ML030970705
| ML030970705 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/22/2003 |
| From: | Travis Tate NRC/NRR/DLPM/LPD1 |
| To: | Bellamy M Entergy Nuclear Operations |
| Tate T, NRR/DLPM, 415-8474 | |
| References | |
| TAC MB7318 | |
| Download: ML030970705 (16) | |
Text
April 22, 2003 Mr. Mike Bellamy Site Vice President Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360
SUBJECT:
PILGRIM NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: EMERGENCY CORE COOLING SYSTEM REQUIREMENTS DURING SHUTDOWN CONDITIONS (TAC NO. MB7318)
Dear Mr. Bellamy:
The Commission has issued the enclosed Amendment No. 200 to Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station. This amendment is in response to your application dated January 23, 2003, as supplemented by letters dated February 24, and April 17, 2003. In the April 17, 2003, letter you requested that the proposed change to Note (1) of Technical Specification (TS) Table 3.2.B be withdrawn. Accordingly, this portion of the request is not considered in the U.S. Nuclear Regulatory Commission staffs safety evaluation.
This amendment revises the TS requirements for certain emergency core cooling system subsystems during shutdown conditions. Specifically, this amendment revises the core spray and low pressure coolant injection systems TS to be applicable during the Run, Startup, and Hot Shutdown Modes. This amendment also modifies the high drywell pressure instrumentation TS to require the instrumentation to be operable during the Run, Startup, and Hot Shutdown Modes. Additionally, unnecessary TS requirements are removed based on the plants operating Mode and other administrative changes are made.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.
Sincerely,
/RA/
Travis L. Tate, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-293
Enclosures:
- 1. Amendment No. 200 to License No. DPR-35
- 2. Safety Evaluation cc w/encls: See next page
Pilgrim Nuclear Power Station cc:
Resident Inspector Mr. Mike Bellamy U. S. Nuclear Regulatory Commission Site Vice President Pilgrim Nuclear Power Station Entergy Nuclear Operations, Inc.
Post Office Box 867 Pilgrim Nuclear Power Station Plymouth, MA 02360 600 Rocky Hill Road Plymouth, MA 02360 Chairman, Board of Selectmen 11 Lincoln Street Mr. C. Stephen Brennion Plymouth, MA 02360 Licensing Superintendent Pilgrim Nuclear Power Station Chairman, Duxbury Board of Selectmen 600 Rocky Hill Road Town Hall Plymouth, MA 02360-5599 878 Tremont Street Duxbury, MA 02332 Mr. Jack Alexander Manager, Reg. Relations and Office of the Commissioner Quality Assurance Massachusetts Department of Pilgrim Nuclear Power Station Environmental Protection 600 Rocky Hill Road One Winter Street Plymouth, MA 02360-5599 Boston, MA 02108 Mr. David F. Tarantino Office of the Attorney General Nuclear Information Manager One Ashburton Place Pilgrim Nuclear Power Station 20th Floor 600 Rocky Hill Road Boston, MA 02108 Plymouth, MA 02360-5599 Dr. Robert M. Hallisey, Director Ms. Jane Perlov Radiation Control Program Secretary of Public Safety Commonwealth of Massachusetts Executive Office of Public Safety Executive Offices of Health and One Ashburton Place Human Services Boston, MA 02108 174 Portland Street Boston, MA 02114 Mr. Stephen J. McGrail, Director Attn: James Muckerheide Regional Administrator, Region I Massachusetts Emergency Management U. S. Nuclear Regulatory Commission Agency 475 Allendale Road 400 Worcester Road King of Prussia, PA 19406 Framingham, MA 01702-5399 John M. Fulton Chairman Assistant General Counsel Nuclear Matters Committee Pilgrim Nuclear Power Station Town Hall 600 Rocky Hill Road 11 Lincoln Street Plymouth, MA 02360-5599 Plymouth, MA 02360
Pilgrim Nuclear Power Station cc:
Mr. William D. Meinert Nuclear Engineer Massachusetts Municipal Wholesale Electric Company P.O. Box 426 Ludlow, MA 01056-0426
April 22, 2003 Mr. Mike Bellamy Site Vice President Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360
SUBJECT:
PILGRIM NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: EMERGENCY CORE COOLING SYSTEM REQUIREMENTS DURING SHUTDOWN CONDITIONS (TAC NO. MB7318)
Dear Mr. Bellamy:
The Commission has issued the enclosed Amendment No. 200 to Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station. This amendment is in response to your application dated January 23, 2003, as supplemented by letters dated February 24, and April 17, 2003. In the April 17, 2003, letter you requested that the proposed change to Note (1) of Technical Specification (TS) Table 3.2.B be withdrawn. Accordingly, this portion of the request is not considered in the U.S. Nuclear Regulatory Commission staffs safety evaluation.
This amendment revises the TS requirements for certain emergency core cooling system subsystems during shutdown conditions. Specifically, this amendment revises the core spray and low pressure coolant injection systems TS to be applicable during the Run, Startup, and Hot Shutdown Modes. This amendment also modifies the high drywell pressure instrumentation TS to require the instrumentation to be operable during the Run, Startup, and Hot Shutdown Modes. Additionally, unnecessary TS requirements are removed based on the plants operating Mode and other administrative changes are made.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.
Sincerely,
/RA/
Travis L. Tate, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-293
Enclosures:
- 1. Amendment No. 200 to License No. DPR-35
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
J. Clifford C. Anderson, RI PUBLIC C. Raynor PDI-2 Reading T. Tate OGC G. Hill (2), T-5C3 ACRS R. Dennig S. Richards F. Arner, RI J. Bobiak, RI FAkstulewicz GThomas ADAMS Accession Number: ML030970705 TS(s): ML Package: ML
- See previous concurrence.
OFFICE PDI-2/PM PDI-2/LA SRXB RTSB OGC PDI-2/SC NAME TTate CRaynor FAkstulewicz* RDennig SCole JClifford DATE 4/22/03 4/22/03 4/10/03 4/17/03 4/22/03 4/22/03 OFFICIAL RECORD COPY
ENTERGY NUCLEAR GENERATION COMPANY ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 200 License No. DPR-35
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Entergy Nuclear Operations, Inc. (the licensee) dated January 23, 2003, as supplemented by letters dated February 24, and April 17, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations; 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; 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 Title 10 of the Code of Federal Regulations 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 3.B of Facility Operating License No. DPR-35 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 200, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
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: April 22, 2003
ATTACHMENT TO LICENSE AMENDMENT NO. 200 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 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 3/4.2-11 3/4.2-11 3/4.2-17 3/4.2-17 3/4.5-1 3/4.5-1 3/4.5-2 3/4.5-2 3/4.5-10 3/4.5-10 3/4.7-3 3/4.7-3 B3/4.5-1 B3/4.5-1 B3/4.5-2 B3/4.5-2
--- B3/4.5-2a
--- B3/4.5-2b B3/4.5-22 B3/4.5-22
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 200 TO FACILITY OPERATING LICENSE NO. DPR-35 ENTERGY NUCLEAR GENERATION COMPANY ENTERGY NUCLEAR OPERATIONS, INC.
PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
By application dated January 23, 2003, as supplemented by letters dated February 24 and April 17, 2003, Entergy Nuclear Operations, Inc. (Entergy or the licensee) requested changes to the Technical Specifications (TSs) for the Pilgrim Nuclear Power Station (Pilgrim). In the April 17, 2003, letter the licensee requested that the proposed change to Note (1) of TS Table 3.2.B be withdrawn. The supplements dated February 24, and April 17, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staffs (staff) original proposed no significant hazards consideration determination as published in the Federal Register on March 18, 2003 (68 FR 12952).
Entergy is planning to implement a design modification to the reactor vessel level instrumentation system to correct the anomalies in the water level indications that have previously occurred at Pilgrim. TS changes are needed to allow the modification to be installed during the upcoming refueling outage (RFO) 14 without an unnecessary extension in the outage duration due to instrumentation operability requirements during refueling operations. The proposed changes would revise the TS requirements for certain emergency core cooling system (ECCS) subsystems during shutdown conditions. Specifically, the proposed changes would revise the core spray (CS) and low pressure coolant injection (LPCI) systems TS requirements to be applicable during the Run, Startup, and Hot Shutdown Modes. The proposed changes would also modify the high drywell pressure instrumentation TS to require the instrumentation to be operable during the Run, Startup, and Hot Shutdown Modes.
Additional proposed changes involve removing unnecessary TS requirements based on the plants operating mode, and other administrative changes.
2.0 REGULATORY EVALUATION
The regulatory requirements for which the staff based its acceptance are Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46 and 10 CFR 50.36(c)(2)(ii). The design acceptance criteria for ECCS for light-water reactors is contained in 10 CFR 50.46. For a loss-of-coolant accident (LOCA), 10 CFR 50.46 specifies design acceptance criteria based on:
(1) the peak cladding temperature, (2) maximum cladding oxidation, (3) maximum hydrogen
generation, (4) coolable geometry, and (5) long-term cooling. Section 50.36(c)(2)(ii) of 10 CFR requires that a TS Limiting Condition for Operation be established for each item meeting one or more of four criteria. Criterion 3 of 10 CFR 50.36(c)(2)(ii) is defined as a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident (DBA) or transient that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier. The high drywell pressure instrumentation, instrumentation that initiates or controls the core and containment cooling systems, CS systems, and LPCI system are ECCS systems and components that meet Criterion 3 of 10 CFR 50.36(c)(2)(ii).
In addition to the regulatory requirements, the staff used NUREG 1433, Standard Technical Specifications, General Electric Plants, BWR/4" (STS), for determining acceptability of the proposed changes. This staff publication contains TSs for general applicability to boiling water reactor plants.
3.0 TECHNICAL EVALUATION
The staff has reviewed the licensees technical analysis in support of its proposed license amendment which is described in Section 4 of the licensees submittal. The detailed evaluation is discussed below.
3.1 ECCS Design The ECCS is designed to provide protection in the event of a LOCA due to a rupture of the primary-system piping. Although DBAs are not expected to occur during the lifetime of a plant, plants are designed and analyzed to ensure that the radiological dose from a DBA will not exceed the 10 CFR Part 100 limits. The LOCA analysis considers a spectrum of break sizes and locations, including a rapid circumferential rupture of the largest recirculation system pipe.
Assuming a single failure of the ECCS, the LOCA analyses identify the break sizes that most severely challenge the ECCS and the primary containment.
Pilgrim's ECCS consists of the high-pressure coolant injection (HPCI) system, the LPCI mode of the residual heat removal system (RHR) system, the low-pressure CS system, and the automatic depressurization system (ADS). The suppression pool serves as the source of water for each of these subsystems of the ECCS. Although not credited in the safety analyses, the condensate storage tank is capable of providing a source of water for the HPCI and CS systems.
The HPCI system (with other ECCS subsystems as backups) is designed to maintain reactor water inventory during small- and intermediate-break LOCAs, isolation transients, and loss-of-feedwater (LOFW) events. The HPCI system is designed to pump water into the reactor vessel over a wide range of reactor operating pressures. The HPCI system also serves as a backup to the reactor core isolation cooling (RCIC) system. The HPCI system is required to start and operate reliably over its design operating range. During the LOFW event and isolation transients, the RCIC system maintains water level above the top of active fuel (TAF).
For the main steam isolation valve closure event, the safety/relief valves (S/RVs) open and close as required to control pressure and the HPCI system restores water level.
The CS system is designed to provide reactor core cooling at low reactor pressures and initiates automatically in the event of a LOCA. In conjunction with other ECCS subsystems, the CS system provides adequate core cooling for all LOCA events. The system also provides spray cooling for long-term core cooling after a LOCA. The CS system consists of two independent subsystems. Each subsystem has a motor driven pump, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the reactor.
The LPCI mode of the RHR system is designed to provide reactor core cooling at low reactor pressures and is automatically initiated in the event of a LOCA. In conjunction with other ECCS subsystems, the LPCI mode is used to provide adequate core cooling for all LOCA events.
There are four motor-driven LPCI pumps with piping and valves to transfer water from the suppression pool to the reactor.
The ADS uses S/RVs to reduce reactor pressure after a small-break LOCA with HPCI failure, allowing LPCI and CS to provide cooling flow to the vessel. The plant design requires S/RVs to have a minimum flow capacity. After a delay, the ADS actuates either on low-water-level plus high drywell pressure, or on low-water-level alone.
3.2 High Drywell Pressure Instrumentation Pilgrim TS 3.2.B currently requires the instrumentation functions listed in TS Table 3.2.B to be operable whenever the supported systems are required to be operable. TS Table 3.2.B contains the instrumentation that initiates or controls the core and containment cooling systems.
Currently, the high drywell pressure instrumentation is required to be operable during all modes of operation: Run, Startup, Hot Shutdown, Cold Shutdown, and Refueling (Modes 1, 2, 3, 4 and 5, respectively). The licensee proposes a change to add a note to TS Table 3.2.B which will limit the applicability of the high drywell pressure instrumentation to be operable in the Run, Startup, and Hot Shutdown Modes only.
In its application, the licensee states that the high drywell pressure instrumentation actuates a redundant and diverse initiation signal to the reactor water level actuation instrumentation. The licensee also states that the high drywell pressure signal is actuated by the increase in the drywell pressure that results from high-energy fluid escaping from the reactor into the drywell during a LOCA, and thereby pressurizing the drywell. When the unit is in cold shutdown or refueling mode, the fluid in the reactor coolant system will be at a lower energy or the drywell may be opened. In the event of a LOCA while in these modes, pressures in the drywell would not increase enough to actuate the high drywell pressure signal. As a result, a signal would not be provided to the reactor water level actuation instrumentation. The reactor water level actuation instrumentation would still be required to be operable, with initiation on low-low reactor water level, when the associated systems are required to be operable.
The staff has reviewed the licensees analysis and determined that the proposed change is consistent with the function of the high drywell pressure instrumentation to provide an initiation signal as a result of a LOCA that results in drywell pressurization. Additionally, the staff finds the instrumentation would continue to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) for ECCS instrumentation. The staff also finds the proposed applicability is consistent with that specified in STS, Table 3.3.5.1-1, item 2.b., for high drywell pressure instrumentation. Therefore, the staff finds the proposed change acceptable.
3.3 Instrumenta0tion That Initiates or Controls the Core and Containment Cooling Systems Proposed changes to Note (1) to TS Table 3.2.B have been withdrawn.
3.4 CS and LPCI Requirements - Run, Startup, and Hot Shutdown Modes Pilgrims TSs 3.5.A.1, 3.5.A.2, 3.5.A.3, and 3.5.A.4 currently require both CS systems and the LPCI system to be operable whenever irradiated fuel is in the vessel. The licensees proposed changes involve the reformatting and rewording of these requirements to separate the requirements for each system into their own LCO. The requirements are still applicable to the CS and LPCI systems during the Run, Startup, and Hot Shutdown Modes. The proposed reformatting also includes the incorporation of new TS 3.5.A.5 to replace the existing TS 3.5.A.5 as part of the separation of the CS and LPSI requirements.
The staff has reviewed the licensees proposed reformatting related to the CS and LPCI system requirements during the Run, Startup, and Hot Shutdown Modes and concludes that no changes in the current requirements are introduced by reformatting and rewording of the existing requirement. The applicability of the CS and LPCI systems in the Cold Shutdown and Refueling Modes are addressed in Section 3.4 below. Therefore, the staff finds the proposed changes to replace the current TSs 3.5.A.1, 3.5.A.2, 3.5.A.3, 3.5.A.4, and 3.5.A.5 with the proposed reformatted version acceptable.
3.5 CS, LPCI, and Containment Cooling Requirements - Cold Shutdown and Refueling Modes The current TS 3.5.F.3 requirements allow the LPCI, CS, and containment cooling systems to be inoperable with irradiated fuel in the vessel during Cold Shutdown conditions provided no work is being done that has the potential for draining the reactor vessel. Additionally, the current TS 3.5.F.4 provides a requirement for one CS or LPCI system to be operable during refueling operations.
The proposed change would replace current TSs 3.5.F.3 and 3.5.F.4 with a requirement that two CS/LPCI subsystems be operable when the unit is in Cold Shutdown or Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than, or equal to, the 114-foot elevation. The proposed change includes a requirement to restore an inoperable CS/LPCI subsystem to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or take immediate action to suspend activities with the potential for draining the reactor vessel. Also, if both required CS/LPCI subsystems are inoperable, the proposed change would require immediate action to suspend activities with the potential for draining the reactor vessel and restore one CS/LPCI subsystem to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If one CS/LPCI subsystem cannot be restored in this condition, the proposed change requires that immediate action be taken to restore secondary containment and one standby gas treatment system to operable and to restore isolation capability in each required secondary containment penetration flow path.
The staff has reviewed the licensees analysis of the proposed changes for the CS and LPCI system requirements during Cold Shutdown and Refueling Modes. In its application dated January 23, 2003, the licensee stated that the long-term cooling analysis following a design basis LOCA demonstrates that only one low-pressure ECCS injection/spray subsystem is required, post-LOCA, to maintain adequate reactor vessel water level. Based on engineering judgment, it is reasonable to assume that while in Cold Shutdown or Refuel Modes, one
low-pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level, since the low pressure conditions for these two modes are consistent with the post-LOCA plant conditions following use of ADS. Therefore, the staff determined that the requirement to have two CS/LPCI subsystems operable provides sufficient redundancy to ensure coolant is available for the reactor core. With one required subsystem declared inoperable, the remaining operable subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. The staff recognizes that the overall system reliability is reduced in this condition because a single failure in the remaining operable subsystem concurrent with a vessel draindown event could prevent the ECCS from performing its intended safety function.
Based on engineering judgment considering the remaining available subsystem and the low probability of a vessel draindown event, the staff finds that the 4-hour completion time allows for prompt action to be taken to ensure the required cooling capacity is provided. If the inoperable subsystem cannot be restored in the required completion time, the required actions minimize any potential fission product release to the environment.
For the condition in which the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to the 114-foot elevation, the licensee states that greater than 300,000 gallons of water is provided over the fuel in the reactor vessel. In its supplement to the application dated February 24, 2003, the licensee stated that the 114-foot elevation corresponds to approximately 46.5 feet above the TAF, and approximately 21 feet above the top of the RPV flange. This provides sufficient inventory for core cooling to allow time for operator action to terminate the inventory loss prior to fuel uncovery in the event of an inadvertent draindown. The staff finds that, although a LPCI or CS system is not required to be operable in this condition, the shutdown cooling mode of the RHR system is operable during the Cold Shutdown and Refueling Modes. The RHR system can be used for restoring the reactor water level in case of a draindown event.
In response to the staffs request for additional information, the licensee, in its supplement dated February 24, 2003, stated that there are no requirements for containment cooling to be operable in the Cold Shutdown Mode. The licensee stated the containment cooling requirements may have been included in TS 3.5.F.3 because the RHR system also provides a containment cooling function when operating in the LPCI mode. However, in this mode the RHR system is considered a subsystem of the Core Standby Cooling Systems. Therefore, the reference to the containment cooling function is not necessary.
Based on the above discussion, the staff finds the proposed changes for the CS and LPCI systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) and are acceptable for ensuring sufficient inventory is available for reactor core cooling during Cold Shutdown and Refueling Modes.
Additionally, the staff finds the proposed changes consistent with the STS requirements.
3.6 Administrative Modifications The current TSs 3.5.F.2, 3.5.F.3, and 3.5.F.4 are deleted and are replaced by the proposed TSs 3.5.A.5, and 6 as discussed in Section 3.4. The current TS 3.7.A.1.n contains a cross-reference to TS 3.5.F.3, which is being deleted by the proposed change. The necessary requirements for the condition discussed in this TS are provided in TS 3.5.F.5; therefore, this change is editorial in nature and is acceptable.
3.7 TS BASES The licensee included the TS Bases pages and modifications associated with the proposed changes. The staff does not object to the proposed changes.
3.8 Conclusion The NRC staff has reviewed the licensees analysis for the proposed TS changes and concludes that the proposed changes maintain the level of protection necessary to ensure adequate coolant inventory is available in the event of a LOCA. The proposed changes ensure the acceptance criteria of 10 CFR 50.46 for ECCS are met. Therefore, the proposed TS changes are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Massachusetts State Official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change 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 (68 FR 12952). 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 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: G. Thomas Date: April 22, 2003