review of IRs included the initiation non-conformance reports by contractors and equipment vendors, as well as by I

b. Findings

No findings of significance were identified. Other Activities II . I

.1 Quarterly

Resident Inspectors Observations of Security Personnel 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 inspectors 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 Inspection Results for TI 2515/172, RCS Dissimilar Metal Butt Welds

a. Inspection Scope

Temporary Instruction (TO 2515/172 provides for confirmation that owners of pressurized-water reactors (PWR) have implemented the industry guidelines of the Materials Reliability Program-139 (MRP) regarding nondestructive examination and evaluation of certain dissimilar metal welds in reactor coolant systems and components with welds containing Alloy 600/82/182.

The TI requires documentation of specific questions in this inspection report. The questions and responses for the T1 R 18 outage are included in Attachment B of this report. This T1 R 18 report supplements the information provided in NRC Inspection Report 05000289/2008004 and it documents the completion of TI 2515/172.

The RPS data base has been updated to reflect the completion of this TI for the year 2009. In summary, TMI Unit 1 has 14 MRP-139 applicable Alloy 600/82/182 RCS welds. Those welds are:

• one 10" pressurizer surge line

• one 1 0" pressurizer hot leg surge nozzle

• one 4" pressurizer spray

• four 28" RCS cold leg (CL) outlet nozzles (at the RC pump);

• four 28" RCS CL inlet nozzles (at the RC pump);

• two 14" core flood nozzles (at the reactor pressure vessel); and

• one 12" hot leg decay heat drop line.

b. Findings

No findings of significance were identified . . 3 ,gteam Generator Replacement (50001)

a. Inspection Scope

The inspectors verified that engineering evaluations and design changes associated with SGR were completed in conformance with requirements in the facility license, the applicable codes and standards, licensing commitments.

and the regulations.

The inspectors reviewed applicable Engineering Change Requests (ECRs), which controlled 26 the replacement of the steam generators.

ECRs were reviewed in accordance with the requirements of IP 71111.02, Evaluation of Changes, Tests and Experiments and IP 71111.17B, Permanent Plant Modifications.

The inspectors reviewed the administrative procedures that were used to control the screening, preparation, and issuance of safety evaluations to ensure that the procedure adequately covered the requirements of 10 CFR 50.59. The evaluation of the large break loss of coolant accident (LOCA) analysis for the new steam generators (EOTSGs) was included in the scope of inspection.

The inspection included a review of various engineering and work activities associated with establishing a temporary containment access opening in the side of the containment shell for the SGR and its subsequent restoration.

Prior to beginning SGR activities the inspectors reviewed the rigging, lifting, and handling procedures used to move the old steam generators (OTSGs) from the containment and move the EOTSGs into place inside the containment The inspectors verified that Exelon appropriately analyzed and addressed the potential for damage to existing plant SSCs due to the moving of these heavy loads, including the transport . routes before performing these steam generator movements.

The component drop analysis was reviewed to verify that the potential offsite releases at the exclusion area boundary were within 10 CFR Part 100 limits and that equipment to maintain safe shutdown was unaffected.

The inspectors reviewed the ECRs and implementing Work Packages for the site haul path for movement of the replacement EOTSGs and the OTSGs. The review included the engineering evaluation of all potentially affected underground piping systems, structures, and electrical conduits.

The inspectors also reviewed the rigging and handling procedures for the OTSGs and the EOTSGs both inside and outside of containment.

Access into containment for the steam generator replacement was through a temporary construction opening created in the containment wall. The inspectors reviewed ECR TM 06-00816 associated with the steam generator replacement modification.

The inspectors performed a review of applicable procedures, work practices, and documentation of the steam generator replacement project (SGRP) to assess the control of safety related aspects associated with the major phases of the SG replacement.

This inspection was focused on SGRP activities that restore the pressure boundary of the RCS, secondary systems, and the affected containment systems, the exclusion of foreign materials, and plant modifications that could affect plant risk during the replacement activities or subsequent plant operation.

The inspectors also examined the steam generator supports and related piping system supports.

The inspectors performed a review of those procedures which governed field activities involving the cutting and machining of weld preparations at existing main steam (MS), feedwater (FW), and RCS connections on the EOTSGs. The inspectors examined welding procedures, welder qualifications, weld filler metal selection, non-destructive test procedures, examiner qualifications, acceptance criteria and test results for compliance with the requirements of the ASME Boiler and Pressure Vessel Code (Code), Sections IX and XI. The inspectors examined training and qualification records of selected personnel performing welding and non-destructive examination.

The inspectors selected a sample of work packages (WP) for review. The selection of these field planning and Enclosure I I 27 I work documents was based on their risk significance and represented the installation by welding of RCS, main stream (MS). and feedwater (FW) piping, fittings, and pipe supports.

The tracking and processing of the work planning and control documents was examined.

The inspectors verified that selected work documents which controlled specific work tasks were appropriate for the work being performed and contained provisions and 'hold points' to monitor weld progress, provide necessary in-process NDE. and to close out the activities at the completion of work.. Work activities inspected included the lifting, rigging and removal of the OTSGs out of the containment, and the placement of the EOTSGs into the containment.

Observations were made of welders using the automated process for welding of the RCS piping and completed welds were examined.

The as machined weld preparations on selected samples of MS, FW and RCS piping spools were examined and assessed for dimensional accuracy by comparison with specification requirements.

The inspectors observed the welding of various piping spools in the fabrication shop and the field fit-up, tack welding and automated welding of a sample of these components to their respective steam generator hot and cold leg nozzles. The inspectors observed the drilling activity of several bolt holes on the '8' EOTSG skirt flange to enlarge 30 of the 48 bolt holes to enable component and RCS piping fit-up installation.

The inspectors observed various work activities associated with creating a temporary construction access opening in the Side of the exterior containment wall and the containment wall restoration.

As part of operating experience review the inspectors Observed that the concrete appeared to be structurally sound, with no indication of delamination or other degradation.

The exposed surfaces of the containment liner plate were observed to be in good condition, with no signs of corrosion.

The inspectors observed tendon sheath V131 repair, manual shielded metal arc welding of the containment liner plate, NDE of construction opening steel liner plate field weld FW-171, and reviewed magnetic testing (MT) examination data records of the containment liner restoration field weld and vacuum box testing examination data record of the containment liner field weld. The inspectors observed preparations for and placement of concrete in the opening mockup test placement assembly.

For restoration of the temporary opening in the containment wall, the inspectors verified that the concrete was handled and tested properly, that quality control practices were being adequately followed, and confirmed that concrete work activities were performed in accordance with applicable Codes AC1318-63, Building Code Requirements for Structural Concrete, 1963 Edition; AWS D1.4, Structural Welding Code -Reinforcing Steel, 1998 Edition; and, ASME Section XI, 1992 Edition with 1992 Addenda. One field weld from each of the MS, FW, and RCS was selected by the inspectors for an in-depth review of fabrication work documentation, including the severance at the OTSG nozzle to the completion of re-attachment to the EOTSG including and final NDE testing of the new welds. The final NDE tests for the acceptance of these risk significant welds involved surface and radiographic testing. Additionally, ultrasonic testing (UT) was performed for these respective welds to establish pre-service baseline quality levels for comparison during subsequent ASME Section XI exams (inservice inspection).

The inspectors reviewed the test activities and inspection reports to assess the completeness of the examination processes.

The inspectors also conducted interviews

! Enclosure with design and field engineers, and with construction and examination personnel responsible for implementing these activities.

The inspectors reviewed the manufacturing records and primary and secondary shop fabrication hydrostatic testing records. The inspectors reviewed the non-destructive test results of the pre-service examinations of the EOTSG shell, nozzle and welds, and the baseline examination of the steam generator tubes. The inspectors also reviewed the SG Degradation Assessment developed in advance of the pre-service tube inspections performed with respect to the limited range of degradation mechanisms that may pertain to the as-built condition.

The inspectors reviewed the post-steam generator replacement NDE of selected new welds for the risk significant RCS, MS, and FW systems. The results of these inspections were satisfactory.

Additionally, the inspectors reviewed the pre-service service ultrasonic inspection baseline examinations performed on the new RCS, MS, and FW system piping welds. The inspectors reviewed the procedure for the post containment restoration reactor building integrated leak rate test at about 50.6 psig. discussed the test plan with responsible engineers, observed the test preparations, the test equipment, and selected portions of the test evolution.

The inspectors reviewed the planning and implementation of security considerations associated with relocation of vital and protected area barriers affected by the steam generator replacement activities. The inspectors also conducted a walkdown and review of security boundaries affected by the steam generator replacement project and conducted a review of Exelon's compensatory measures.

The inspectors observed searches and subsequent security seal verification of the steam generators prior to their introduction into the vital area.

b. Findings

No findings of significance were identified.

40A6 Meetings, Including Exit Exit Meeting Summary On January 15, 2010, the resident inspectors presented the inspection results to Mr. William Noll and other members of the TMI staff who acknowledged the findings.

The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary.

No proprietary information was identified.

40A7 Licensee Identified Violations The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NeV.

• Technical specification 6.S, 1.e requires that written procedures covering the Fire Enclosure 29 Protection Program shall be properly implemented.

Procedure AP-1038. Fire Protection Program, establishes TMI Unit 1 Fire Protection Program functions.

Procedure 1303-12.9, Fire Barrier Seal Inspection, requires that penetration seals be verified functional to ensure compliance with procedure 1038. Contrary to these requirements, inspections of the fire seals for the five safety-related 120 volt vital inverters were not performed.

In addition, fire seal inspections performed in July 2008 as part of prior extent-of-condition reviews for multiple degraded fire seals (IR 793088 and Green NCV 05000289/2008-004-01), did not identify two degraded fire seals for vital inverters

'C' and 'E'and did not initiate corrective measures.

Consequently, the seals which were damaged during inverter replacement activities in November 2007, did not meet the required 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating for a period of two years. However, engineers determined the degraded fire seals would have provided at least a one hour fire rating and due to defense-in-depth the degraded fire seals would not have affected operability of the vital inverters in the event of a fire. This finding adversely affected the reliability of equipment required to achieve and maintain a safe shutdown condition following a severe fire, because the degraded fire seals adversely affected the confinement defense-in-depth element of fire protection.

The finding is greater than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone.

This finding is of very low safety significance because the safe shutdown of the plant was not affected due to available defense-in-depth fire protection features such as incipient and ionization detectors, control room alarms, the plant's fire brigade team, and also because the degraded seals would have provided greater than a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire rating protection in the event of a fire. This issue was placed in Exelon's corrective action program as IR 995903. Corrective actions included extent-of-condition walkdowns, repair of the fire seals, and ensuring these seals are inspected every two years. I ATTACHMENT:

=SUPPLEMENTAL

INFORMATION=

I I ! I I I I Enclosure ! . I

Licensee Personnel

. D. Atherholt

R. Atkinson C. Baker R. Bleistine

T. Bradley J. Byrne W. Carsky G. Chevalier

R. Davis D. DeBoer D. DiVitore T. Dougherty

D. Etheridge

T. Geyer J. Heishman J. Karkoska R. Libra F. Linsenbach

W. McSorley A. Miller J. Murray G. Navratil D.Neff W.Noll J. Piazza T. Roberts J. Schork B. Swenson M. Sweigart W. Taylor D. Trostle L. Weber LWeir C. Wend Other D. Dyckman SUPPLEMENTAL

KEY POINTS OF Manager, Regulatory

Assurance

Manager, Steam Generator

Replacernent

Project Manager, Chemistry I\lormandeau Associates

Normandeau

Associates

Licensing

Director, Operations

Senior Chemist Manager, Radiation

Protection

Director, Operations

Manager. Radiological

Engineering

Plant Manager Manager. Radiation

Protection

Technical

Support Manager, Programs Director, Maintenance

Manager, Site Security Director, Work Management

Manager. OTSG Replacement

Radiation

Protection

Mechanical

Design Engineer Regulatory

Assurance

Manager, Operations

Training Engineer Manager, Emergency

Preparedness

Site Vice President

Senior Eng ineering Manager Supervisor, Radiation

Protection

Lead LORT Instructor

Vp, Projects Supervisor, RadwastefEnvironmental

Project Manager, SGT Operations

Security Analyst Senior Chemist Manager, Nuclear Oversight

Services Manager, Radiation

Protection

Nuclear Safety Specialist

Pennsylvania

Department

of Environmental

Protection

Bureau of Radiation

Protection

LIST OF ITEMS OPENED, CLOSED, AND

Closed URI 05000289/2008009-02

Potential

C02 Migration

Outside the Relay Room Fire Area (Section 1 R05.2) Opened and Closed NCV 05000289/2009005-01

Potential

C02 Migration

Outside the Relay Room Fire Area (Section 1 R05.2) LIST OF DOCUMENTS

REVIEWED Section 1R04: Equipment

Alignment

Procedures

1103-11, RCS Water Level Control, Rev. 66 1107-1, Normal Electrical

System, Rev. 78 1107 -2A, Emergency

Electrical-

4KV and 480 volt, Rev. 19 OP-TM-212-000, Decay Heat Removal System, Rev. 11 OP-TM-212-111, Shifting

DH Train A from DHR Standby To DHR Operating

Mode, Rev. 5 OP-TM-533-000, Decay Heat River System, Rev. 9C, Interim Change 26967 OP-TM-543-000, Decay Heat Closed System, Rev. 8 OP-TM-543-101, Shifting DC Train A from Standby to DHR Operating

Mode, Rev. 1 OP-TM-731-525.

De-Energizing

1S 480V Switchgear.

Rev. 3 CC-AA-112, Temporary

Configuration

Changes, Rev. 15 Drawings 308-946, RCS Benchmark

Elevations.

Rev. 3 302-640, Decay Heat Removal, Rev. 82 302-651, Reactor Coolant System, Rev. 57 302-630, Spent Fuel Cooling System, Rev. 31 Other TMI-1 Technical Specification 3.8 AR-A2075238 (WO-R2044279).

EE-MCC-ES-1B-6K, Install Temporary

Power SF-P-1B Section 1 R05: Fire Protection

Procedures

OP-TM-EOP-020, Cooldown from Outside of Control Room, Rev. 9 and 10 OP-TM-AOP-001, Fire, Rev. 4 and 6 OP-TM-EOP-0201, Cooldown from Outside of Control Room Basis Document, Rev. 5 Drawings 1-FHA-035, Fire Area Layout Control Room Tower, Elevation

2, Rev. 13 1-FHA-036.

Fire Area Layout Control Room Tower, Elevation

355, Rev. 14 1-FHA-037.

Fire Area Layout Control Room Tower, Section A-A, Rev. 6 1-FHA-038, Fire Area Layout Control Room Tower, Section B-6, Rev. 6 Attachment

TM-08-00963, C02 Discharge

into Control Building Relay Room, Rev.

App R Control Room Evacuation

Transient

Analysis, Rev. Issue Reports 815641 and Section 1 R013: Maintenance

Risk Assessment

and Emergent Work

1103-11, RCS Water Level Control, Rev. OP-TM-220-567, Drain Cold Leg(s) to RCBT Via RC Drain Pump, Rev.

308-946, RCS Benchmark

Elevations

Schematic

Layout, Rev. Section 1 R19: Post Maintenance

Testing

OP-TM-642-231, ES Train Emergency

Sequence and Power Transfer Test, Rev.

208-207, Electrical

Elementary

Diagram, Reactor Building Spray Pump BS-P-1A, Rev. TMI Updated Final Safety Analysis SDBD-T1-214, Rev. Issue 982522 983373 983631 983658 983712 983741 984012 984026 987787 987'922 Section 1 R20: Refueling

and Other Outage Activities ,I'

1103-11, RCS Water Level Control, Rev. 1505-1, Fuel and Control Component

Shuffles, Rev. 1507-3, Main Fuel Handling Bridge Operating

Instructions, Rev. 1507 -5, Spent Fuel Handling Bridge Operating

Instructions, Rev. MA-AA-716-008-1008, Reactor Services Refuel Floor FME Plan, Rev. OP-TM-642-231, ES Train 'N Emergency

Sequence and Power Transfer Test, Rev. OU-AP-4001, PWR Fuel Handling Practices, Rev. Section 20S1: Access Controls, Section 20S2: ALARA Planning and Controls.

and 20S3: Radiation

Monitoring

Instrumentation

and Protective

Equipment

Self-contained

breathing

apparatus

testing/inspection

results NOS Outage Observations-

2009 Various 2009 radiation

monitor calibration

and operability

check data Various 2009 radiological

survey records for work activities

including

records Various 2009 radiation

work permits for work activities

and associated

ALARA plans Various 2009 personnel

whole body count data results, personnel

exposure investigation

logs Radiological

controls contamination

data Various 2009 Station ALARA Committee

Meeting Minutes Attachment

Section 40A1: Performance

Indicator

LS-AA-2001, Collecting

and Reporting

of NRC Performance

Indicator

Data, Rev.

Mitigating

System Performance

Index Data Acquisition

& Reporting, Rev. Other

TMI-2006-004, MSPI Basis Document, Rev. TMI-2006-004, MSPI Basis Document, Rev. NE199-02, Regulatory

Assessment

Performance

Indicator

Guideline, Rev, NEI 99-02. Regulatory

Assessment

Performance

Indicator

Guideline.

Rev, TMI Unit 1 Plant Operatiqns

Review Committee

Meeting 2009-14 NRC PI

Section 40A2: Identification

and Resolution

of Problems Procedures

1001 J.1 , Surveillance

Testing Program, Rev. 9 1103-11, RCS Water Level Control, Rev. 65 1302-5.17. Makeup Tank Level & Pressure Instrumentation, Rev. 25 1302-14.1, Calibration

of 1ST Related Instruments, Rev. 63 1401-4.4C , Remove -Install OTSG Upper Head Inspection

Port Cover, Interim Change 28224 AD-AA-2110, Management

and Oversight

of Supplemental

Workforce, Rev. 6 OP-AA-1 08-112, Plant Status and Configuration.

Rev. 4 OP-TM-211-000, Makeup and Purification

System, Rev, 19 Drawings S09-0212-001, Pressure Relief Damper General Assembly, Rev. B Other Documents

Issue Reports 689443 698291 699314 739315 801459 931968 978376 797990 523284 640134 798088 798071 939840 940189 549376 267630 829380 515161 Work Orders C2020778 R2115356 . R2124339 R2058605 R2027331 R2008699 R 1801613 C-1101-624-E510-008, TMI-1 Makeup Tank Pressure Error (MU17-PT), Rev. 1 C-1101-624-5350-002, Makeup Tank Level (MU14-LT & MU-LT-778)

-Loop Error & Baseline Calibration, Rev, 6 NRC Information Notice 2009-11, Configuration

Control Errors. dated July 7,2009 NRC Information Notice 2009-22, Recent Human Performance

Issues at Nuclear Power Plants, dated October 2, 2009 Section 1 R08: Inservice

Inspection, Section 4OA5: Inspection

Results for TI 2515/172 ReS Dissimilar

Metal Butt Welds, and Section 40A5: Steam Generator

Replacement

Engineering

Change Requests (ECRs) TM 06-00709, OTSG Replacement

-FW and EFW Piping and Temp Supports.

Rev. 0 TM 06-00816.

OTSG Replacement

-Containment

Structural

Opening, Attachment

15, Rev. a TM 06-00935, Steam Generator

Component, Rev. 2 Attachment

1\ A-5 TM 07-00576, Hot Leg Piping and RTD Replacement, Rev. 1 TM 06-00759, RCS Piping & Temporary

Supports, Rev. 1 10 CFR 50.59 Safety

Safety Evaluation-000224-030, 50.59 Evaluation

for TMI-1 Once Through Steam

Replacement, Rev. 1 10 CFR 50.59 Screening

ECR 06-00935, TMI-1 Once Through Steam Generator

Replacement, Rev. Calculations

2-9121663, EOTSG Tube-to-Tubesheet

Weld Qualification

for LBLOCA Load-TMI1 and AN01, Rev. 1 32-9018796-003, Support Skirt Stress Analysis for TMI1 EOTSG, Rev. 3 38455-CALC-C-018, Postulated

Load Drop Assessments, Rev. 1 Calibration

Certificates

SGT003-09-08-24585-9, Calibration

Certificate

for Infrared Thermometer

SGT-039/S/N 0015, dated 8/25/09 AREVA Condition

Reports CR-2009-4602

Issue Reports (for SGRP) 934670,936594,982894,985146,990432,990534,993030,997062,998734,1003668, 1005356 Issue Reports (For lSI) 981777,984237,984792,989092,989737,989754,990534,990602, 991197,991611, 991616,991618.992274,994594,996574,997631.

Issue Report 998789 Drawings 007-TMI-DT-09-602-05, Steam Generators

Transport

Arrangement

-From Reconfiguration

Area to Plant 14 axles + 12 axles, dated 7/16/09 38455-00816-001, Containment

Opening Liner Plate Restoration

Elevation

& Details Installation, Rev. 8 NDE MT-FW-0036, MT of Pipe to Elbow Weld 422/FW36, dated 11112/09 (WO

UT-FW-0036, UTof Pipe to Elbow Weld 4221FW36, dated 11/10/09 (WO

PT -RCS H/L Riser Pipe Weld Prep Area, dated 12/6109

PT-RCS H/L Assembly Nozzle End, dated 12/11/09

MT Examination

Report MTS/12..:16-09/02, Weld on Liner Plate FW-171 (1 st Layer), 12/16/2009 (WP-3702)

MT Examination

Report MF 12-17-09/4, FW-171 Liner Plate (2 nd Layer), dated 12/23/2009

VB-LP-SGT, Rev. 0, Attachment

1, Liner Plate Bubble Test Report Seam Weld FW-171 , dated 12/18/2009

Steam Generator

Tube Inspections

AREVA Engineering

Report 51*9102157,Technical

Summary for Eddy Current Testing ofTMI-1 EOTSG Tubing, Rev. 1 AREVA Information

Record 51-9067993, TMI-1 EOTSG Pre-Service

Inspection

Degradation

Assessment, Rev. 2 Work Packages WP 1720, Construction

Opening Steel Liner Removal, Rev. 0 WP 2525A, Remove Lower Supports SG-A, Rev. 0 WP 3055A, Reinstall

Lower Supports SG-A, Rev. 0 WP 3530A, Remove RCS Temporary

Supports/Restraints

SG A, Rev. 0 WP 3720, Construction

Opening Steel Liner Installation, Rev. 0 WP 3040A, EOTSG "AU Installation, Rev. 0 WP 3040B, EOTSG "B" Installation, Rev. 0 Specifications

38455-SPEC-C-004, Reactor Buildill9

-Construction

Opening Concrete, Rev. 11 QEP 10.05, Rigging and Handling, Rev. 1 Exelon Specification

SP-1101-12-162.

Conformed

SpeCification

for Replacement

Steam Generators

for Three Mire Island Unit 1, Rev. 0 Welding Procedure

Specifications

and Procedure

Qualification

Records GTM/1.3-4A, ASME Section IX Welding Procedure

SpeCification, Rev. 0 Welder Qualification/Certification

For Welders P0350, 07345, G4369, W2766, 6-7251, B-5529. "IDE Examination

Test Reports C2020536-CFA-NDE-280, Core Flood Dissimilar

Metal Weld UT Data Sheet (Pre-Onlay), Core Flood Nozzle 1A to Safe End, dated 11124/09 Work Orders WO C2019694, Replace degraded 30" diameter pipe upstream of valves NR-V-3 and NR-V-2, dated 12/2/08 Procedurl3s

54-IS/-835-12, Ultrasonic

Examination

of Ferritic Piping Welds, Rev. 12 ER-AP-331, Boric Acid Corrosion

Control Program, Rev. 4 ER-AP-331-1001, Boric Acid Corrosion

Control Inspection

Locations, Implementation

and Inspection

Guidelines, Rev. 4 ER-AP-331-1002, Boric Acid Corrosion

Control Program Identification, Screening, and Evaluation, Rev. 5

Boric Acid Corrosion

Control Training and Qualification, Rev. 3 ER-AA-335-003, Magnetic Particle Examination, Rev. 3 ER-AA-335-018, Detailed, General, VT-1, VT-1C, VT-3 and VT-3C Visual Examination

ofASME Class MC and CC Containment

Surfaces and Components, Rev. 5 54-ISI-823-01, 10 Automated

Ultrasonic

Examination

of Dissimilar

Metal Core Flood Piping Welds, Rev. 1 ER-AP-330-1

001, Alloy 600 Management

Plan, Rev. 0 ER-AP-335-001, Bare Metal Visual Examination

for Alloy 600/821182

Materia/s, Rev. 1 Attachment

j" i I

Welding Procedure

Specifications

55-WP3/8/43/F43TBSC3, Machine Temper Bead Gas Tungsten Arc Welding Rev. 1 55-WP3f8/F60L

TB3. Machine Temper Bead Overlay, Gas Tungsten Arc Welding, Rev.O Other 33-9018765, Three Mile Island Unit 1 Replacement

Steam Generator, Rev. 0 51-9037616, Conduct of Operations

between AREVA NP and Exelon for TMI-1 EOTSG Project. dated 9/28/07. Rev. 4 51-9007383, TMI-1 EOTSG LOCA Safety Evaluation, Rev. 3 51-9124284.

LBLOCA -EOTSG Degraded Tube Assessment.

Rev. 1 51-9125139, Summary Report for Qualification

of EOTSG for LBLOCA Loading. Rev. 1 55-010035-03, Operating

Instruction.

RCS Welding for B & W Unit SG Replacement

Project, dated 3/17/04 MPR Associates

Independent

Third Party Review of EOTSG Documents.

dated 12/16/09 ADAMS ALARA . ASME CF CFR CL C02 DRP ECCS ECR EDG EOTSG

EPD ESAS FW

LIST OF ACRONYMS Agencywide

Documents

and Management

System As Low As is ReC\sonably

Achievable

American Society of Mechanical

Engineers

Core Flood Code of Federal Regulations

Cold Leg Carbon Dioxide Division of Reactor Projects Emergency

Core Cooling System Engineering

Change Request Emergency

Diesel Generator

Enhanced Once-Through

Steam Generator

Electronic

Personnel

Dosimeter

Engineered

Safeguards

and Actuation

System Feedwater

High Pressure Injection

High Radiation

Area . Inspection

Manual Chapter Issue Report Inservice

Inspection

Inservice

Testing Loss of Coolant Accident Low Pressure Injection

Maintenance

Rule Materials

Reliability

Program Main Steam Mitigating

System Performance

Indicator

Main Steam Safety Valve Magnetic Testing Non-cited

Violation

Nondestructive

Examination

Nuclear Energy Institute

NRC OTSG PADEP PARS POI PI PMT PT PWR RCS RWP SOP SG SGR SI sse SSFF TI TMI TS UFSAR URI UT VAC VHRA VT WO WP 1R17 T1R18 T1R19 A-8 I I I I Nuclear Regulatory

Commission

I Once Through Steam Generator

I Pennsylvania

Department

of Environmental

Protection

Publicly Available

Records I Performance

Demonstration

Initiative

l. Performance

Indicator

Post-Maintenance

Test Penetrant

Testing Pressurized

Water Reactor Reactor Coolant System Radiation

Work Permit I Significance

Determination

Process Steam Generator

Steam Generator

Replacement

Stress Improvement

Structures, Systems, and Components

Safety System Functional

Failures Temporary

Instruction

Three Mile Island, Unit 1 Technical

Specifications

Updated Final Safety Analysis Report Unresolved

Item Ultrasonic

Testing Volts Alternating

Current Very High Radiation

Area Visual Testing Work Order Work

Fall 2007 17 Refueling

Outage Fall 2009 18 th Refueling

Outage Scheduled

Fall 2011 Refueling

Outage Attachment

B-1 ATTACHMENT

B-1 TI 2515/172 Documentation

Questions

for TMI Unit 1 Introduction:

Temporary

Instruction (TI) 2515/172 provides for confirmation

that owners of pressurized-water

reactors have implemented

the industry guidelines

of the Materials

Reliability

Program 139 regarding

nondestructive

examination

and evaluation

of certain dissimilar

metal welds in reactor coolant systems (RCS) containing

Alloy 600/82/182.

The TI requires documentation

of specific questions

in an inspection

report. The questions

and responses

are included in this Attachment

B. In summary, Three Mile Island (TM!) has one four inch spray nozzle which is currently

placed in category '0' and 'J' and is planned for inspection

or mitigation

by application

of a structural

weld overlay in 2011. TMI also has two 14 inch core flood nozzles which were ultrasonically

examined (Performance

Demonstration

Initiative

[POI] Qualified)

in the T1R18. One, core flood (CF) nozzle 'A', dissimilar

weld was mitigated

in T1 R18 with the application

of a protective

barrier of Alloy 52 to the inside diameter of the nozzle (weld onlay) placing it in MRP-139 PWSCC category 'A'. The second, CF nozzle 'B', was mitigated

in T1 R18 with the application

of a protective

barrier of Alloys 82 and 52 to the inside diameter of the nozzle (weld onlay) placing it in PWSCC category 'E', TMI has four 28 inch CL RC pump suction and four 28 inch CL RC pump discharge

welds. There are no plans (or commitments)

to perform mitigation

activities

on these welds. A bare metal visual examination

and ultrasonic

examination (using POI qualified

techniques)

was completed

during the 1TR18. The examinations

were acceptable

and 100% coverage was achieved for each weld. TMI has two 10 inch surge line nozzles (one directly attached to the pressurizer

and the second attached to the 36 inch carbon steel hot leg). Both of these dissimilar

metal welds were mitigated

with the application

of a structural

weld overlay in the Fall of 2007 (1R17) and the Fall of 2003 (1R15). respectively.

Also, TMI has one 12 inch hot leg decay heat nozzle which was mitigated

during the 1 R17 with the application

of a structural

weld overlay. The EPRI Materials

Reliability

Program has issued Interim Guidance on November 1, 2007 regarding

dissimilar

metal butt welds of <4" (but> 1 "), Volumetric

Exam Requirements (Mandatory

Element).

The MRP guidance provided that dissimilar

metal butt welds greater than or equal to 2" NPS in the following

service conditions

and not already included within the volumetric

examination

requirements

of MRP-139 should be added: * those at pressurizer

temperatures;

• those athot leg temperatures;

and * those that serve an ECCS functioh.

TMI has identified

four nozzles in this classification

that are two and one-half inch in diameter and perform an ECCS function.

One of the welds in one nozzle had been previously

mitigated

in 2005 due to a replacement

of an internal thermal sleeve Attachment

",'" B-2 ,"" (unrelated

to dissimilar

metal weld). All eight welds were replaced in the T1 R18 and have been incorporated

into the ASME Section XI lSI program. a. For MRP-139 baseline inspections:

Qa1. Have the baseline inspections

been performed

or are they scheduled

to be performed

in accordance

with MRP-139 guidance?

A. Yes. Baseline manual ultrasonic

test (UT-POI qualified)

inspections

have been performed

on the pressurizer

surge nozzle, pressurizer

hot leg surge nozzle. hot leg decay heat drop line weld and one high pressure injection (HPI) nozzle. The four remaining

H PI nozzles were replaced in the T1 R 18 and their "category" has been changed to U A." The pressurizer

spray nozzle was UT examined (POI qualified)

during the Fall 2007 but has not yet been mitigated.

The spray nozzle weld will be mitigated (weld overlay) in the Fall of 201" and will be MRP-139 UT (POI qualified)

examined at that time. The eight cold legs (four inlets and four outlets) RCS suction and discharge

nozzle welds

UT examined (POI qualified)

in the T1 Ri8. One CF nozzle 'A', dissimilar

weld was mitigated

during T1 R18 with the application

of a protective

barrier of Alloy 52 to the inside diameter of the nozzle (weld onlay) placing it in MRP-139 PWSCC category "A". The second, CF nozzle 'B', was partially

mitigated

in T1R18 with the application

of a protective

barrier of Alloys 82 and 52 to the inside diameter of the nozzle (weld onlay) placing it in PWSCC category "E" with the MRP-139 baseline examination (UT-POI qualified)

performed

on both CF welds 'A' and 'B'. Oa2. Is the licensee planning to take deviations

from the MRP-139 baseline inspection

requirements

of MRP-139? If so, what deviations

are planned and what is the general basis for the deviation?

If inspectors

determine

that a licensee is planning to deviate from any MRP-139 baseline inspection

requirements, NRR should be informed bye-mail as soon as possible. ,A.. No MRP-139 deviations

have been taken at TMI and none are planned. b. For each examination

inspected, was the activity:

Ob1. Performed

in accordance

with the examination

guidelines

in MRP-139. Section 5.1 for unmitigated

welds or mechanical

stress improved welds and consistent

with NRC staff relief request authorization

for weld overlaid welds? A. Yes. Qb2. Performed

by qualified

personnel?

Briefly describe the personnel

training/qualification

process used by the licensee for this activity.

A. Yes. The examinations

were performed

by personnel

qualified

to the requirements

of ASME Section XI, Appendix VIII. Qb3. Performed

such that deficiencies

were identified, dispositioned, and resolved.

• c;* A. Yes. Indications

identified

in the ultrasonic

examination

were evaluated

for relevance, characterized

and entered into the licensee's

corrective

action process for disposition

and resolution.

c. For each weld overlay inspected, was the activity:

Qc1. Performed

in accordance

with ASME Code welding requirements

and consistent

with NRC staff relief request authorizations?

Has the licensee submitted

a relief request and obtained NRR staff authorization

to install the weld overlays?

A. No weld overlays were used in the Fall 2009 outage. Qc2. Performed

by qualified

personnel? (Briefly describe the personnel

training/qualification

process used by the licensee for this activity).

A. N/A Qc3. Performed

such that deficiencies

were identified.

dispositioned, and resolved?

A. N/A d. For each mechanical

stress improvement (SI) used by the licensee during the outage, was the activity performed

in accordance

with a documented

qualification

report for stress improvement

processes

and in accordance

with demonstrated

procedures?

.§p.ecifically:

Qd1. Are the nozzle. weld. safe end. and pipe configurations.

as applicable, consistent

with the configuration

addressed

in the SI qualification

report? .

A. N/A, the mechanical

stress improvement

process was not used. Qd2. Does the SI qualification

report address the location radial loading is applied, the applied load, and the effect that plastic deformation

of the pipe configuration

may have on the abHity to conduct volumetric

examinations?

A. NfA Qd3. Do the licensee's

inspection

procedure

records document that a volumetric

examination

per the ASME Code,Section XI, Appendix VIII was performed

prior to and after the application

of the SI? A. N/A Qd4. Does the SI qualification

report address limiting flaw sizes that may be found during pre-SI and post-SI inspections

and that*any flaws identified

during the volumetric

examination

are to be within the limiting flaw sizes established

by the SI qualification

report? A. N/A Od5. Performed

such that deficiencies

were identified, dispositioned, and resolved?

A N/A Attachment

'. e. For the inservice

inspection

program: Qe1. Has the licensee prepared an MRP-139 lSI program? If not, briefly summarize

the licensee's

basis for not having a documented

program and when the licensee plans to complete preparation

of the program. AYes. The licensee has an MRP-139 lSI program which is implemented

through ER-AP-330-1001

Rev. 0, Alloy 600 Management

Plan. The Alloy 600 Management

Plan program defines the processes, objectives

and key elements for maintaining

the integrity

and operability

of each alloy 600/82/182

component

for the remaining

life of the plant. This program makes provision

for the incorporation

of MRP-139 designated

welds into the existing ASME Section XI lSI program upon completion

of the specified

MRP-139 mitigation

activities.

This program provides the basis to support management

strategies

needed to address technical

operating

experience

with all Alloy 600/82/182

pressure boundary butt welds including

materials, commitments, remediation, inspection, and regulatory

requirements.

The subject welds will be included in the risk-informed

lSI program upon completion

of the remediation

plan for TMI Unit 1 Alloy 600. Oe2. In the MRP-139 Inservice

Inspection

Program, are the welds appropriately

categorized

in accordance

with MRP-139? If any welds are not appropriately

categorized, briefly explain the discrepancies.

A Yes. All 14 welds identified

during this inspection

are appropriately

categorized

in accordance

with MRP-139. Qe3. In the MRP-139 lSI Program, are there lSI frequencies, which may differ between the first and second 1 O-year intervals

after the MRP-139 baseline inspection, consistent

w1th the lSI frequencies

called for by MRP-139? AAII MRP-139 applicable

welds were mitigated

and/or inspected

during the Fall 2009 outage. Qe4. If any welds are categorized

as H or I, briefly explain the licensee's

basis for the categorization

and the licensee's

plans for addressing

potential

PWSCC. A There are three SI welds at TMI that are categorized

as 'I' and 'K'. Due to restrictive

access and unacceptable

surface profile, ultrasonic

testing cannot be performed

such that the required examination

volume will be achieved.

These three welds were replaced in the T1R18. Oe5. If the licensee is planning to take deviations

from the lSI requirements

of 139, what are the deviations

and what are the general bases for the deviations?

Was the NEI 03-08 process for filing deviations

followed?

A.No deviations

are currently

planned from the lSI requirements

of MRP-139 at TMI. Attachment