ML051230384
| ML051230384 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/24/2005 |
| From: | Pickett D NRC/NRR/DLPM/LPD2 |
| To: | Singer K Tennessee Valley Authority |
| Pickett , NRR/DLPM, 415-1364 | |
| References | |
| TAC MB9513, TAC MB9514 | |
| Download: ML051230384 (14) | |
Text
May 24, 2005 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 ISSUANCE OF AMENDMENTS REGARDING USE OF OPERATOR ACTION TO MITIGATE A HIGH ENERGY LINE BREAK WITH CONSEQUENTIAL LOSS OF AIR SYSTEM (TAC NOS. MB9513 AND MB9514) (SQN-TS-03-09)
Dear Mr. Singer:
The Commission has issued the enclosed Amendment No. 302 to Facility Operating License No. DPR-77 and Amendment No. 292 to Facility Operating License No. DPR-79 for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated June 5, 2003 (SQN-TS-03-09), as supplemented by letters dated June 3 and October 26, 2004.
The amendments authorize modification to the Updated Final Safety Analysis Report for both units to acknowledge credit for possible operator action to ensure that the containment design pressure is not exceeded in the event of a high energy line break inside containment with a consequential failure of the station control and service air system inside containment.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Douglas V. Pickett, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328
Enclosures:
- 1. Amendment No. 302 to License No. DPR-77
- 2. Amendment No. 292 to License No. DPR-79
- 3. Safety Evaluation cc w/enclosures: See next page
ML051230384 NRR-058 OFFICE PDII-2/PM PDII-2/LA SPSB/SC OGC PDII-2/SC NAME DPickett BClayton RDennig by memo dated KKannler EMarinos for MMarshall DATE 05 /13/ 05 05 /13/ 05 03 /22 /05 5/19/ 05 5/24/ 05
Mr. Karl W. Singer SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:
Mr. Ashok S. Bhatnagar, Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Larry S. Bryant, General Manager Nuclear Engineering Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Randy Douet Site Vice President Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 General Counsel Tennessee Valley Authority ET 11A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. John C. Fornicola, Manager Nuclear Assurance and Licensing Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Fredrick C. Mashburn Senior Program Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. Paul L. Pace, Manager Licensing and Industry Affairs ATTN: James D. Smith Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Mr. David A. Kulisek, Plant Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Senior Resident Inspector Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission 2600 Igou Ferry Road Soddy Daisy, TN 37379 Mr. Lawrence E. Nanney, Director Division of Radiological Health Dept. of Environment & Conservation Third Floor, L and C Annex 401 Church Street Nashville, TN 37243-1532 County Mayor Hamilton County Courthouse Chattanooga, TN 37402-2801 Ms. Ann P. Harris 341 Swing Loop Road Rockwood, Tennessee 37854
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 302 License No. DPR-77
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 5, 2003, as supplemented by letters dated June 3 and October 26, 2004, 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 to authorize revision of the Updated Final Safety Analysis Report (UFSAR) as set forth in the application for amendment by the licensee dated June 5, 2003. The licensee shall update the UFSAR by modifying the licensing basis to acknowledge credit for possible operator action to ensure that the containment design pressure is not exceeded in the event of a high energy line break inside containment with a consequential failure of the station control and service air system inside containment.
3.
This license amendment is effective as of its date of issuance. Implementation of the amendment is the incorporation into the next UFSAR update made in accordance with 10 CFR 50.71(e), of the changes to the description of the facility as described in TVAs application dated June 5, 2003, and evaluated in the staffs Safety Evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA E. C. Marinos for/
Michael L. Marshall, Jr., Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Date of Issuance: May 24, 2005
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 292 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 5, 2003, as supplemented by letters dated June 3 and October 26, 2004, 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 to authorize revision of the Updated Final Safety Analysis Report (UFSAR) as set forth in the application for amendment by the licensee dated June 5, 2003. The licensee shall update the UFSAR by modifying the licensing basis to acknowledge credit for possible operator action to ensure that the containment design pressure is not exceeded in the event of a high energy line break inside containment with a consequential failure of the station control and service air system inside containment.
3.
This license amendment is effective as of its date of issuance. Implementation of the amendment is the incorporation into the next UFSAR update made in accordance with 10 CFR 50.71(e), of the changes to the description of the facility as described in TVAs application dated June 5, 2003, and evaluated in the staffs Safety Evaluation attached to this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA E. C. Marinos for/
Michael L. Marshall, Jr., Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Date of Issuance: May 24, 2005
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 302 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 292 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By application dated June 5, 2003 (Agencywide Distribution and Access Management Systems accession No. ML031610858), as supplemented by letters dated June 3 (ML041670548) and October 26, 2004 (ML043010100), the Tennessee Valley Authority (TVA, the licensee) proposed amendments to operating license Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN) Units 1 and 2. These license amendments authorize changes to the Updated Final Safety Analysis Report (UFSAR) by modifying the licensing and design bases for both units to acknowledge credit for possible operator action to ensure that the containment design pressure is not exceeded in the event of a high energy line break (HELB) inside containment with a consequential failure of the station control and service air (SCSA) system inside containment. Such a consequential failure of the SCSA system could result in an uncontrolled flow of air which could potentially result in exceeding the containment design pressure. The licensees letter dated June 5, 2003, states that administrative controls are currently in place to mitigate this postulated event.
The letters dated June 3 and October 26, 2004, provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
The SCSA system is described in Section 9.3.1.2 of the SQN UFSAR. The SCSA system is a nonsafety-related system. The compressor and dryer units are located in the turbine building below plant grade. The turbine building is not a Class 1 structure. SCSA system operation is not required for safe plant shutdown or to mitigate any design-basis accidents. The SCSA system provides air to systems inside containment through a single line for each unit.
Containment isolation for this system is accomplished by an air-operated, fail-close valve and an inboard check valve. The piping downstream of the inboard check valve is TVA Class G.
This classification indicates that the piping could be damaged by a seismic event or a pipe break in another system.
Section 9.3.1.3 of the SQN UFSAR states that loss of system pressure from an accident such as a pipe break would result in the shutdown of both units if the break was not isolated before the system pressure fell below the minimum operating level.
The licensees application describes a scenario where a primary or secondary pipe breaks at a location which results in breaking the SCSA line. Failure of the outboard isolation valve to close is assumed to be the worst case single failure. As a result of this scenario, air flow into containment via the SCSA system would not be isolated. UFSAR Table 9.3.1-1 lists the operating pressure of the SCSA system as 105 pounds per square inch gauge (psig). The licensee states that the station air compressors will maintain header pressure, supplying a flow of 1500 standard cubic feet per minute into the containment. The containment design pressure is given in UFSAR Table 6.2.1-1 as 12 psig. The UFSAR currently assumes that containment integrity is maintained automatically - without operator action. To mitigate this scenario, the licensee proposes using operator action. These operator actions include: (1) identifying that the outboard containment isolation valve has not closed when the high containment pressure containment isolation setpoint has been reached, and (2) cycling the breaker to the containment isolation valve in the shutdown board room to de-energize the valve operator, which allows the air operated containment isolation valve to go to its fail-closed position. If these actions do not occur, the operator can close a manual valve upstream of the failed-open containment isolation valve. This manual valve has a chain operator and is located just inside the Unit 1 pipe chase.
However, if radiation levels do not permit access to this manual valve, operators would need to perform an emergency shutdown of the nonaccident unit and then terminate the SCSA air leak by shutting down the station air compressors.
In accordance with the licensees Corrective Action Program, the licensee prepared a problem evaluation report addressing this scenario and submitted this license amendment request to the Nuclear Regulatory Commission (NRC) to credit operator action.
3.0 REGULATORY EVALUATION
This safety evaluation addresses (1) the postulated accident scenario and the licensees analysis and proposed response, and (2) operator performance in response to the postulated accident scenario.
The following requirements and guidance are applicable:
General Design Criterion (GDC) 4, Environmental and dynamic effects design basis, requires that structures, systems and components be designed to accommodate the effects of environmental conditions associated with accidents, including possible pipe whip and discharge of fluids.
GDC 50, Containment design basis, requires that the containment be designed to accommodate the pressure and temperatures resulting from a loss-of-coolant accident (LOCA) without exceeding its design leakage rate.
The SQN UFSAR Section 6.2.4, Containment Isolation, states that containment integrity exists when a penetration, which is required to be closed, is capable of being closed by an automatic isolation valve system. The licensee has identified a scenario in which automatic isolation may not be possible.
The NRC staffs review of operator response covers operator actions, human-system interfaces, procedures, and training related to the proposed change. The NRCs acceptance criteria for operator performance are based on GDC-19, Title 10, Code of Federal Regulations (10 CFR), Sections 50.54(I) and (m), 10 CFR 50.59, 10 CFR 50.120, 10 CFR Part 55, and Generic Letter 82-33, Supplement 1 to NUREG-0737, Requirements for Emergency Response Capability. Specific regulatory review criteria are contained in Standard Review Plan Sections 13.2.1, 13.2.2, 13.5.2.1, and 18.0.
4.0 TECHNICAL EVALUATION
4.1 Accident Scenario The licensees proposed resolution of the identified scenario is to credit operator action. These actions were described in the introduction to this safety evaluation report input. In order to carry out these actions, the operator must first identify the problem. Given that the break is large enough to pressurize the containment to the Phase B containment isolation signal, at which point the SCSA system should be isolated, the operator must recognize that the air-operated valve has not closed. The licensee states that Procedure FR-Z.1, High Containment Pressure, requires the operator to verify that the containment isolation valves have closed.
This indication is available to the operator from qualified position indication status lights in the control room. If the SCSA system containment isolation valve has not closed, the licensee states that the operator would follow Procedure EA-32-3, Isolating Non-Essential Air to Containment, which directs the operator to attempt to close the outboard SCSA system containment isolation valve by cycling its breaker.
Under the worst-case accident scenario, the SCSA system outboard containment isolation valve is mechanically bound and radiological conditions preclude entry to the pipe chase to close the manual valve. Procedure EA-32-3 then directs the operator to perform an emergency shutdown of the nonaccident unit using procedure AOP-C.03, Emergency Shutdown. After shutdown of the nonaccident unit, the SCSA system leak is terminated by shutting down the station air compressors.
The licensee performed an analysis of the limiting scenario, which is the large break LOCA causing the break in the SCSA system line. Modifying the UFSAR containment peak pressure analysis to include the flow of air into containment, the licensee concluded that two hours would be available to the operator to mitigate this scenario before containment design pressure would be exceeded. Since the worst single failure is the failure of the air operated isolation valve in the SCSA system to close, the licensee credited both trains of containment spray. The licensee assumed loss of one train of safety injection and loss of off-site power. These assumptions are in compliance with the single failure criterion and are conservative.
As stated previously, loss of the SCSA system due to any cause would result in shutdown of both units. Except for the consideration of the time limitation for carrying out these actions before exceeding the containment design pressure, the most limiting outcome, shutdown of the nonaccident unit, is within the existing design basis of the plant. Therefore, while this scenario could, if carried far enough, result in shutdown of the nonaccident unit, it would be an acceptable outcome for this accident scenario.
Both SQN units have a safety-related air system, the auxiliary control air system (ACAS), which provides air to equipment which needs air to mitigate an accident or for safe shutdown. The licensee states that upon loss of pressure in the SCSA system, the SCSA is isolated from this equipment and the equipment is provided air by the ACAS. Therefore, loss of the SCSA will not 1
Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, U.S. Nuclear Regulatory Commission, NUREG 1560 Volume 2, December 1997 (Table 12.13).
affect the safety function of air-operated equipment needed to mitigate an accident or credited for safe shutdown.
With the addition of these operator actions to the licensing basis, the design-basis accidents considered in Section 6.2 of the UFSAR remain bounding.
The licensee also addressed the risk of this scenario for both the accident unit and the nonaccident unit. In both cases, as would be expected given the low probability of the initiating event and the number of additional postulated failures, the licensee concluded that the risk from this scenario is low. In addition, although the licensee states that the proposed mitigating actions will maintain the containment pressure below the design pressure of 12 psig, the containment failure pressure is five times this value.1 4.2 Human Factors Considerations The licensee provided an evaluation of the operator actions using guidance from NRC Information Notice (IN) 97-78, Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times, and American Nuclear Society/American National Standards Institute (ANS/ANSI) 58.8, Time Response Design Criteria for Safety-Related Operator Actions, which is referenced in IN 97-78. Using the criteria identified in IN 97-78, the licensee performed the following evaluation of the proposed actions.
The first criterion is indication that operator action is necessary. As discussed previously, indication that the outboard containment isolation valve has failed to close is available through qualified position indication status lights.
The next criterion is for operators to have emergency-operating procedures for the necessary actions and to receive training on the actions. The licensee indicated that procedures E-0, Reactor Trip or Safety Injection, and FR-Z.1 provide guidance to ensure that the containment isolation valves are closed. If the SCSA system containment isolation valve is not closed, FR-Z.1 directs operators to enter EA-32-3. EA-32-3 provides guidance on isolating the SCSA system leak, including cycling the breakers to the accident units outboard SCSA system containment isolation valve and closing a manual valve upstream of the containment isolation valve. If radiation levels prohibit access to the manual valves, EA-32-3 directs operators to perform an emergency shutdown of the nonaccident unit using procedure AOP-C.03, Emergency Shutdown, and then shut down the station air compressors. The licensee stated that operator training has been completed for these procedures.
Another criterion evaluated is the time available for operators to diagnose and complete the actions. According to the licensee, under the worst-case accident scenario, an SCSA system leak to containment must be stopped within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to prevent exceeding containment design pressure. The licensee provided a time line for the operator actions involved to show time available for each step. In response to the staffs request for additional information, the licensee submitted a letter dated October 26, 2004, to demonstrate that the actions in the sequence can be successfully accomplished within the allotted times. The licensee conducted walkthrough simulations on EA-32-3 with approximately two-thirds (39 out of 56) of the on-shift nonlicensed operators. The simulations included time to communicate task assignments, brief the operators, and reach the required locations.
According to the licensees time line, EA-32-3 is entered 35 minutes after a LOCA begins on the affected unit. Then 15 minutes is available to cycle the breaker to close the SCSA system outboard containment isolation valve. The licensees walkthrough simulations show that all 39 operators completed this action in less than 12.5 minutes, with an average of 8 minutes. These times included communicating the task assignment, travel time to the control room, providing briefing on the task, travel time from the control room to the location where the action is taken, and identifying and opening the circuit breaker.
Should the valve fail to close when the breaker is cycled, the next action is to close the manual isolation valve upstream of the outboard containment isolation valve, and 20 minutes is allocated for completing this task. The results of the licensees walkthrough showed that all 39 operators completed action within 16.5 minutes, with an average time of 12 minutes. These times included communicating the task assignment, providing briefing on the task, travel time to the location of the valve, identifying and operating the valve, and time for compliance with radiological precautions (e.g., obtaining dosimetry).
If radiation levels prohibit access to manual valve upstream of the stuck-open containment isolation valve, the next step is for operators to perform an emergency shutdown of the nonaccident unit. Forty minutes is allocated for this task. The licensee stated that their analysis showed that the total time needed to perform the task is 25 minutes and indicated that the ability to perform a rapid shutdown is demonstrated periodically during licensed operator training on the plant simulator. The licensee further stated that if sufficient time is not available, EA-32-3 directs operators to initiate a manual trip as necessary to meet the 2-hour time limit.
Therefore, the licensee concluded that providing timed scenarios to document the ability to meet the time limit for this step was not necessary because: this task is performed from the control room entirely, at the minimum, a 10-minute time margin is available; this task is periodically covered in operator simulator training, and the backup action is to insert a manual reactor trip if necessary to meet the time line. The staff finds these reasons to be acceptable.
The last action in the sequence is to stop the air compressors, and 10 minutes is available for this task. The licensee indicated that all 39 operators who participated in the walkthrough completed this action under 10 minutes, with an average of 6.5 minutes. These times included communicating the task assignment and travel time to the location of the air compressor controls. The licensee explained that briefing times were not included because EA-32-3 directs briefing the operators in advance, during completion of the shutdown of the nonaccident unit. EA-32-3 also has an appendix which contains, as a prudent measure, additional actions to isolate air receivers and to temporarily open vent paths to speed depressurization of the air header. The licensee explained that these actions are not considered to be time critical because the accident scenarios for this request involve an unisolated air leak inside containment.
The next criterion evaluated by the licensee is the availability of indications to show if an action performed its intended function. The licensee provided a table with indication listed for each operator action in the sequence. For cycling the breaker for the SCSA system outboard containment isolation valve, there are indicators in the control room that show valve position.
For closing the manual isolation valves upstream of the outboard containment isolation valve, procedure EA-32-3 directs the operators to seek confirmation from the technical support center (TSC) that the SCSA leak has been isolated successfully. The licensee explained that the TSC makes the determination based on plant conditions, accident progress, operating status of equipment, containment pressure, and the number of station air compressors running. Finally, for shutdown of the station air compressors, and, if necessary, depressurizing the SCSA system header, the indication is decreasing pressure in the SCSA system. During the sequence, the TSC can stop the shutdown of the nonaccident unit and depressurization of air header if it determines that nonessential air flow to the accident unit has been isolated or if the non-essential air header inside containment is intact.
Another criterion discussed by the licensee is the postaccident environmental conditions and working conditions for actions that take place outside of the main control room. The licensee provided the locations of all the actions involved, which are the vital power board rooms, the Unit 1 elevation 690-ft pipe chase, the Turbine Building, and the control room. The licensee indicated that radiological conditions may prohibit entry into the Unit 1 elevation 690-ft pipe chase following a large break LOCA. The licensee explained that this possibility is acknowledged by procedure EA-32-3, which has the option to terminate the SCSA system leak by shutting down the station air compressors as an alternative. The licencee stated that all the other areas will have acceptable environmental conditions post-large break LOCA for entry.
The next criterion evaluated is not allowing a single operator error of omission, or a single failure, to result in exceeding any limiting design requirement for the design-basis event under consideration. The licensee explained that the manual actions associated with this change are not required unless a single failure has already occurred. Specifically, the outboard containment isolation valve must fail to close before manual action is necessary to isolate an SCSA leak. Therefore, the licensee stated that an assumption of an error of commission or omission by the operators in performing the manual actions is not necessary for this request.
The staff agrees with this position.
Finally, the last criterion evaluated is the risk significance of the operator actions. As stated previously, the risk from this scenario is acceptably low.
Based on the above evaluation, the licensee determined that the criteria of NRC IN 97-78 and ANS/ANSI 58.8 have been met and, therefore, the accident scenarios described in the submittal are capable of being mitigated by the proposed operator actions.
4.3 Summary The NRC staff has reviewed the licensees request to revise the licensing and design bases to acknowledge the possibility of crediting operator actions to ensure the containment design pressure is not exceeded in the event of an HELB inside containment with a consequential failure of the SCSA system inside containment. The staff concludes that the licensee has proposed a reasonable mitigating strategy to prevent exceeding containment design pressure and has provided reasonable assurance that the allocated operator action times are sufficient for operators to successfully perform the required tasks. In addition, the staff finds that the licensee has given adequate consideration to indications available to the operators, the postaccident environmental conditions, procedures needed for implementing the sequence of actions, and training on the tasks. The staff further concludes that the licensee will continue to meet the applicable requirements of GDC-4, GDC-19, GDC-50, 10 CFR 50.54(I) and (m),
10 CFR 50.59, 10 CFR 50.120, and 10 CFR Part 55. Therefore, the staff finds this request acceptable.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 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 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 (68 FR 37584). 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.
7.0 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 these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Richard Lobel June Cai Dated: May 24, 2005