ML040550107
| ML040550107 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/05/2004 |
| From: | Dan Collins NRC/NRR/DLPM/LPD1 |
| To: | Bakken A Public Service Enterprise Group |
| Boska J, NRR, 301-415-2901 | |
| Shared Package | |
| ML040550110 | List: |
| References | |
| TAC MC1436 | |
| Download: ML040550107 (14) | |
Text
October 5, 2004 Mr. A. Christopher Bakken, III President & Chief Nuclear Officer PSEG Nuclear LLC-X04 Post Office Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE:
REMOVAL OF REACTOR VESSEL HEAD SPRAY PIPING (TAC NO. MC1436)
Dear Mr. Bakken:
The Commission has issued the enclosed Amendment No. 152 to Facility Operating License No. NPF-57 for the Hope Creek Generating Station. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 17, 2003, and supplemented by letters dated July 15, and August 23, 2004. The amendment revises the TSs to delete the primary containment isolation valves and instrumentation associated with the permanent removal of the reactor vessel head spray piping.
A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Daniel S. Collins, Sr. Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-354
Enclosures:
- 1. Amendment No. 152 to License No. NPF-57
- 2. Safety Evaluation cc w/encls: See next page
Hope Creek Generating Station cc:
Mr. John T. Carlin Vice President - Nuclear Assessment PSEG Nuclear - N10 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. David F. Garchow Vice President - Engineering/Technical Support PSEG Nuclear - N28 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Michael Brothers Vice President - Site Operations PSEG Nuclear - N10 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. James A. Hutton Plant Manager PSEG Nuclear - X15 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Steven Mannon Acting Manager - Nuclear Safety and Licensing PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Ms. R. A. Kankus Joint Owner Affairs Exelon Generation Company, LLC Nuclear Group Headquarters KSA1-E 200 Exelon Way Kennett Square, PA 19348 Lower Alloways Creek Township c/o Mary O. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Brian Beam Board of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Hope Creek Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038
October 5, 2004 Mr. A. Christopher Bakken, III President & Chief Nuclear Officer PSEG Nuclear LLC-X04 Post Office Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE:
REMOVAL OF REACTOR VESSEL HEAD SPRAY PIPING (TAC NO. MC1436)
Dear Mr. Bakken:
The Commission has issued the enclosed Amendment No. 152 to Facility Operating License No. NPF-57 for the Hope Creek Generating Station. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 17, 2003, and supplemented by letters dated July 15, and August 23, 2004. The amendment revises the TSs to delete the primary containment isolation valves and instrumentation associated with the permanent removal of the reactor vessel head spray piping.
A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Daniel S. Collins, Sr. Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-354
Enclosures:
- 1. Amendment No. 152 to License No. NPF-57
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION JClifford OGC GMeyer, RGN-I PUBLIC DCollins ACRS PDI-2 Reading CRaynor TBoyce GHill (2)
RDennig Adams Accession Numbers: Package: ML040550110, Letter: ML040550107, TS(s): ML
- See Safety Evaluation dated January 24, 2003 OFFICE PDI-2/PM PDI-2/LA SRXB/SC SPSB/SC EMEB/SC IROB/SC OGC PDI-2/SC NAME DCollins CRaynor FAkstulewicz RDennig KManoly TBoyce HMcGurren Rlaufer for JClifford DATE 9/21/04 9/21/04 4/24/04 9/17/04 9/14/04 9/20/04 9/30/04 10/01/04 OFFICIAL RECORD COPY
PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 152 License No. NPF-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by PSEG Nuclear LLC dated November 17, 2003, and supplemented July 15, and August 23, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions regulations set forth in 10 CFR Chapter I; 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 Commissions 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 Facility Operating License No. NPF-57 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 152, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented prior to restart from the fall 2004 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA by Richard J. Laufer for/
James W. Clifford, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 5, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 152 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3/4 3-21 3/4 3-21 3/4 3-78 3/4 3-78 3/4 4-13 3/4 4-13 3/4 6-21 3/4 6-21
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 152 TO FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354
1.0 INTRODUCTION
By letter dated November 17, 2003, and supplemented by letters dated July 15, and August 23, 2004 (ADAMS Accession Nos. ML033360696, ML042100229, and ML042450535),
PSEG Nuclear LLC (PSEG or the licensee) requested changes to the Hope Creek Generating Station (Hope Creek) Technical Specifications (TSs). The supplements dated July 15, and August 23, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 20, 2004 (69 FR 2746).
The proposed changes would revise the TSs to delete certain primary containment isolation valves and instrumentation associated with the reactor vessel (RV) head spray piping. These requested changes would support permanent removal of the RV head spray piping from Hope Creek. Also, on TS page 3/4 3-21, the nomenclature for the valves remaining in the list would be expanded to provide the full system nomenclature.
Each refueling outage before the reactor pressure vessel (RPV) head can be removed, the head spray piping has to be disconnected from the RPV head at a flanged connection; then, at the end of the outage, it is reinstalled and tested. The application states that the removal and reinstallation activities result in additional personnel radiation dose and add several hours to the critical path for the refueling outage. Permanent removal of the reactor head spray line would reduce personnel exposure, save refueling outage time and cost, and eliminate the local leak rate testing and motor operated valve testing associated with the reactor head spray line containment penetration.
2.0 REGULATORY EVALUATION
The reactor head spray line is part of the residual heat removal (RHR) system. During the initial phase of the shutdown cooling mode of the RHR system, a small portion (1000 gpm) of the discharge flow can be directed to the head spray nozzle in the RV head. The head spray flow condenses steam generated from the RV walls while the RV is being filled with water, thereby keeping reactor pressure down during initial phases of the shutdown cooling mode. It also prevents temperature stratification in the RV. Its design objective is to assure that the vessel head cools in parallel with the remainder of the vessel metal below the waterline during the filling of the RV.
The head spray line was provided in the original RHR system design, because it was anticipated that vessel cooldown and head removal would be critical path activities and that rapid head cooldown would reduce outage time. Operating experience at many boiling water reactor (BWR) plants has indicated that vessel head cooldown is not on the critical path since critical path is devoted to parallel activities such as shield block removal and main steamline isolation valve leak rate testing. Although the head spray provides the intended function, the incentive for its inclusion in the RHR design may not be realized where head cooldown is not on the critical path. As a result, many BWR operators do not use the head spray during the shutdown cooling mode.
The reactor head spray line at Hope Creek runs from outside the primary containment (also known as the drywell) through the primary containment steel shell and reactor cavity seal plate and connects to the RPV head. The primary containment at Hope Creek is a steel and concrete pressure vessel of the General Electric Mark I design. Although the head spray function is not a safety-related function, the head spray piping and components have safety-related functions of maintaining the reactor coolant pressure boundary (RCPB) integrity.
Additionally, valves within the line perform containment isolation functions.
The regulatory requirements that the Nuclear Regulatory Commission (NRC or the Commission) staff considered in its review of this amendment application include those contained in:
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, General Design Criteria (GDC) for Nuclear Power Plants Criterion 14, Reactor coolant pressure boundary, GDC Criterion 50, "Containment design basis,"
GDC Criterion 52, "Capability for containment leakage rate testing,"
GDC Criterion 53, "Provisions for containment testing and inspection,"
GDC Criterion 55, Reactor coolant pressure boundary penetrating containment, 10 CFR 50.55a, Codes and Standards, subparagraph (c), Reactor Coolant Pressure Boundary; 10 CFR Part 50 Appendix J, Primary Containment Leakage Testing for Water Cooled Power Reactors; and Hope Creek TS 6.8.4.f, which is not being changed by this amendment, The NRC staff also considered the guidance contained in:
Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, NEI [Nuclear Energy Institute] 94-01, Revision 0, dated July 26, 1995, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J.
ANSI/ANS-56.8-1994, Containment System Leakage Testing Requirements.
The NRC staffs evaluation of the proposed physical modifications to Hope Creek and the requested TS changes relative to these regulatory requirements are discussed further in Section 3.0 of this safety evaluation.
3.0 TECHNICAL EVALUATION
3.1 Scope of the Modification In its application, the licensee states that deletion of the head spray piping will involve removal of the Class 1 process piping for the rector vessel head spray line, which is part of the RCPB, on both sides of primary containment penetration (P10). The primary containment isolation valves for P10 (F022 and F023) and the associated valve power, controls, and flow instrumentation will also be removed. The RCPB will be established by the installation of a bolted circular metal plate, known as a blank flange, to the existing flanged connection at the RV head. A short section of the existing process piping will be left within the P10 containment penetration and pipe caps will be welded onto each end of that pipe segment. Additionally, the relief valve for the P10 Class 1 piping, which is installed on a 3/4-inch line, will be removed such that the 3/4-inch line will remain as a test connection. That test connection will have a normally-installed blind flange as a foreign material exclusion boundary. The reactor head insulation package will be restored after removal of the 6-inch pipe, pipe supports, and check valve.
The proposed changes to the Hope Creek TSs are:
1.
Valves HV-F022 and HV-F023 will be removed from Table 3.3.2-1, "Isolation Actuation Instrumentation." The system designator "1BC" will be added to the valves remaining in the table, which is an administrative change done for clarity.
2.
Valves 1BC-HV-F022 and 1BC-HV-F023 will be removed from Table 3.3.7.4-2, "Remote Shutdown Systems Controls."
3.
Valves BC-V021 and BC-V020 will be removed from Table 3.4.3.2-1, "Reactor Coolant System Pressure Isolation Valves."
4.
Item 3.(d) is removed from Group 3 of Table 3.6.3-1, "Primary Containment Isolation Valves." Item 3.(e) will be renumbered as 3.(d).
3.2 Compliance with 10 CFR Part 50 By letter dated June 14, 2004, the NRC staff requested additional information from PSEG regarding how Hope Creek will meet the applicable requirements of 10 CFR Part 50 following the proposed modifications. PSEGs response was provided in the July 15, and August 23, 2004, supplements. Information provided and the NRC staffs evaluation of that information is discussed below.
a.
GDC 14, Reactor Coolant Pressure Boundary, and 10 CFR 50.55a, Codes and Standards GDC 14 requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. In addition, 10 CFR 50.55a(c) contains requirements that the RCPB must meet relative to the American Society of Mechanical Engineers (ASME) design code Edition and Addenda. Specifically, RCPB components must meet the requirements for ASME Section III, Class 1. This regulation further specifies, in 10 CFR 50.55a(c)(4), that:
For a nuclear power plant whose construction permit was issued prior to May 14, 1984 the applicable Code Edition and Addenda for a component of the reactor coolant pressure boundary continue to be the Code Edition and Addenda that were required by Commission regulations for such component at the time of issuance of the construction permit.
The construction permit for Hope Creek was issued on November 4, 1974. The requirements of 10 CFR 50.55a(c) that were in effect at that time state:
(2) For construction permits issued on or after January 1, 1971, pressure vessels which are part of the reactor coolant pressure boundary shall meet the requirements for Class A or Class 1 vessels set forth in editions of Section III of the ASME Boiler and Pressure Vessel Code and Addenda in effect on the date of order of the pressure vessel: provided, however, that if the pressure vessel is ordered more than 18 months prior to the date of issuance of the construction permit, compliance with the requirements for Class A or Class 1 vessels set forth in editions of Section III of the ASME Boiler and Pressure Vessel Code and Addenda in effect 18 months prior to the date of issuance of the construction permit is required....
For Hope Creek, the applicable construction code for the RCPB is ASME Section III, Class 1, 1968 Edition, Winter 69 Addenda. Hence, any new RCPB components installed under the proposed modification must meet the requirements of this construction Code.
The Hope Creek RV head was designed and manufactured with three flanged nozzle connections: head vent, head spray, and a spare. The proposed modification will remove the head spray connection and install a blind flange on the RV head. The application states that the blind flange to be installed is designed to ASME Section III, Class 1, 1968 Edition, Winter 69 Addenda.
In its August 23, 2004 supplement, the licensee clarified that this replacement flange installation will be subjected to a post-modification leakage test at normal reactor operating temperature and pressure to ensure the integrity of the bolted connection.
The NRC concludes that the installation of a blind flange designed to the original construction code and leak tested at normal reactor operating temperature and pressure is acceptable for ensuring the integrity of the RCPB. Therefore, the NRC staff finds that the proposed modification will comply the requirements of GDC 14 and 10 CFR 50.55a.
b.
GDC 50, Containment Design Basis; 10 CFR Part 50 Appendix J, Primary Containment Leakage Testing for Water Cooled Power Reactors; and Hope Creek TS 6.8.4.f, Primary Containment Leakage Rate Testing Program GDC 50 requires, in part, that reactor containment structures be designed so that the containment and its internal components can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure, Pa, and temperature conditions resulting from a design basis loss of coolant accident.
Section 6.2.1.1.3.1 of the Hope Creek Updated Final Safety Analysis Report states that the containment drywell design pressure is 62 psig and Pa is 48.1 psig.
Appendix J to 10 CFR Part 50 contains requirements for periodic leakage testing of the primary reactor containment. Per this appendix, licensees are required to perform periodic local leakage rate tests (LLRTs) of containment isolation valves to ensure satisfactory performance. Additionally, periodic integrated leakage rate tests (ILRTs) are required to ensure that the overall containment system (i.e., structure and isolation valves) perform adequately to prevent release of quantities of radioactive material that would have a significant radiological effect on the health of the public.
Hope Creek TS 6.8.4.f requires a primary containment leakage rate testing program that is in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. The regulatory position in RG 1.163 states that NEI 94-01, Revision 0, dated July 26, 1995, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, prepared by NEI, provides methods acceptable to the NRC staff with certain exceptions, stated in the RG.
NEI 94-01, in turn, references ANSI/ANS-56.8-1994, Containment System Leakage Testing Requirements, which specifies in Section 3.2.5 that systems open to the containment atmosphere shall be vented during periodic ILRTs.
The licensee stated in its July 15, and August 23, 2004, supplements that the post-modification configuration of penetration P10 will be such that the outboard pipe cap will meet the requirements of ASME Section III, Class 2, and its associated welds will provide an ASME Section III, Class MC containment boundary. The post-modification testing will include a 48.1 psig air test on the volume contained by the spared pipe within the P10 penetration and the newly installed pipe caps. The application further states that as-found LLRT will be performed on the reactor head spray containment isolation valves before they are removed and that containment penetration P10 will be vented during periodic ILRTs to ensure that the Class MC welds are subjected to the ILRT test pressure.
The NRC staff concludes that the proposed 48.1 psig post-modification pressure test will subject the new Class MC welds at the pipe caps to Pa. This is sufficient to ensure that the requirements of GDC 50 will be met, and is consistent with the testing requirements of 10 CFR Part 50 Appendix J. Furthermore, because the containment isolation valves in penetration P10 will be removed and replaced with passive components (i.e., pipe caps), the NRC staff finds that it is acceptable for penetration P10 to be removed from the LLRT program and tested under the ILRT program. Venting of penetration P10 to the containment atmosphere during the ILRT as set forth in the application will expose the pipe cap and Class MC weld that become the new containment boundary for penetration P10 to the ILRT test pressure as specified in the guidance of ANSI/ANS-56.8-1994. Hence, the requirements of TS 6.8.4.f, Primary Containment Leakage Rate Testing Program, will continue to be met.
c.
GDC 52, "Capability for containment leakage rate testing; and GDC Criterion 53, "Provisions for containment testing and inspection GDC 52 requires that the reactor containment be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure. GDC 53 requires that the reactor containment be designed to permit (1) appropriate periodic inspections of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations that have resilient seals and expansion bellows.
The proposed modification provides a vent connection for penetration P10 inside containment that will ensure that the Class MC containment boundary will be exposed to the ILRT test pressure. Additionally, the application states that the penetrations expansion bellows will not be altered by the proposed modification and it will continue to be tested by the periodic ILRT. Hence, the NRC staff concludes that the post-modification configuration will meet the requirements of GDC 52, and the proposed modification does not affect Hope Creeks compliance with GDC 53.
d.
GDC Criterion 55, Reactor coolant pressure boundary penetrating containment, GDC 55 contains requirements for containment isolation valves in RCPB lines that penetrate containment. Because the proposed modification will permanently remove the RV head spray line penetrating containment and its associated containment insolation valves, GDC 55 will no longer apply.
3.3 Plant Safety and Operations Head spray has no active safety-related function and no credit has been taken for its use in current safety analyses or emergency operating procedures. Removal of the head spray line is beneficial in that it removes a potential location for a primary system pipe break. The need for head spray and the reduction in line break probability has little effect on overall risk.
Consequently, the head spray removal has no significant impact on plant safety.
The functions of venting the head and removal of radioactive gases from the vessel head are performed by the head vent system, which in some plants is connected to the head spray line.
For Hope Creek, the head spray is entirely separate from the head vent, so that head spray removal has no impact on the head venting function.
3.4 Vessel Thermal Duty The loads on the vessel head flange and studs depend on the cooldown rate of the RV flange and head. Use of head spray can promote rapid cooling of the vessel head, which may result in an increased thermal duty on the vessel flange and head. However, the use of head spray has been considered in the thermal duty design of the vessel and the vessel has been analyzed for a cooldown rate conservatively bounding the maximum shutdown cooling rate of 100 F/hr with or without the head spray.
The removal of head spray eliminates the potential for rapid cooling of the vessel head, which will reduce the thermal duty on the vessel flange and head during a normal shutdown cooling evolution.
3.5 Maintenance/Maintainability Removal of the head spray pipe and components will decrease system maintenance requirements and reduce in-service inspection work. In addition, it should no longer be necessary to conduct isolation valve surveillance tests or to maintain the valves. Plant maintainability will be enhanced since removal of equipment, particularly in the drywell, will provide additional space which will enhance the maintainability of the other components in their immediate vicinity.
3.6 Plant Personnel Radiation Exposure Removal of the head spray pipe connected to the RV will reduce radiation exposure since personnel will no longer remove and install the head spray pipe each refueling outage.
Additional exposure reduction is expected because removal of the head spray pipe and components will result in the reduction in the required in-service inspection of the head spray line and the elimination of surveillance testing and maintenance on the isolation valves.
3.7 Conclusion The NRC staff has completed its review of the proposed changes to the Hope Creek TSs and finds that the RPV head spray portion of the RHR system has no active safety-related function.
Following the removal of the head spray piping, the plant systems will remain in compliance with the licensing basis, specifically GDC 14, 50, 52, 53, 55, and other applicable requirements of 10 CFR 50.55a, 10 CFR Part 50 Appendix J, and the Hope Creek TSs. Therefore, the NRC staff concludes that the licensees request to remove the RV head spray portion of the RHR system and revise the TSs accordingly is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey State Official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (69 FR 2746). Accordingly, the amendment meets 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 amendment.
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 Commissions 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.
Principal Contributors: T. Ford J. Boska D. Collins Date: October 5, 2004