ML062920138
| ML062920138 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/16/2006 |
| From: | John Honcharik NRC/NRR/ADRO/DORL/LPLIII-2 |
| To: | Crane C Exelon Generation Co |
| Gratton C, NRR/DORL 415-1055 | |
| Shared Package | |
| ML062920155 | List: |
| References | |
| TAC MD2166, TAC MD2167 | |
| Download: ML062920138 (14) | |
Text
November 16, 2006 Mr. Christopher M. Crane, President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENTS RE: STANDBY LIQUID CONTROL SYSTEM SODIUM PENTABORATE ISOTOPIC ENRICHMENT (TAC NOS. MD2166 AND MD2167)
Dear Mr. Crane:
The Commission has issued the enclosed Amendment No. 222 to Renewed Facility Operating License No. DPR-19 and Amendment No. 214 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3. The amendments are in response to your application dated June 2, 2006, as supplemented by letters dated August 18 and October 5, 2006.
The amendments revise Technical Specification Surveillance Requirement (SR) 3.1.7.10, Standby Liquid Control System Sodium Pentaborate Isotopic Enrichment such that the required enrichment would increase from $ 30.0 atom percent boron-10 to $ 45.0 atom percent boron-10. The original amendment request dated June 2, 2006, also requested a revision to SR 3.4.3.1 to increase the allowable as-found main steam safety valve lift setpoint tolerance which will be addressed in future NRC correspondence.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA by C. Gratton for/
John Honcharik, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249
Enclosures:
- 1. Amendment No. 222 to DPR-19
- 2. Amendment No. 214 to DPR-25
- 3. Safety Evaluation cc w/encls: See next page
November 16, 2006 Mr. Christopher M. Crane, President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENTS RE: STANDBY LIQUID CONTROL SYSTEM SODIUM PENTABORATE ISOTOPIC ENRICHMENT (TAC NOS. MD2166 AND MD2167)
Dear Mr. Crane:
The Commission has issued the enclosed Amendment No. 222 to Renewed Facility Operating License No. DPR-19 and Amendment No. 214 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3. The amendments are in response to your application dated June 2, 2006, as supplemented by letters dated August 18 and October 5, 2006.
The amendments revise Technical Specification Surveillance Requirement (SR) 3.1.7.10, Standby Liquid Control System Sodium Pentaborate Isotopic Enrichment such that the required enrichment would increase from $ 30.0 atom percent boron-10 to $ 45.0 atom percent boron-10. The original amendment request dated June 2, 2006, also requested a revision to SR 3.4.3.1 to increase the allowable as-found main steam safety valve lift setpoint tolerance which will be addressed in future NRC correspondence.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA by C. Gratton for/
John Honcharik, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249
Enclosures:
- 1. Amendment No. 222 to DPR-19
- 2. Amendment No. 214 to DPR-25
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
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RidsNrrLADClarke RidsRgn3MailCenter RidsNrrDorlDpr RidsNrrDssSbwb Package: ML062920155 Amendment: ML062920138 TS Pages: ML063210154 OFFICE LPL3-2/PM LPL3-2/LA DSS/SBWB/BC DIRS/ITSB/BC OGC LPL3-2/BC NAME JHoncharik:mw DClarke GCranston TKobetz BPoole DCollins DATE 11/16/2006 11 /15/2006 11/15/2006 11/15/2006 11/16/2006 11/16/2006 OFFICIAL RECORD COPY
Dresden Nuclear Power Units 2 and 3 cc:
Site Vice President - Dresden Exelon Generation Company, LLC 6500 N. Dresden Road Morris, IL 60450-9765 Plant Manager - Dresden Nuclear Power Station Exelon Generation Company, LLC 6500 N. Dresden Road Morris, IL 60450-9765 Manager Regulatory Assurance - Dresden Exelon Generation Company, LLC 6500 N. Dresden Road Morris, IL 60450-9765 U.S. Nuclear Regulatory Commission Dresden Resident Inspectors Office 6500 N. Dresden Road Morris, IL 60450-9766 Chairman Grundy County Board Administration Building 1320 Union Street Morris, IL 60450 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Illinois Emergency Management Agency Division of Disaster Assistance &
Preparedness 110 East Adams Street Springfield, IL 62701-1109 Document Control Desk - Licensing Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Senior Vice President - Midwest Operations Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Vice President - Regulatory & Legal Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Assistant General Counsel Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348 Manager Licensing - Dresden Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
EXELON GENERATION COMPANY, LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 222 Renewed License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Exelon Generation Company, LLC (the licensee) dated June 2, 2006, as supplemented by letters dated August 18 and October 5, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-19 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 222, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Daniel S. Collins, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed License No. DPR-19 and Technical Specifications Date of Issuance: November 16, 2006
EXELON GENERATION COMPANY, LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 214 Renewed License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Exelon Generation Company, LLC (the licensee) dated June 2, 2006, as supplemented by letters dated August 18 and October 5, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B. of Renewed Facility Operating License No. DPR-25 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 214, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Daniel S. Collins, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed License No. DPR-25 and Technical Specifications Date of Issuance: November 16, 2006
ATTACHMENT TO LICENSE AMENDMENT NOS. 222 AND 214 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Insert License DRP-19 License DRP-19 Page 3 Page 3 License DRP-25 License DRP-25 Page 4 Page 4 TSs TSs 3.1.7-3 3.1.7-3 (2)
Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)
Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 222, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Operation in the coastdown mode is permitted to 40% power.
Renewed License No. DPR-19 Amendment No. 222 f.
Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)
Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14).
3.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 2957 megawatts (thermal), except that the licensee shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in the licensees telegrams; dated February 26, 1971, have been verified in writing by the Commission.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 214, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D.
Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.
E.
Restrictions Operation in the coastdown mode is permitted to 40% power.
Renewed License No. DPR-25 Amendment No. 214
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 222 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 214 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25 EXELON GENERATION COMPANY, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249
1.0 INTRODUCTION
By letter to the Nuclear Regulatory Commission (NRC, the Commission) dated June 2, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML062130031), as supplemented by letters dated August 18 (ADAMS Accession Number ML062420088) and October 5, 2006 (ADAMS Accession Number ML062790121), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) for the Dresden Nuclear Power Station, Units 2 and 3 (Dresden). The proposed changes would revise TS Surveillance Requirement (SR) 3.1.7.10, Standby Liquid Control System
[SLCS] Sodium Pentaborate Isotopic Enrichment such that the required enrichment would increase from $ 30.0 atom percent boron-10 to $ 45.0 atom percent boron-10. The licensees original amendment request dated June 2, 2006, also requested a revision to increase the main steam safety valve (MSSV) lift setpoint tolerance which will be addressed in future NRC correspondence.
The August 18 and October 5, 2006, supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 15, 2006 (71 FR 46931).
2.0 REGULATORY EVALUATION
The NRC staff finds that the licensee, in Section 5.0 of its original submittal, identified the applicable regulatory requirements.
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A provides the General Design Criteria (GDCs) for nuclear power plants. GDC 26, Reactivity control system redundancy and capability, states:
Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
The provisions of 10 CFR 50.62, ?Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants, require, in part, that each boiling water reactor (BWR) have a SLCS with the capability of injecting into the reactor pressure vessel a borated water solution at a prescribed flow rate, concentration and boron-10 isotopic enrichment. The boron in the solution absorbs neutrons, thus providing reactivity control to shut down the reactor in the event the control rods fail to insert into the core.
Section 50.62 is known as the ?ATWS Rule.
In proprietary report NEDE 31096P-A, Anticipated Transients Without Scram: Response to NRC ATWS Rule, General Electric (GE) provided guidance on modifications to SLCS to ensure licensee compliance with the ATWS Rule. The NRC approved the methods presented in NEDE 31096P-A for use by BWR licensees to demonstrate compliance with the ATWS Rule.
In response to variations in the SLCS system pressure due to changes made to SLCS operation, the NRC issued Information Notice (IN) 2001-13, Inadequate Standby Liquid Control System Relief Valve Margin. IN 2001-13 alerted licensees to NRC staff findings regarding inadequate SLCS relief valve margin. Licensees were advised to review this information for applicability to their facilities, and the IN suggested appropriate action to avoid similar problems, including evaluating their SLCS to ensure that modifications made to the systems would not cause SLCS relief valves to lift at required injection pressures, thus causing the borated water to discharge to SLCS pump recirculation, rather than to the reactor vessel.
Requirements for inservice testing of pumps and valves are contained in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 1998 edition through 2000 addenda (OM Code).
In consideration of the above regulatory guidance, the basis for this review is continued compliance with the requirements of both 10 CFR 50.62 and the ASME OM Code. The information contained in IN 2001-13 was also considered during the NRC staffs review of this license amendment request.
3.0 TECHNICAL EVALUATION
The NRC staff has reviewed the licensees regulatory and technical analyses in support of its proposed license amendment, which are described in Sections 4.0 and 5.0 of the licensees June 2, 2006, submittal.
The licensee stated that a MSSV lift setpoint tolerance change has resulted in changes to the predicted transient responses at both Dresden units. As a result, the maximum postulated pressure in the reactor vessel has increased. To ensure that the SLCS remains capable of injecting the required equivalent amount of boron-10 into the reactor vessel during an ATWS, and therefore, that the licensee remains in compliance with the ATWS Rule, changes to the SLCS are required.
On behalf of the licensee, GE performed analyses of transients to support the proposed increase in MSSV lift setpoint tolerance. The licensee concluded that the maximum postulated reactor vessel pressure at the SLCS discharge sparger is now 1301 psig. This pressure is the peak predicted during SLCS discharge in the pressure-limiting ATWS event. During review of GEs transient analyses, the NRC staff found that the results of the analyses were inadequate.
Accordingly, the NRC staff issued a request for additional information (RAI) to obtain a transient graph of reactor lower plenum pressure during the pressure-limiting ATWS scenario. The licensee provided this information to the NRC staff in its August 18, 2006, supplement. This information provided the NRC staff with the necessary data to understand the plants transient responses during the pressure-limiting ATWS event, such that the NRC staff could conclude that there was an adequate basis to support a predicted peak pressure of 1301 psig.
In order to provide the required SLCS flow, the licensee proposed to use a single-pump operation with increased boron-10 isotope enrichment to meet the requirements of the ATWS Rule. The licensee proposed to increase the boron-10 enrichment from 30 percent to 45 percent. The licensee indicated that this change is proposed to prevent the SLCS relief valves from opening during SLCS discharge in the bounding scenario.
The SLCS modifications are being requested to prevent the SLCS relief valves from lifting at times when the system must inject borated water into the reactor at its peak predicted pressure.
This request is consistent with IN 2001-13 and with GE's recommendations as presented in the licensee's submittal.
In light of the increase in predicted peak reactor vessel pressure, the NRC staff issued an RAI to clarify why the licensee was not seeking a revision to SR 3.1.7.7, SLCS Discharge Pressure, in tandem with the proposed revision to SR 3.1.7.10. The NRC staff was concerned that the required SLCS surveillance discharge pressure, 1275 psig, was inadequate to demonstrate the design-basis capability of the pump.
In its August 18, 2006, response to the RAI, the licensee referenced NUREG-1433, Standard Technical Specifications, General Electric Plants, BWR/4, Volume 2, Bases. According to NUREG-1433, the purpose of SR 3.1.7.7 is to ensure that pump performance has not degraded during the fuel cycle. The minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLCS will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component operability, trend performance, and detect incipient failures by indicating abnormal performance.
The NRC staffs position is that SR 3.1.7.7 trends the pumps performance using an established reference value and is not intended to provide reasonable assurance of the design-basis capability of the pump. Therefore, the NRC staff requested additional information seeking justification of the use of surveillance testing to assure the operational readiness of the SLCS pumps. In response, the licensee stated that based on the guidance in NUREG-1433, as discussed above, only one point on the design curve is indicative of overall pump performance.
In light of this information, the NRC staff agreed that a revision to SR 3.1.7.7 is not required.
However, as stated in NUREG-1482, Revision 1, Guidelines for Inservice Testing at Nuclear Power Plants, inservice testing assesses component performance in relation to operating characteristics when the component was known to be operating acceptably. Since the proposed change in SLCS injection pressure changes the design-basis requirement of the pump, the NRC staff requires a demonstration of the design-basis capability of the pump prior to implementation of the proposed license amendment. Such testing will provide assurance of known operational readiness, and satisfy the NRC staffs concern that the pump has not demonstrated its design-basis capability.
The NRC staff also questioned the adoption of reference values and associated tolerance bands established in the ASME OM Code that could, in some circumstances, allow the pump to degrade to a point where it would no longer be capable of meeting its design-basis injection requirement. The NRC staff requested additional information from the licensee that indicated there was adequate margin in the SLCS design to allow for the code-allowable pump degradation.
In its October 5, 2006, supplement, the licensee provided a system evaluation of the SLCS capability in accordance with NRC-approved guidance published in NEDE 31096P-A. The NRC staff reviewed this evaluation, determined that the licensee applied the guidance appropriately, and concluded that there was adequate design margin in the SLCS system to allow for a one-time design-basis demonstration of pump performance. The NRC staff finds that inservice/surveillance testing in accordance with the ASME OM Code requirements provides reasonable assurance that the pump is operationally ready. This conclusion is supported by the design-basis demonstration committed to by the licensee and by the design margin proposed by the licensee, as demonstrated by the SLCS evaluation.
The NRC staff concludes that the proposed TS SR revision is in support of acceptable modifications, and that the proposed amendment is, therefore, acceptable. Specifically, as discussed above, the NRC staff has determined that the predicted peak reactor pressure vessel pressure was calculated acceptably. Also, the requested TS SR revisions are in support of modifications consistent with NRC guidance in IN 2001-13. The test discussed below provides reasonable assurance that the increased design-basis requirements of the SLCS can be met.
The NRC staff further finds that the licensees system evaluation shows that there is adequate design margin in the SLCS to provide assurance that testing in accordance with the ASME OM Code requirements will demonstrate pump operational readiness and design-basis capability.
3.1 Commitment Prior to implementing the license amendment to increase the enrichment of sodium pentaborate used in the SLCS, the licensee will perform a one-time test of the SLCS pumps at a sufficiently high pressure to verify that the SLCS pumps meet the design-basis flow requirement. The acceptance criteria for the test will include appropriate density considerations. For the test performed during the next Dresden, Unit 2 refueling outage, the pressure will be set at 1385 psig to meet the flow requirement of 40 gallons per minute (gpm).
For the test performed during the next Dresden, Unit 3 refueling outage, the pressure will be set at 1425 psig to meet the flow requirement of 40 gpm.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (71 FR 46931; August 15, 2006). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: B. Parks Date: November 16, 2006