ML063540168
| ML063540168 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 12/12/2006 |
| From: | Cowan P AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 5928-06-20498 | |
| Download: ML063540168 (20) | |
Text
AmerGen Energy Company, LLC www.exeloncorp.com AmerQKbG An Exelon Company SM 20o Exelon Way Kennett Square, PA 19348 10 CFR 50.90 December 12, 2006 5928-06-20498 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Technical Specification Change Request No. 333 - Relocation of Technical Specification Requirements for Refuel and Spent Fuel Pool Area Radiation Monitors In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," 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 refueling area and spent fuel storage area radiation monitor operability requirements to the Updated Final Safety Analysis Report (UFSAR) and plant procedures, since these radiation monitors do not meet the criteria for inclusion in the TS as presented in 10 CFR 50.36(c)(2)(ii). These changes are consistent with Standard Technical Specifications (NUREG-1430, Revision 3). To further support the proposed change, the current TMI Unit 1 Fuel Handling Accident in the Fuel Handling Building, described in UFSAR Section 14.2.2.1 .b.1, has been reanalyzed without credit for Fuel Handling Building ventilation exhaust filtration. This reanalysis is provided for NRC review and approval in Enclosure 4.
The proposed amendment has been reviewed by the TMI Unit 1 Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the AmerGen Quality Assurance Program.
Using the standards in 10 CFR 50.92, AmerGen has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed analysis performed in accordance with 10 CFR 50.91 (a)(1). Pursuant to 10 CFR 50.91 (b)(1), a copy of this Technical Specification Change Request is provided to the designated official of the Commonwealth of Pennsylvania, Bureau of Radiation Protection, as well as the chief executives of the township and county in which the facility is located.
U.S. Nuclear Regulatory Commission December 12, 2006 Page 2 We request approval of the proposed change by September 30, 2007, with the amendment being implemented within 30 days of issuance. This will allow an orderly implementation of these changes 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 have any questions or require additional information, please contact David J. Distel at (610) 765-5517.
Executed on the Ikday 1declare under penalty of perjury that the foregoing is true and correct.
of December, 2006.
Respectfully, Pamela B. Cowan Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC
Enclosures:
- 1) TMI Unit 1 Technical Specification Change Request No. 333 - Description and Assessment
- 2) TMI Unit 1 Technical Specification Change Request No. 333 - Markup of Proposed Technical Specification and Bases Page Changes
- 3) List of Commitments
- 4) 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 I D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 F. E. Saba, USNRC Project Manager, TMI Unit 1 D. Allard, Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection Chairman, Board of County Commissioners of Dauphin County, PA Chairman, Board of Supervisors of Londonderry Township, Dauphin County, PA TMI 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 Area Radiation Monitors Description and Assessment
Enclosure 1 Description 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 construction permit," AmerGen Energy Company, LLC (AmerGen) is requesting an amendment to Facility 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 Technical Specifications (TS) to relocate the reactor building refueling area and spent fuel storage area radiation monitor operability requirements to the Updated Final Safety Analysis Report (UFSAR) and plant procedures, since these radiation monitors do not meet the criteria for inclusion in the TS as presented in 10 CFR 50.36(c)(2)(ii). These changes are consistent with Standard Technical Specifications (NUREG-1430, Revision 3).
AmerGen requests that the following changed replacement pages be inserted into the existing 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 requirements stated below:
"Radiation levels in the Reactor Building refueling area shall be monitored by RM-G6 and RM-G7. Radiation levels in the spent fuel storage area shall be monitored by RM-G9. If any of these instruments become inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refueling operation, 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 the following sub-items and associated surveillance requirements: "a. RM-G6 (FH Bridge #1 Aux)", "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-G9 and thus are being deleted.
3.0 BACKGROUND
Radiation Monitors RM-G6, RM-G7, and RM-G9 are designed to provide refueling and spent fuel pool area radiation monitoring for personnel protection during fuel loading and refueling operations. These channels monitor radiation levels in the associated fuel handling areas.
Enclosure 1 Description and Assessment Page 2 of 10 Radiation Monitors RM-G6 and RM-G7 are area gamma monitors currently located on the Reactor Building fuel handling bridges. RM-G6 is located on the Auxiliary Bridge at approximately elevation 346 ft. in the Reactor Building. RM-G7 is located on the Main Bridge at approximately elevation 346 ft. in the Reactor Building. Although isolation of the Reactor Building is not credited in the fuel handling accident analysis for TMI Unit 1, as described in existing UFSAR Section 14.2.2.1 .b.2, these monitors alarm any excessive radiation in the vicinity of the refueling water surface and provide the Control Room operators sufficient information to initiate evacuation and closure of the Reactor Building in the event of a fuel handling accident. These functions are not affected by the proposed change.
Radiation Monitor RM-G9 is an area gamma monitor currently located in the Fuel Handling Building (FHB) on the East Wall in the vicinity of the spent fuel pool water surface at approximately elevation 346 ft.
As specified in the existing TS, if RM-G6, RM-G7, or RM-G9 become inoperable, portable survey instrumentation, having appropriate ranges and sensitivities to fully protect individuals involved in refueling operation, is used until permanent instrumentation is returned to service. The above described functions and operability requirements are currently described in the TMI Unit 1 UFSAR Section 11.4.2. Additionally, the TMI Unit 1 UFSAR and plant procedures will be revised to incorporate the relocated TS channel check, 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 in radiation level and off-standard conditions. Radiation levels are closely monitored during refueling operations to establish the allowable exposure times for plant personnel in order to not exceed the integrated doses specified in 10 CFR 20.
Radiation Monitors RM-G6 and RM-G7 provide no interlock functions. As described in the TMI Unit 1 UFSAR, area gamma monitor RM-G9 interlocks are designed to trip and isolate the normal Fuel Handling Building (FHB) Ventilation System upon detection of a high radiation signal. As described in UFSAR Section 14.2.2.1.b.1, the current TMI Unit 1 UFSAR Chapter 14 accident analysis for the Fuel Handling Accident in the FHB assumes the FHB is ventilated and discharges through 90% efficient charcoal filters (FHB Engineered Safety Feature Air Treatment System) to the unit vent. The RM-G9 interlock function is redundant to the interlock function provided by the FHB exhaust ventilation duct atmospheric radiation monitor RM-A4. The surveillance and operability requirements for RM-A4 are specified in the TMI Unit 1 Offsite Dose Calculation Manual (ODCM). This function is not described in the current TS. However, TMI Unit 1 had previously committed in support of TS Amendment No. 248, dated December 12, 2003, to continue to test the normal FHB Ventilation System fan stop and damper interlocks as part of the monthly and quarterly surveillances associated with RM-G9 and RM-A4 (FHB exhaust radiation monitor) to provide assurance that the system will isolate, and these tests will be continued. AmerGen has reanalyzed the TMI Unit 1 UFSAR Chapter 14 accident analysis for the Fuel Handling Accident in the FHB without credit for FHB ventilation exhaust filtration to further support the proposed change and no longer assume the RM-G9 interlock functions to isolate the normal FHB Ventilation System. This reanalysis utilizes the methodology described in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and in
Enclosure 1 Description and Assessment Page 3 of 10 accordance with 10 CFR 50.67, "Accident source term." The NRC approved full scope Implementation of an alternative source term in accordance with 10 CFR 50.67, in TMI Unit 1 Amendment No. 235, dated September 19, 2001. The Fuel Handling Accident in the FHB reanalysis is provided in Enclosure 4. The revised dose consequences remain well within the allowable dose criteria as specified in Regulatory Guide 1.183 and 10 CFR 50.67. The resulting dose increases are more than the 10% minimal increase threshold allowed by 10 CFR 50.59. Therefore, the revised accident analysis (Enclosure 4) is submitted for NRC review and approval. No change to the existing system design or operation is proposed.
4.0 TECHNICAL ANALYSIS
Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TS as part of the license. The Commission's regulatory requirements related to the content for the TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in eight specific categories. The categories are (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. However, the regulation does not specify the particular requirements to be included in a plant's TS.
The Commission amended 10 CFR 50.36 (60 FR 36593, July 19,1995), and codified four criteria to be used in determining whether a particular item is required to be included in a limiting condition for operation (LCO) as follows: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; or (4) a structure, system, or component which operating experience or probabilistic 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 the regulation must be retained in the TS, while those requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The TMI Unit 1 UFSAR and plant procedures are such licensee-controlled documents.
Consistent with these criteria, AmerGen proposes to relocate the reactor building refueling area and spent fuel storage area radiation monitor operability requirements from the TMI Unit 1 TS to the UFSAR and plant procedures. The four criteria of 10 CFR 50.36 are addressed below for this relocation:
Enclosure 1 Description and Assessment Page 4 of 10 (1) The reactor building refueling area and spent fuel storage area radiation monitors are not "instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." This function is primarily performed by reactor building sump monitoring equipment as well as containment atmospheric monitoring instrumentation.
(2) The reactor building refueling area and spent fuel storage area radiation monitors are not used as an initial condition of a design-basis accident or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier. The associated radiation monitors provide refueling and spent fuel pool area radiation monitoring for personnel protection during fuel loading and refueling operations.
(3) The reactor building refueling area and spent fuel storage area radiation monitor functions currently addressed in the TMI Unit 1 TS are not used as part of the primary success path which functions or actuates to mitigate a design-basis accident or transient. The radiation monitor functions addressed in the TMI Unit 1 TS are to provide refueling and spent fuel pool area radiation monitoring for personnel protection during fuel loading and refueling operations. The interlock function to isolate the normal FHB Ventilation System, associated with RM-G9, is not addressed by the current TS and is no longer being assumed in the postulated design basis Fuel Handling Accident in the FHB analysis, as described in Enclosure 4.
(4) Operating experiences or probabilistic safety assessments have not shown the reactor building refueling area and spent fuel storage area radiation monitors to be significant to public health and safety. The radiation monitors RM-G6, RM-G7, and RM-G9 are considered to be non-risk contributors to the core damage frequency and offsite dose assessment models, and as such are not part of the TMI Unit 1 probabilistic risk assessment. If RM-G6, RM-G7, or RM-G9 become inoperable, portable survey instrumentation, having appropriate ranges and sensitivities to fully protect individuals involved in refueling operations, will continue to be used until permanent instrumentation is returned to service.
The relocation of the reactor building refueling area and spent fuel storage area radiation monitor operability requirements to the Updated Final Safety Analysis Report (UFSAR) and plant procedures is fully consistent with the NRC NUREG-1430, "Standard Technical Specifications, 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, dated October 6, 2004.
The proposed TS changes are administrative in nature. Relocation of the reactor building refueling area and spent fuel storage area radiation monitor operability requirements from the TS to licensee-controlled documents does not affect the plant design, hardware, or system operation and will not affect the ability of the radiation monitors to perform their design function to protect refueling personnel. The current specified operability requirements and actions taken in the event these monitors become inoperable are not
Enclosure 1 Description and Assessment Page 5 of 10 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 design function and operability requirements for these radiation monitors. The TMI Unit 1 UFSAR and plant procedures will be further revised to incorporate the relocated TS channel check, test, and calibration surveillance requirements for RM-G6, RM-G7, and RM-G9. Any future changes to the associated radiation monitor operability requirements would be controlled by the appropriate regulatory processes, e.g., 10 CFR 50.59 and 10 CFR 50.65. Therefore, the proposed changes do not adversely affect nuclear safety or 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 FHB accident analysis, described in UFSAR Section 14.2.2.1 .b.1, has been reanalyzed utilizing alternative source term methodology in accordance with Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and 10 CFR 50.67, "Accident source term." The NRC approved full scope implementation of an alternative source term in accordance with 10 CFR 50.67, in TMI Unit 1 Amendment No. 235, dated September 19, 2001. The revised analysis is described 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. The revised analysis no longer assumes the RM-G9 interlock functions to isolate the normal FHB Ventilation System, and eliminates credit for filtration of the postulated accident release through the FHB Engineered Safety Feature (ESF) Ventilation System.
Consistent with the existing UFSAR Section 14.2.2.1 .b.1 analysis, the revised analysis assumes 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 existing analysis, an irradiated fuel decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is assumed, and 56 fuel rods (all rods in the outer row in one assembly) are assumed to be damaged. The 72-hour decay time is based on TS Section 3.8.10, which requires at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between reactor shutdown and the removal of irradiated fuel. The assembly damaged is assumed to be the highest powered assembly in the core region to be discharged with a radial peaking factor of 1.7. In accordance with Regulatory Guide 1.183, Position 3.3 for non-LOCA design basis accidents (DBA's) in which fuel damage is assumed, the release from the fuel gap is assumed to occur instantaneously with the onset of fuel damage. In accordance with Regulatory Guide 1.183, Appendix B.1.3, the chemical form of radioiodine 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 the fuel 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 of 200 is assumed for the elemental and organic iodine, with a minimum depth of water above the damaged fuel of 23 feet or greater consistent with the existing TMI Unit 1 UFSAR 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 a decontamination factor of 1 is assumed, and particulate radionuclides are assumed to be retained by the water in the spent fuel pool.
Enclosure 1 Description and Assessment Page 6 of 10 All activity not retained in the water is immediately released from the fuel to the air above the water and then released to the environment over a 2-hour period. The released activity is then discharged to the environment without mixing or dilution in the FHB.
Noble gas and iodine release fractions are doubled consistent with the conservatisms described in Amendment No. 257, dated October 13, 2005. An iodine overall effective decontamination factor of 200 is assumed.
The design inputs and assumptions utilized in the EAB, LPZ, and Control Room Habitability analyses are listed in Section 5.2 of Enclosure 4. In accordance with Regulatory Guide 1.183, Position 5.1.4, these design inputs are compatible with the alternative source term characteristics and TEDE dose criteria, and the assumptions are based on Regulatory Guide 1.183, Position 3, and Appendix B guidance. With the exception 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 in the Reactor Building, which was reviewed and approved by the NRC for TMI Unit 1 in Amendment No. 257, dated October 13, 2005.
The atmospheric dispersion factors (X/Qs) used are those that have previously been reviewed and approved for TMI Unit 1 for alternative source term application by the NRC in Amendment No. 257, dated October 13, 2005.
The TMI Unit 1 Control Room emergency filtration system including the charcoal and HEPA filters are credited with 75 percent and 99 percent efficiency, respectively. These assumptions are bounded by the existing TMI Unit 1 TS, Section 3.15, requirements of 95% and 99.95% for the charcoal and HEPA filter efficiencies, respectively. The Control Room is assumed to be manually isolated and the emergency ventilation system to be manually initiated 30 minutes after the postulated accident occurs. This is conservative since the actions necessary to accomplish this are performed within the Main Control Room 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 air intake and 28,000 cfm minimum recirculation flow are conservatively assumed based on system 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 the duration of the accident, which bounds by a factor of approximately three the value measured using a tracer gas test. However, for the first 30 minutes, an additional 60,000 cfm of unfiltered intake is assumed. TMI Unit 1 control building envelope tracer gas testing 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 system in the emergency lineup configuration. Unfiltered inleakage flow rates were determined to be 233 +/- 129 scfm for the "A" ventilation train and 189 +/- 103 scfm for the "B" ventilation train. 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 all rooms inside the control building envelope were at a positive pressure relative to adjacent areas outside the envelope, and the Main Control Room was maintained at a positive pressure of at least 0.1 inches w.g. with respect to adjacent areas of the control building envelope. Routine preventive maintenance on critical components within the control building envelope, including differential pressure testing and monitoring of results, and implementation of a boundary control program that includes the control building envelope,
Enclosure 1 Description and Assessment Page 7 of 10 ensures continued control building envelope integrity. These test results and programmatic controls verify that the control building envelope is being adequately maintained, and that the proposed analysis conservatively bounds measured unfiltered inleakage into the Control Room. The above assumptions and parameters for Control Room Habitability are consistent with those previously reviewed and approved by the NRC in TMI Unit 1 Amendment No. 257, dated October 13, 2005.
The resulting dose consequences for the revised TMI Unit 1 Fuel Handling Accident in the FHB, using alternative source term methodology in accordance with Regulatory Guide 1.183 are tabulated below.
Fuel Handling Accident (FHA)
Occurring In Fuel Handling Building Post-FHA TEDE Dose (Rem)
Control EAB LPZ Room Current Licensing 7.20E-02 2.30E-01 4.01 E-02 Basis Dose*
(with credit for FHB exhaust filtration)
Re-Calculated 6.69E-01 1.21 E+00 2.11 E-01 Licensing Basis Dose (without credit for FHB exhaust filtration)
Allowable 5.OOE+00 6.30E+00 6.30E+00 Dose I II
- Note - Current Licensing Basis Dose values reflect the existing UFSAR design basis accident analysis parameters using the approved alternative source term methodology.
The revised dose consequences remain well within the allowable dose criteria as specified in Regulatory Guide 1.183 and 10 CFR 50.67. The resulting dose increases are more than the 10% minimal increase threshold allowed by 10 CFR 50.59. Therefore, the revised accident analysis (Enclosure 4), which no longer assumes the RM-G9 interlock functions to isolate the normal FHB Ventilation System, and eliminates credit for filtration of the postulated release through the FHB ESF Ventilation System, is submitted for NRC review and approval.
In conclusion, the proposed relocated requirements for the reactor building refueling area and spent fuel storage area radiation monitors are not required to be in the TS under
Enclosure 1 Description and Assessment Page 8 of 10 10 CFR 50.36 or Section 182a of the Atomic Energy Act, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. In addition, sufficient regulatory controls over the relocated requirements exist (e.g., 10 CFR 50.59, 10 CFR 50.71(e)) to assure continued protection of 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 involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed relocation is administrative in nature and does not involve the modification of any plant equipment or affect plant operation. The associated radiation monitors provide refueling and spent fuel pool area radiation monitoring for personnel protection during fuel loading and refueling operations. The associated instrumentation is not assumed to be an initiator of any analyzed event, nor are these functions assumed in the mitigation of consequences of accidents.
Additionally, the associated required actions for inoperable components do not impact the initiation or mitigation of any accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The associated radiation monitors are designed to provide refueling and spent fuel pool area radiation monitoring for personnel protection during fuel loading and refueling operations. The proposed change is administrative in nature and does not require any physical alteration of plant equipment, and does not change the method by which any safety related system performs its function. As such, no new or different types of equipment will be installed, and the design function and basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Enclosure 1 Description and Assessment Page 9 of 10
- 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 not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the Technical Specifications to the UFSAR and plant procedures. Further, the proposed change to relocate current Technical Specification requirements to the UFSAR and plant procedures is consistent with regulatory guidance and previously approved changes for other stations, and is administrative in nature. Therefore, the proposed change does not involve a significant reduction in any margin of safety.
Based on the above, AmerGen concludes that the proposed amendment presents no significant 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 or relief from regulatory requirements and does not affect conformance with any General Design 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 reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the 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 requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and does not change an inspection or surveillance requirement. The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in
Enclosure 1 Description and Assessment Page 10 of 10 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
7.0 REFERENCES
The NRC has approved similar changes (e.g., relocation of TS requirements which do not meet the criteria of 10 CFR 50.36(c)(2)(ii)) in a number of amendments. The proposed change is consistent with the approach utilized by the St. Lucie Plant, Units 1 and 2 in amendment application dated October 29, 2003, and approved by NRC in Amendment Nos. 194 and 136, respectively, dated October 6, 2004.
Additionally, the proposed change is fully consistent with the TS requirements of the NRC 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 3-44 3-45 4-5a
CONTROLLED COPY 3.8 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 responsible manner.
Specification 3.8.1 Radiation Il,,ol- 6n the Ra,* tor Building r.fueling area shall .lM be mnite.d by G6 an RMiG67. -adiation 19-48o16 intho- cpent fuel ct8rage arco she!' be rnmited by flM G9. If any of 1ho~e inctrUMont8 bocm npable, pezlable -surveyisrm~~o iavim the aPpeproiatc rangoc and seonciivity toflypoct individuals irnvelved inrefueling hall be used until the pormanont intuonain arturred ie serviee.
-prti~
3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with continuous indication available, whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be 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, the boron 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 Reactor Building 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 the personnel and emergency air locks shall be capable of being closed.* The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces.
NOTE The equipment hatch may be open ifall of the following conditions are met:
- 1) The Reactor Building Equipment Hatch Missile Shield Barrier is capable of being closed within 45 minutes,
- 2) A designated crew is available to close the Reactor Building Equipment Hatch Missile Shield 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 providing direct 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 of being 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 specific individual(s) is designated and available to close the air lock doors and other penetrations as part of a required evacuation of containment. Any obstruction(s) (e.g., cable and hoses) that could prevent closure of an air lock door or other penetration will be capable of being quickly removed.
3-44 Amendment No. 27498,236, ,
CONTROLLED COPY 3.8.8 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 the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.
3.8.9 The reactor building purge isolation valves, and associated radiation monitors which initiate purge isolation, shall be tested and verified to be operable no more than 7 days prior to initial fuel movement in the reactor building.
3.8.10 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.8.11 During the handling of irradiated fuel in the Reactor Building at least 23 feet of water shall be maintained above the level of the reactor pressure vessel flange, as determined by a shiftly check and a daily verification. If the water level is less than 23 feet above the reactor pressure vessel flange, place the fuel assembly(s) being handled into a safe position, then cease fuel handling until the water level has been restored to 23 feet or greater above the reactor pressure vessel flange.
Bases Detailed written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.7 of the UFSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public.health and safety. If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation. Continuous monitoring of radiation levels_ ",-"eutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a uniform boron concentration. The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core (Reference 1). The boron concentration will be sufficient to maintain the core keff < 0.99 if all the control rods were removed from the core, however only a few control rods will be removed at any one time during fuel shuffling and replacement. The ke, with all rods in the core and with refueling boron concentration is approximately 0.9. Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected 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 be open during movement of irradiated fuel in the containment provided a minimum of one door in each of the air locks, and penetrations are capable of being closed in the event of a fuel handling accident, and the plant is in REFUELING SHUTDOWN or REFUELING OPERATION with at least 23 feet of water above the fuel seated within the reactor pressure vessel. The minimum water level specified is the basis for the accident analysis assumption of a decontamination factor of 200 for the release to the containment atmosphere from the postulated damaged fuel rods located on top of the fuel core seated in the reactor vessel. Should a fuel handling accident occur inside containment, a minimum of one door in each personnel and emergency air lock, and the open penetrations will be closed following an evacuation of containment. Administrative controls will be in place to assure closure of at least one door in 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 closed within 45 minutes,
- 2) A designated crew is available to close the Reactor Building Equipment Hatch Missile Shield Barrier, and
- 3) Reactor Building Purge Exhaust System is in service.
3-45 Amendment No. 157, 178, 236, 215, 250,--259
TABLE 4.1-1 (Continued)
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS
- 27. Makeup Tank Instrument Channels:
a Level D(1) NA R (1) When Makeup and Purification System is in operation.
- b. Pressure D(1) NA R
- 28. RadiatiQn onitoring Systems*
WOR) IVIR QR (1) Using the installed check source when E*re j - -- background is less than twice the expected M(2) Q(2) increase in cpm which would result from the check source alone. Background readings U3 greater than this value are sufficient in
- c. RM G9 (FHW4ridgo -FH-Bldg)- -W(1)(3) M(3) E(3)
CD themselves to show that the monitor is
- d. RM-A2P (RB Atmosphere particulate) W(1)(4) M(4) E(4) functioning.
+Rpcability L
- f. RM-A2G (RB Atmosphere gas) W(1)(4) M(4) E(4)
(3)
.requird-t bea cu rrent onlY i.- r ;,atcd fu. l.
AMGOr epcrbe
-given inT.S. 3.8.1.
t When handling
.nurmcnts erc (4) RM-A2 operability requirements are L
given in T.S. 3.1.6.8
- 29. High and Low Pressure N/A N/A R Injection Systems: I Flow Channels
- 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 3 5928-06-20498 Page 1 of 1
SUMMARY
OF AMERGEN COMMITMENTS The following table identifies regulatory commitments made in this document by AmerGen. (Any other actions discussed in the submittal represent intended or planned actions by AmerGen. They are described to the NRC for the NRC's information and are not regulatory commitments.)
COMMITMENT COMMITTED DATE OR "OUTAGE" The TMI Unit 1 UFSAR and plant procedures will be Upon implementation of further revised to incorporate the relocated TS channel amendment for the proposed check, 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, and CR Doses Due to Fuel Handling Accidents"