three mile island, unit 1, technical specification change ... · radiation monitor rm-g9 is an area...

AmerQKbG SM

AmerGen Energy Company, LLC www.exeloncorp.com An Exelon Company20o Exelon Way

Kennett Square, PA 19348 10 CFR 50.90

December 12, 20065928-06-20498

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

Three Mile Island Nuclear Station, Unit 1Facility Operating License No. DPR-50NRC Docket No. 50-289

Subject: Technical Specification Change Request No. 333 - Relocation of TechnicalSpecification Requirements for Refuel and Spent Fuel Pool Area RadiationMonitors

In accordance with 10 CFR 50.90, "Application for amendment of license or constructionpermit," AmerGen Energy Company, LLC (AmerGen) proposes changes to Appendix A,Technical Specifications (TS), of the Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1)Facility Operating License.

The proposed changes would revise the TMI Unit 1 TS to relocate the reactor building refuelingarea and spent fuel storage area radiation monitor operability requirements to the Updated FinalSafety Analysis Report (UFSAR) and plant procedures, since these radiation monitors do notmeet the criteria for inclusion in the TS as presented in 10 CFR 50.36(c)(2)(ii). These changesare consistent with Standard Technical Specifications (NUREG-1430, Revision 3). To furthersupport the proposed change, the current TMI Unit 1 Fuel Handling Accident in the FuelHandling Building, described in UFSAR Section 14.2.2.1 .b.1, has been reanalyzed withoutcredit for Fuel Handling Building ventilation exhaust filtration. This reanalysis is provided forNRC review and approval in Enclosure 4.

The proposed amendment has been reviewed by the TMI Unit 1 Plant Operations ReviewCommittee and approved by the Nuclear Safety Review Board in accordance with therequirements of the AmerGen Quality Assurance Program.

Using the standards in 10 CFR 50.92, AmerGen has concluded that these proposed changesdo not constitute a significant hazards consideration, as described in the enclosed analysisperformed in accordance with 10 CFR 50.91 (a)(1). Pursuant to 10 CFR 50.91 (b)(1), a copy ofthis Technical Specification Change Request is provided to the designated official of theCommonwealth of Pennsylvania, Bureau of Radiation Protection, as well as the chief executivesof the township and county in which the facility is located.

U.S. Nuclear Regulatory CommissionDecember 12, 2006

We request approval of the proposed change by September 30, 2007, with the amendment beingimplemented within 30 days of issuance. This will allow an orderly implementation of thesechanges prior to the Fall 2007 refueling outage (1 R1 7) for TMI Unit 1.

Regulatory commitments established by this submittal are identified in Enclosure 3. If you haveany questions or require additional information, please contact David J. Distel at (610) 765-5517.

1 declare under penalty of perjury that the foregoing is true and correct. Executed on the Ikday

of December, 2006.

Respectfully,

Pamela B. CowanDirector - Licensing and Regulatory AffairsAmerGen Energy Company, LLC

Enclosures: 1) TMI Unit 1 Technical Specification Change Request No. 333 - Description andAssessment

2) TMI Unit 1 Technical Specification Change Request No. 333 - Markup ofProposed Technical Specification and Bases Page Changes

3) List of Commitments4) TMI Unit 1 Calculation No. C-1 101 -900-E000-083, Revision 4, "EAB, LPZ, and

CR Doses Due to Fuel Handling Accidents"

cc: S. J. Collins, Administrator, USNRC Region ID. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1F. E. Saba, USNRC Project Manager, TMI Unit 1D. Allard, Director, Bureau of Radiation Protection - Pennsylvania Department

of Environmental ProtectionChairman, Board of County Commissioners of Dauphin County, PAChairman, Board of Supervisors of Londonderry Township, Dauphin County, PATMI Unit 1 File No. 06041

ENCLOSURE 1

TMI Unit 1 Technical Specification Change Request No. 333

Relocation of Technical Specification Requirements for Refuel and Spent Fuel Pool AreaRadiation Monitors

Description and Assessment

Enclosure 1Description and Assessment

Page 1 of 10

ENCLOSURE 1

DESCRIPTION AND ASSESSMENT

1.0 INTRODUCTION

In accordance with 10 CFR 50.90, "Application for amendment of license or constructionpermit," AmerGen Energy Company, LLC (AmerGen) is requesting an amendment toFacility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1(TMI Unit 1). The proposed amendment would revise the TMI Unit 1 TechnicalSpecifications (TS) to relocate the reactor building refueling area and spent fuel storagearea radiation monitor operability requirements to the Updated Final Safety AnalysisReport (UFSAR) and plant procedures, since these radiation monitors do not meet thecriteria for inclusion in the TS as presented in 10 CFR 50.36(c)(2)(ii). These changes areconsistent with Standard Technical Specifications (NUREG-1430, Revision 3).

AmerGen requests that the following changed replacement pages be inserted into theexisting Technical Specifications:

Revised TMI Unit 1 TS Pages: 3-44, 3-45, and 4-5a.

2.0 DESCRIPTION OF PROPOSED AMENDMENT

2.1 Revise TMI Unit 1 TS 3.8.1 and associated Bases to delete the specification requirementsstated below:

"Radiation levels in the Reactor Building refueling area shall be monitored by RM-G6 andRM-G7. Radiation levels in the spent fuel storage area shall be monitored by RM-G9. Ifany of these instruments become inoperable, portable survey instrumentation, having theappropriate ranges and sensitivity to fully protect individuals involved in refuelingoperation, shall be used until the permanent instrumentation is returned to service."

2.2 Revise TMI Unit 1 TS Table 4.1-1, Item 28, Radiation Monitoring Systems, to delete thefollowing sub-items and associated surveillance requirements: "a. RM-G6 (FH Bridge #1Aux)", "b. RM-G7 (FH Bridge #2 Main)", and "c. RM-G9 (FH Bridge-FH Bldg)". Table 4.1-1, Item 28, Remarks column Notes (2) and (3) only apply to RM-G6, RM-G7, or RM-G9and thus are being deleted.

3.0 BACKGROUND

Radiation Monitors RM-G6, RM-G7, and RM-G9 are designed to provide refueling andspent fuel pool area radiation monitoring for personnel protection during fuel loading andrefueling operations. These channels monitor radiation levels in the associated fuelhandling areas.


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Radiation Monitors RM-G6 and RM-G7 are area gamma monitors currently located on theReactor Building fuel handling bridges. RM-G6 is located on the Auxiliary Bridge atapproximately elevation 346 ft. in the Reactor Building. RM-G7 is located on the MainBridge at approximately elevation 346 ft. in the Reactor Building. Although isolation of theReactor Building is not credited in the fuel handling accident analysis for TMI Unit 1, asdescribed in existing UFSAR Section 14.2.2.1 .b.2, these monitors alarm any excessiveradiation in the vicinity of the refueling water surface and provide the Control Roomoperators sufficient information to initiate evacuation and closure of the Reactor Building inthe event of a fuel handling accident. These functions are not affected by the proposedchange.

Radiation Monitor RM-G9 is an area gamma monitor currently located in the Fuel HandlingBuilding (FHB) on the East Wall in the vicinity of the spent fuel pool water surface atapproximately elevation 346 ft.

As specified in the existing TS, if RM-G6, RM-G7, or RM-G9 become inoperable, portablesurvey instrumentation, having appropriate ranges and sensitivities to fully protectindividuals involved in refueling operation, is used until permanent instrumentation isreturned to service. The above described functions and operability requirements arecurrently described in the TMI Unit 1 UFSAR Section 11.4.2. Additionally, the TMI Unit 1UFSAR and plant procedures will be revised to incorporate the relocated TS channelcheck, test, and calibration surveillance requirements for RM-G6, RM-G7, and RM-G9.

These monitors provide high alarm and alert setpoints to call attention to an increase inradiation level and off-standard conditions. Radiation levels are closely monitored duringrefueling operations to establish the allowable exposure times for plant personnel in orderto not exceed the integrated doses specified in 10 CFR 20.

Radiation Monitors RM-G6 and RM-G7 provide no interlock functions. As described in theTMI Unit 1 UFSAR, area gamma monitor RM-G9 interlocks are designed to trip and isolatethe normal Fuel Handling Building (FHB) Ventilation System upon detection of a highradiation signal. As described in UFSAR Section 14.2.2.1.b.1, the current TMI Unit 1UFSAR Chapter 14 accident analysis for the Fuel Handling Accident in the FHB assumesthe FHB is ventilated and discharges through 90% efficient charcoal filters (FHBEngineered Safety Feature Air Treatment System) to the unit vent. The RM-G9 interlockfunction is redundant to the interlock function provided by the FHB exhaust ventilation ductatmospheric radiation monitor RM-A4. The surveillance and operability requirements forRM-A4 are specified in the TMI Unit 1 Offsite Dose Calculation Manual (ODCM). Thisfunction is not described in the current TS. However, TMI Unit 1 had previously committedin support of TS Amendment No. 248, dated December 12, 2003, to continue to test thenormal FHB Ventilation System fan stop and damper interlocks as part of the monthly andquarterly surveillances associated with RM-G9 and RM-A4 (FHB exhaust radiationmonitor) to provide assurance that the system will isolate, and these tests will becontinued. AmerGen has reanalyzed the TMI Unit 1 UFSAR Chapter 14 accident analysisfor the Fuel Handling Accident in the FHB without credit for FHB ventilation exhaustfiltration to further support the proposed change and no longer assume the RM-G9interlock functions to isolate the normal FHB Ventilation System. This reanalysis utilizesthe methodology described in Regulatory Guide 1.183, "Alternative Radiological SourceTerms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and in


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accordance with 10 CFR 50.67, "Accident source term." The NRC approved full scopeImplementation of an alternative source term in accordance with 10 CFR 50.67, in TMIUnit 1 Amendment No. 235, dated September 19, 2001. The Fuel Handling Accident inthe FHB reanalysis is provided in Enclosure 4. The revised dose consequences remainwell within the allowable dose criteria as specified in Regulatory Guide 1.183 and10 CFR 50.67. The resulting dose increases are more than the 10% minimal increasethreshold allowed by 10 CFR 50.59. Therefore, the revised accident analysis(Enclosure 4) is submitted for NRC review and approval. No change to the existingsystem design or operation is proposed.

4.0 TECHNICAL ANALYSIS

Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicantsfor nuclear power plant operating licenses to include the TS as part of the license. TheCommission's regulatory requirements related to the content for the TS are set forth in10 CFR 50.36. That regulation requires that the TS include items in eight specificcategories. The categories are (1) safety limits, limiting safety system settings, andlimiting control settings; (2) limiting conditions for operation; (3) surveillancerequirements; (4) design features; (5) administrative controls; (6) decommissioning;(7) initial notification; and (8) written reports. However, the regulation does not specify theparticular requirements to be included in a plant's TS.

The Commission amended 10 CFR 50.36 (60 FR 36593, July 19,1995), and codified fourcriteria to be used in determining whether a particular item is required to be included in alimiting condition for operation (LCO) as follows: (1) installed instrumentation that is usedto detect, and indicate in the control room, a significant abnormal degradation of thereactor coolant pressure boundary; (2) a process variable, design feature, or operatingrestriction that is an initial condition of a design-basis accident or transient analysis thateither assumes the failure of, or presents a challenge to, the integrity of a fission productbarrier; (3) a structure, system, or component that is part of the primary success path andwhich functions or actuates to mitigate a design-basis accident or transient that eitherassumes the failure of, or presents a challenge to, the integrity of a fission productbarrier; or (4) a structure, system, or component which operating experience orprobabilistic safety assessment has shown to be significant to public health and safety.

LCOs and related requirements that fall within or satisfy any of the criteria in theregulation must be retained in the TS, while those requirements that do not fall within orsatisfy these criteria may be relocated to licensee-controlled documents. The TMI Unit 1UFSAR and plant procedures are such licensee-controlled documents.

Consistent with these criteria, AmerGen proposes to relocate the reactor buildingrefueling area and spent fuel storage area radiation monitor operability requirements fromthe TMI Unit 1 TS to the UFSAR and plant procedures. The four criteria of 10 CFR 50.36are addressed below for this relocation:


Page 4 of 10

(1) The reactor building refueling area and spent fuel storage area radiation monitors arenot "instrumentation that is used to detect, and indicate in the control room, asignificant abnormal degradation of the reactor coolant pressure boundary." Thisfunction is primarily performed by reactor building sump monitoring equipment as wellas containment atmospheric monitoring instrumentation.

(2) The reactor building refueling area and spent fuel storage area radiation monitors arenot used as an initial condition of a design-basis accident or transient analysis thateither assumes the failure of, or presents a challenge to, the integrity of a fissionproduct barrier. The associated radiation monitors provide refueling and spent fuelpool area radiation monitoring for personnel protection during fuel loading andrefueling operations.

(3) The reactor building refueling area and spent fuel storage area radiation monitorfunctions currently addressed in the TMI Unit 1 TS are not used as part of the primarysuccess path which functions or actuates to mitigate a design-basis accident ortransient. The radiation monitor functions addressed in the TMI Unit 1 TS are toprovide refueling and spent fuel pool area radiation monitoring for personnelprotection during fuel loading and refueling operations. The interlock function toisolate the normal FHB Ventilation System, associated with RM-G9, is not addressedby the current TS and is no longer being assumed in the postulated design basis FuelHandling Accident in the FHB analysis, as described in Enclosure 4.

(4) Operating experiences or probabilistic safety assessments have not shown thereactor building refueling area and spent fuel storage area radiation monitors to besignificant to public health and safety. The radiation monitors RM-G6, RM-G7, andRM-G9 are considered to be non-risk contributors to the core damage frequency andoffsite dose assessment models, and as such are not part of the TMI Unit 1probabilistic risk assessment. If RM-G6, RM-G7, or RM-G9 become inoperable,portable survey instrumentation, having appropriate ranges and sensitivities to fullyprotect individuals involved in refueling operations, will continue to be used untilpermanent instrumentation is returned to service.

The relocation of the reactor building refueling area and spent fuel storage area radiationmonitor operability requirements to the Updated Final Safety Analysis Report (UFSAR)and plant procedures is fully consistent with the NRC NUREG-1430, "Standard TechnicalSpecifications, Babcock and Wilcox Plants."

Similar TS changes were previously approved by the NRC for the following plant:St. Lucie Plant, Units 1 and 2, Amendment Nos. 194 and 136, respectively, datedOctober 6, 2004.

The proposed TS changes are administrative in nature. Relocation of the reactor buildingrefueling area and spent fuel storage area radiation monitor operability requirements fromthe TS to licensee-controlled documents does not affect the plant design, hardware, orsystem operation and will not affect the ability of the radiation monitors to perform theirdesign function to protect refueling personnel. The current specified operabilityrequirements and actions taken in the event these monitors become inoperable are not


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being changed by the proposed relocation from TS to the UFSAR and plant procedures.The current TMI Unit 1 UFSAR and plant procedures describe the current TS designfunction and operability requirements for these radiation monitors. The TMI Unit 1UFSAR and plant procedures will be further revised to incorporate the relocated TSchannel check, test, and calibration surveillance requirements for RM-G6, RM-G7, andRM-G9. Any future changes to the associated radiation monitor operability requirementswould be controlled by the appropriate regulatory processes, e.g., 10 CFR 50.59 and10 CFR 50.65. Therefore, the proposed changes do not adversely affect nuclear safetyor plant operations.

Reanalysis of the Fuel Handling Accident in the Fuel Handling Building

The existing TMI Unit 1 postulated design basis Fuel Handling Accident in the FHBaccident analysis, described in UFSAR Section 14.2.2.1 .b.1, has been reanalyzedutilizing alternative source term methodology in accordance with Regulatory Guide 1.183,"Alternative Radiological Source Terms for Evaluating Design Basis Accidents at NuclearPower Reactors," and 10 CFR 50.67, "Accident source term." The NRC approved fullscope implementation of an alternative source term in accordance with 10 CFR 50.67, inTMI Unit 1 Amendment No. 235, dated September 19, 2001. The revised analysis isdescribed in TMI Unit 1 Calculation No. C-1 101 -900-EOOO-083, Revision 4, "EAB, LPZ,and CR Doses Due to Fuel Handling Accidents," and provided in Enclosure 4. Therevised analysis no longer assumes the RM-G9 interlock functions to isolate the normalFHB Ventilation System, and eliminates credit for filtration of the postulated accidentrelease through the FHB Engineered Safety Feature (ESF) Ventilation System.

Consistent with the existing UFSAR Section 14.2.2.1 .b.1 analysis, the revised analysisassumes the postulated release is through the FHB ventilation exhaust to the unit vent,but no credit is assumed for the existing charcoal filters. Consistent with the existinganalysis, an irradiated fuel decay time of 72 hours is assumed, and 56 fuel rods (all rodsin the outer row in one assembly) are assumed to be damaged. The 72-hour decay timeis based on TS Section 3.8.10, which requires at least 72 hours between reactorshutdown and the removal of irradiated fuel. The assembly damaged is assumed to bethe highest powered assembly in the core region to be discharged with a radial peakingfactor of 1.7. In accordance with Regulatory Guide 1.183, Position 3.3 for non-LOCAdesign basis accidents (DBA's) in which fuel damage is assumed, the release from thefuel gap is assumed to occur instantaneously with the onset of fuel damage. Inaccordance with Regulatory Guide 1.183, Appendix B.1.3, the chemical form ofradioiodine released from the fuel to the water is assumed to be 95% cesium iodide (Csl),4.85 percent elemental iodine, and 0.15 percent organic iodide; the Csl released from thefuel is assumed to completely dissociate in the water, and the iodine instantaneously re-evolves as elemental and organic iodine due to the low pH of the water.

In accordance with Regulatory Guide 1.183, Appendix B.2, a decontamination factor of200 is assumed for the elemental and organic iodine, with a minimum depth of waterabove the damaged fuel of 23 feet or greater consistent with the existing TMI Unit 1UFSAR accident analysis. In accordance with Regulatory Guide 1.183, Appendix B.3,the retention of noble gases in the water in the spent fuel pool is negligible and adecontamination factor of 1 is assumed, and particulate radionuclides are assumed to beretained by the water in the spent fuel pool.


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All activity not retained in the water is immediately released from the fuel to the air abovethe water and then released to the environment over a 2-hour period. The releasedactivity is then discharged to the environment without mixing or dilution in the FHB.Noble gas and iodine release fractions are doubled consistent with the conservatismsdescribed in Amendment No. 257, dated October 13, 2005. An iodine overall effectivedecontamination factor of 200 is assumed.

The design inputs and assumptions utilized in the EAB, LPZ, and Control RoomHabitability analyses are listed in Section 5.2 of Enclosure 4. In accordance withRegulatory Guide 1.183, Position 5.1.4, these design inputs are compatible with thealternative source term characteristics and TEDE dose criteria, and the assumptions arebased on Regulatory Guide 1.183, Position 3, and Appendix B guidance. With theexception of the number of fuel rods assumed to be damaged for this event (56 vs. 208),the analysis methodology is consistent with the analysis for the Fuel Handling Accident inthe Reactor Building, which was reviewed and approved by the NRC for TMI Unit 1 inAmendment No. 257, dated October 13, 2005.

The atmospheric dispersion factors (X/Qs) used are those that have previously beenreviewed and approved for TMI Unit 1 for alternative source term application by the NRCin Amendment No. 257, dated October 13, 2005.

The TMI Unit 1 Control Room emergency filtration system including the charcoal andHEPA filters are credited with 75 percent and 99 percent efficiency, respectively. Theseassumptions are bounded by the existing TMI Unit 1 TS, Section 3.15, requirements of95% and 99.95% for the charcoal and HEPA filter efficiencies, respectively. The ControlRoom is assumed to be manually isolated and the emergency ventilation system to bemanually initiated 30 minutes after the postulated accident occurs. This is conservativesince the actions necessary to accomplish this are performed within the Main ControlRoom on the H&V Panel. The control building envelope free air volume is 250,000 ft3.Control room emergency ventilation system flow rates of 8,000 cfm maximum outside airintake and 28,000 cfm minimum recirculation flow are conservatively assumed based onsystem testing. The most restrictive flow rate tolerance values are used in the analysis.A control building envelope unfiltered inleakage rate of 1000 cfm is also assumed for theduration of the accident, which bounds by a factor of approximately three the valuemeasured using a tracer gas test. However, for the first 30 minutes, an additional 60,000cfm of unfiltered intake is assumed. TMI Unit 1 control building envelope tracer gastesting was performed in August 2000 to establish a measured unfiltered inleakage rate.This testing was performed in accordance with ASTM E741-93 with the ventilation systemin the emergency lineup configuration. Unfiltered inleakage flow rates were determinedto be 233 ± 129 scfm for the "A" ventilation train and 189 ± 103 scfm for the "B" ventilationtrain. Tracer gas methods also quantified the maximum outside air supply flowrate.Consistent with the existing TMI Unit 1 licensing basis, this testing also confirmed that allrooms inside the control building envelope were at a positive pressure relative to adjacentareas outside the envelope, and the Main Control Room was maintained at a positivepressure of at least 0.1 inches w.g. with respect to adjacent areas of the control buildingenvelope. Routine preventive maintenance on critical components within the controlbuilding envelope, including differential pressure testing and monitoring of results, andimplementation of a boundary control program that includes the control building envelope,


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ensures continued control building envelope integrity. These test results andprogrammatic controls verify that the control building envelope is being adequatelymaintained, and that the proposed analysis conservatively bounds measured unfilteredinleakage into the Control Room. The above assumptions and parameters for ControlRoom Habitability are consistent with those previously reviewed and approved by theNRC in TMI Unit 1 Amendment No. 257, dated October 13, 2005.

The resulting dose consequences for the revised TMI Unit 1 Fuel Handling Accident inthe FHB, using alternative source term methodology in accordance with RegulatoryGuide 1.183 are tabulated below.

Fuel Handling Accident (FHA)Occurring In Fuel Handling Building

Post-FHA TEDE Dose (Rem)Control EAB LPZRoom

CurrentLicensing 7.20E-02 2.30E-01 4.01 E-02

Basis Dose*(with credit forFHB exhaust

filtration)

Re-Calculated 6.69E-01 1.21 E+00 2.11 E-01Licensing

Basis Dose(without credit

for FHB exhaustfiltration)

Allowable 5.OOE+00 6.30E+00 6.30E+00Dose I II

* Note - Current Licensing Basis Dose values reflect the existing UFSAR designbasis accident analysis parameters using the approved alternativesource term methodology.

The revised dose consequences remain well within the allowable dose criteria asspecified in Regulatory Guide 1.183 and 10 CFR 50.67. The resulting dose increases aremore than the 10% minimal increase threshold allowed by 10 CFR 50.59. Therefore, therevised accident analysis (Enclosure 4), which no longer assumes the RM-G9 interlockfunctions to isolate the normal FHB Ventilation System, and eliminates credit for filtrationof the postulated release through the FHB ESF Ventilation System, is submitted for NRCreview and approval.

In conclusion, the proposed relocated requirements for the reactor building refueling areaand spent fuel storage area radiation monitors are not required to be in the TS under


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10 CFR 50.36 or Section 182a of the Atomic Energy Act, and are not required to obviatethe possibility of an abnormal situation or event giving rise to an immediate threat to thepublic health and safety. In addition, sufficient regulatory controls over the relocatedrequirements exist (e.g., 10 CFR 50.59, 10 CFR 50.71(e)) to assure continued protectionof public health and safety.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration

AmerGen has evaluated whether or not a significant hazards consideration is involvedwith the proposed amendments by focusing on the three standards set forth in 10 CFR50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in theprobability or consequences of an accident previously evaluated?

Response: No.

The proposed relocation is administrative in nature and does not involve themodification of any plant equipment or affect plant operation. The associatedradiation monitors provide refueling and spent fuel pool area radiation monitoring forpersonnel protection during fuel loading and refueling operations. The associatedinstrumentation is not assumed to be an initiator of any analyzed event, nor arethese functions assumed in the mitigation of consequences of accidents.

Additionally, the associated required actions for inoperable components do notimpact the initiation or mitigation of any accident. Therefore, the proposedchange does not involve a significant increase in the probability or consequences ofan accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or differentkind of accident from any accident previously evaluated?

Response: No.

The associated radiation monitors are designed to provide refueling and spent fuelpool area radiation monitoring for personnel protection during fuel loading andrefueling operations. The proposed change is administrative in nature and does notrequire any physical alteration of plant equipment, and does not change the methodby which any safety related system performs its function. As such, no new ordifferent types of equipment will be installed, and the design function and basicoperation of installed equipment is unchanged. The methods governing plantoperation and testing remain consistent with current safety analysis assumptions.Therefore, the proposed change does not create the possibility of a new or differentkind of accident from any accident previously evaluated.


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3. Does the proposed amendment involve a significant reduction in a margin of

safety?

Response: No.

The proposed change does not negate any existing requirement, and does notadversely affect existing plant safety margins or the reliability of the equipmentassumed to operate in the safety analysis. As such, there are no changes beingmade to safety analysis assumptions, safety limits or safety system settings thatwould adversely affect plant safety as a result of the proposed change. Margins ofsafety are unaffected by requirements that are retained, but relocated from theTechnical Specifications to the UFSAR and plant procedures. Further, the proposedchange to relocate current Technical Specification requirements to the UFSAR andplant procedures is consistent with regulatory guidance and previously approvedchanges for other stations, and is administrative in nature. Therefore, the proposedchange does not involve a significant reduction in any margin of safety.

Based on the above, AmerGen concludes that the proposed amendment presents nosignificant hazards consideration under the standards set forth in 10 CFR 50.92(c), and,accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria

AmerGen has determined that the proposed change does not require any exemptions orrelief from regulatory requirements and does not affect conformance with any GeneralDesign Criteria.

The proposed change is consistent with the criteria specified in 10 CFR 50.36(c)(2)(ii)for inclusion of items in TS, and is consistent with Standard Technical Specifications(NUREG-1430, Revision 3).

In conclusion, based on the considerations discussed above, (1) there is reasonableassurance that the health and safety of the public will not be endangered by operation inthe proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimical tothe common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment does not change a requirementwith respect to installation or use of a facility component located within the restrictedarea, as defined in 10 CFR 20, and does not change an inspection or surveillancerequirement. The proposed amendment does not involve (i) a significant hazardsconsideration, (ii) a significant change in the types or significant increase in theamounts of any effluent that may be released offsite, or (iii) a significant increase inindividual or cumulative occupational radiation exposure. Accordingly, the proposedamendment meets the eligibility criterion for categorical exclusion set forth in


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10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with theproposed amendment.

7.0 REFERENCES

The NRC has approved similar changes (e.g., relocation of TS requirements which donot meet the criteria of 10 CFR 50.36(c)(2)(ii)) in a number of amendments. Theproposed change is consistent with the approach utilized by the St. Lucie Plant, Units 1and 2 in amendment application dated October 29, 2003, and approved by NRC inAmendment Nos. 194 and 136, respectively, dated October 6, 2004.

Additionally, the proposed change is fully consistent with the TS requirements of theNRC NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants."

ENCLOSURE 2

TMI Unit 1 Technical Specification Change Request No. 333

Markup of Proposed Technical Specifications and Bases Page Changes

Revised Technical Specifications & Bases Pages

3-443-454-5a

CONTROLLED COPY3.8 FUEL LOADING AND REFUELING

Apolicability: Applies to fuel loading and refueling operations.

Objective: To assure that fuel loading and refueling operations are performed in a responsiblemanner.

Specification

3.8.1 Radiation Il,,ol- 6n the Ra,• tor Building r.fueling area shall be mnite.d by .lM G6an RM iG67. -a diation 19-48o16 in tho- cpent fuel ct8rage arco she!' be rnmited byflM G9. If any of 1ho~e inctrUMont8 bocm npable, pezlable -survey isrm~~oiavim the aPpeproiatc rangoc and seonciivity toflypoct individuals irnvelved in refueling

-prti~ hall be used until the pormanont intuonain arturred ie serviee.

3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron fluxmonitors, each with continuous indication available, whenever core geometry is beingchanged. When core geometry is not being changed, at least one neutron flux monitor shallbe in service.

3.8.3 At least one decay heat removal pump and cooler shall be operable.

3.8.4 During reactor vessel head removal and while loading and unloading fuel from the reactor, theboron concentration shall be maintained at not less than that required for refueling shutdown.

3.8.5 Direct communications between the control room and the refueling personnel in the ReactorBuilding shall exist whenever changes in core geometry are taking place.

3.8.6 During the handling of irradiated fuel in the Reactor Building at least one door in each of thepersonnel and emergency air locks shall be capable of being closed.* The equipment hatchcover shall be in place with a minimum of four bolts securing the cover to the sealingsurfaces.

NOTE

The equipment hatch may be open if all of the following conditions are met:

1) The Reactor Building Equipment Hatch Missile Shield Barrier is capable of being closedwithin 45 minutes,

2) A designated crew is available to close the Reactor Building Equipment Hatch MissileShield Barrier, and

3) Reactor Building Purge Exhaust System is in service.

3.8.7 During the handling of irradiated fuel in the Reactor Building, each penetration providingdirect access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by an isolation valve, blind flange, manual valve, or equivalent, or capable ofbeing closed,* or

2. Be capable of being closed by an operable automatic containment purge and exhaust

isolation valve.

Administrative controls shall ensure that the Reactor Building Purge Exhaust System is in service,

appropriate personnel are aware that air lock doors and/or other penetrations are open, a specificindividual(s) is designated and available to close the air lock doors and other penetrations as part of arequired evacuation of containment. Any obstruction(s) (e.g., cable and hoses) that could preventclosure of an air lock door or other penetration will be capable of being quickly removed.

3-44

Amendment No. 27498,236, ,

CONTROLLED COPY3.8.8 If any of the above specified limiting conditions for fuel loading and refueling are not met,

movement of fuel into the reactor core shall cease; action shall be initiated to correct theconditions so that the specified limits are met, and no operations which may increase thereactivity of the core shall be made.

3.8.9 The reactor building purge isolation valves, and associated radiation monitors which initiatepurge isolation, shall be tested and verified to be operable no more than 7 days prior to initialfuel movement in the reactor building.

3.8.10 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for atleast 72 hours.

3.8.11 During the handling of irradiated fuel in the Reactor Building at least 23 feet of water shall bemaintained above the level of the reactor pressure vessel flange, as determined by a shiftlycheck and a daily verification. If the water level is less than 23 feet above the reactor pressurevessel flange, place the fuel assembly(s) being handled into a safe position, then cease fuelhandling until the water level has been restored to 23 feet or greater above the reactor pressurevessel flange.

Bases

Detailed written procedures will be available for use by refueling personnel. These procedures, theabove specifications, and the design of the fuel handling equipment as described in Section 9.7 of theUFSAR incorporating built-in interlocks and safety features, provide assurance that no incident couldoccur during the refueling operations that would result in a hazard to public.health and safety. If nochange is being made in core geometry, one flux monitor is sufficient. This permits maintenance on theinstrumentation. Continuous monitoring of radiation levels_ ",-"eutron flux provides immediateindication of an unsafe condition. The decay heat removal pump is used to maintain a uniform boronconcentration. The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, evenwith all control rods withdrawn from the core (Reference 1). The boron concentration will be sufficient tomaintain the core keff < 0.99 if all the control rods were removed from the core, however only a fewcontrol rods will be removed at any one time during fuel shuffling and replacement. The ke, with all rodsin the core and with refueling boron concentration is approximately 0.9. Specification 3.8.5 allows thecontrol room operator to inform the reactor building personnel of any impending unsafe conditiondetected from the main control board indicators during fuel movement.

Per Specification 3.8.6 and 3.8.7, the personnel and emergency air lock doors, and penetrations may beopen during movement of irradiated fuel in the containment provided a minimum of one door in each ofthe air locks, and penetrations are capable of being closed in the event of a fuel handling accident, andthe plant is in REFUELING SHUTDOWN or REFUELING OPERATION with at least 23 feet of waterabove the fuel seated within the reactor pressure vessel. The minimum water level specified is the basisfor the accident analysis assumption of a decontamination factor of 200 for the release to thecontainment atmosphere from the postulated damaged fuel rods located on top of the fuel core seated inthe reactor vessel. Should a fuel handling accident occur inside containment, a minimum of one door ineach personnel and emergency air lock, and the open penetrations will be closed following anevacuation of containment. Administrative controls will be in place to assure closure of at least one doorin each air lock, as well as other open containment penetrations, following a containment evacuation.

Specification 3.8.6 is modified by a NOTE:------------------------------ NOTE-----------------------------The equipment hatch may be open if all of the following conditions are met:

1) The Reactor Building Equipment Hatch Missile Shield Barrier is capable of being closedwithin 45 minutes,

2) A designated crew is available to close the Reactor Building Equipment Hatch MissileShield Barrier, and

3) Reactor Building Purge Exhaust System is in service.-----------------------------------------------------------

3-45Amendment No. 157, 178, 236, 215, 250,--259

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION

27. Makeup Tank Instrument Channels:

a Level

b. Pressure

28. RadiatiQn onitoring Systems*

E• re j - --

c. RM G9 (FHW4ridgo -FH-Bldg)-

CHECK TEST CALIBRATE REMARKS

D(1)

D(1)

NA

NA

R

R

WOR) IVIR QR

M(2) Q(2)

-W(1)(3) M(3) E(3)

(1) When Makeup and Purification System isin operation.

(1) Using the installed check source whenbackground is less than twice the expectedincrease in cpm which would result from thecheck source alone. Background readingsgreater than this value are sufficient inthemselves to show that the monitor isfunctioning.

(2) +Rpcability rcquinents

.requird-t bea cu rrent onlY When handling

U3CD

d. RM-A2P (RB Atmosphere particulate) W(1)(4) M(4) E(4)

e. RM-A21 (RB Atmosphere iodine)

f. RM-A2G (RB Atmosphere gas)

W(1)(4) M(4) Q(4)

W(1)(4) M(4) E(4)! i.- r ;,atcd fu. l.

(3) AM GOr epcrbe t-given inT.S. 3.8.1.

.nurmcnts erc

L

L

(4) RM-A2 operability requirements aregiven in T.S. 3.1.6.8

29. High and Low PressureInjection Systems:Flow Channels

N/A N/A R I

* Includes only monitors indicated under this item. Other T.S. required radiation monitors are included in specifications 3.5.5.2, 4.1.3,Table 3.5-1 item C.3.f, and Table 4.1-1 item 19e.

ENCLOSURE 3

List of Commitments

Enclosure 35928-06-20498

Page 1 of 1

SUMMARY OF AMERGEN COMMITMENTS

The following table identifies regulatory commitments made in this document by AmerGen. (Any otheractions discussed in the submittal represent intended or planned actions by AmerGen. They aredescribed to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT COMMITTED DATEOR "OUTAGE"

The TMI Unit 1 UFSAR and plant procedures will be Upon implementation offurther revised to incorporate the relocated TS channel amendment for the proposedcheck, test, and calibration surveillance requirements for change.RM-G6, RM-G7, and RM-G9.

ENCLOSURE 4

TMI Unit 1 Calculation No. C-1101-900-EOOO-083, Revision 4, "EAB, LPZ, andCR Doses Due to Fuel Handling Accidents"

three mile island, unit 1, technical specification change ... · radiation monitor rm-g9 is an area...

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