ML022480469
| ML022480469 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 09/18/2002 |
| From: | Pulsifer R NRC/NRR/DLPM/LPD1 |
| To: | Thayer J Vermont Yankee |
| Pulsifer R M, NRR/DLPM, 415-3016 | |
| References | |
| TAC MB4610 | |
| Download: ML022480469 (14) | |
Text
September 18, 2002 Mr. Jay K. Thayer Site Vice President - Vermont Yankee P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: MAIN STEAM LINE RADIATION MONITOR (TAC NO.
MB4610)
Dear Mr. Thayer:
The Commission has issued the enclosed Amendment No. 212 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station, in response to your application dated March 19, 2002, as supplemented on June 4, July 16 and 24, August 22, and September 4, 2002.
On July 31, 2002, Vermont Yankee Nuclear Power Corporations (VYNPC) interest in the license was transferred to Entergy Nuclear Vermont Yankee, LLC (ENVY) and Entergy Nuclear Operations, Inc. (ENO). On August 6, 2002, ENO requested that the U.S. Nuclear Regulatory Commission (NRC) continue to review and act on all requests before the Commission which had been submitted by VYNPC before the transfer. Accordingly, the NRC staff has acted upon the request. The July 16 and 24, August 22, and September 4, 2002 supplements were within the scope of the original application and did not change the staffs proposed no significant hazards consideration determination.
The amendment changes the Technical Specifications by removing of the existing scram function and Group 1 isolation valve closure functions of the Main Steam Line Radiation Monitors (MSLRM). An explicit requirement for periodic functional test and calibration of the MSLRM is added to maintain operability of the mechanical vacuum pump isolation function.
J. Thayer A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Robert M. Pulsifer, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosures:
- 1. Amendment No. 212 to License No. DPR-28
- 2. Safety Evaluation cc w/encls: See next page
Vermont Yankee Nuclear Power Station cc:
Regional Administrator, Region I Mr. Raymond N. McCandless U. S. Nuclear Regulatory Commission Vermont Department of Health 475 Allendale Road Division of Occupational King of Prussia, PA 19406 and Radiological Health 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402 Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Mr. Gautam Sen Washington, DC 20037-1128 Manager, Licensing Entergy Nuclear Vermont Yankee, LLC Ms. Christine S. Salembier, Commissioner P.O. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 112 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATTN: James Muckerheide P.O. Box 116 400 Worcester Rd.
Vernon, VT 05354-0116 Framingham, MA 01702-5399 Mr. Michael Hamer Jonathan M. Block, Esq.
Operating Experience Coordinator Main Street Entergy Nuclear Vermont Yankee, LLC P. O. Box 566 P.O. Box 250 Putney, VT 05346-0566 Governor Hunt Road Vernon, VT 05354 Mr. Michael R. Kansler Sr. Vice President and Chief Operating G. Dana Bisbee, Esq. Officer Deputy Attorney General Entergy Nuclear Operations, Inc.
33 Capitol Street Mail Stop 12A Concord, NH 03301-6937 440 Hamilton Ave.
White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General Mr. John J. Kelly One Ashburton Place, 19th Floor Director, Licensing Boston, MA 02108 Entergy Nuclear Operations, Inc.
440 Hamilton Avenue Ms. Deborah B. Katz White Plains, NY 10601 Box 83 Shelburne Falls, MA 01370
J. Thayer A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Robert M. Pulsifer, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosures:
- 1. Amendment No. 212 to License No. DPR-28
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC JAndersen TClark RDennig PDI-2 R/F RPulsifer OGC ACRS SRichards MReinhart GHill (2) CAnderson, RI EMarinos Accession Number: ML022480469
- See previous concurrence
- Input received 9/5/02; no major changes made OFFICE PDI-2/PM PDI-2/LA EEIB
- SPSB ** OGC* PDI-2/SC (A)
NAME RPulsifer TClark EMarinos MReinhart CBray JAndersen DATE 9/16/02 9/17/02 9/5/02 9/5/02 9/10/02 9/18/02 OFFICIAL RECORD COPY
ENTERGY NUCLEAR VERMONT YANKEE, LLC ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 212 License No. DPR-28
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by the Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., (the licensees) dated March 19, 2002, as supplemented on June 4, July 16 and 24, August 22, and September 4, 2002, 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; 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 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, 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 its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James W. Andersen, Acting 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: September 18, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 212 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 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 22 22 24 24 25 25 26 26 27 27 28 28 31 31 35 35 36 36 37 37 45 45 48 48 64 64 74 74 76 76 79 79
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 212 TO FACILITY OPERATING LICENSE NO. DPR-28 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271
1.0 INTRODUCTION
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated March 19, 2002, Vermont Yankee Nuclear Power Corporation (the then licensee) requested changes to the Technical Specifications (TSs) for the Vermont Yankee Nuclear Power Station (VY). Additional information was supplied by letters dated June 4, July 16 and 24, August 22, and September 4, 2002. On July 31, 2002, Vermont Yankee Nuclear Power Corporations (VYNPC) interest in the license was transferred to Entergy Nuclear Vermont Yankee, LLC (ENVY) and Entergy Nuclear Operations, Inc. (ENO). On August 6, 2002, ENO requested that the U.S. Nuclear Regulatory Commission (NRC) continue to review and act on all requests before the Commission which had been submitted by VYNPC before the transfer. Accordingly, the NRC staff has acted upon the request. The July 16 and 24, August 22, and September 4, 2002 supplements were within the scope of the original application and did not change the staffs proposed no significant hazards consideration determination.
The proposed license amendment would make revisions to eliminate the reactor scram and main steamline isolation valve closure requirements associated with the main steamline radiation monitors (MSLRMs). The proposed changes are based on the methodology of the NRC-approved Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report (LTR), NEDO-31400A, Safety Evaluation for Eliminating the Boiling Water Reactor Main Steamline Isolation Valve Closure Function and Scram Function of the Main Steamline Radiation Monitor, October 1992. The elimination of the trip functions identified in the LTR will result in a reduced potential for inadvertent reactor shutdowns and plant transients caused by spurious MSLRM actuation signals. The alarms initiated by the MSLRMs and the mechanical vacuum pump (MVP) trip function would remain functional.
The design intent of the MSLRMs was to provide an early indication of gross fuel failure. The MSLRMs measure the radiation level external to the main steamlines and provide alarm, trip, and isolation functions upon detection of excessive radiation levels. In the original design, the monitors provided signals to initiate a reactor scram and main steam isolation valve (MSIV) closure. At some facilities, these monitors also provided trip signals to other plant equipment.
The MSLRMs have been responsible for several unnecessary reactor scrams due to instrument failures, maintenance errors, and chemistry excursions. In none of these cases was there failed fuel.
NEDO-31400A provides a safety assessment that demonstrates that the MSIV isolation and scram functions of the MSLRMs are not required to ensure compliance with the accident dose guidelines of Title 10 of the Code of Federal Regulations (10 CFR) Part 100. The control rod drop accident (CRDA) is the only design-basis accident (DBA) for which it is assumed that the main steamline isolation signal comes from the MSLRMs. No design-basis event, including the CRDA, credits reactor scram initiated by the MSLRMs. The LTR safety assessment considered two mutually exclusive scenarios for the CRDA. The first scenario assumed isolation of the main steamlines occurred and the fission product activity was released by leakage from the main condenser to the environment. In the second scenario it was assumed that no automatic isolation of the main steamlines occurred and the activity was transported to an augmented offgas system. The release of this activity to the environment would be from the normal offgas release point after holdup in the treatment system. Calculations of the radiological consequences at the exclusion area boundary (EAB) were performed for each scenario for comparison with the applicable dose limits. By letter dated May 15, 1991, the NRC staff accepted the LTR for reference in license applications.
To provide for the proposed elimination of the MSLRMs scram and MSIV isolation functions, the licensee proposed changes to requirements in TS Table 3.1.1, Reactor Protection System (Scram) Instrument Requirements; TS Table 4.1.1, Scram Instrumentation and Logic Systems Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation, Logic Systems and Control Circuits; TS Table 4.1.2, Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Channels; TS 3.2.F, Mechanical Vacuum Pump Isolation; TS 4.2.F, Mechanical Vacuum Pump Isolation; TS Table 3.2.2, Primary Containment Isolation Instrumentation; and TS Table 4.2.2, Minimum Test and Calibration Frequencies - Primary Containment Isolation Instrumentation. The licensee also proposed associated TS Bases changes.
2.0 REGULATORY EVALUATION
The licensee must show, and the staff must find acceptable, that for DBAs the plant continues to meet dose limits given in 10 CFR Part 100 and applicable dose criteria in applicable sections of NUREG-0800, Standard Review Plan (SRP), Chapter 15, for offsite doses, and 10 CFR Part 50, Appendix A, GDC-19 with respect to control room habitability. The NRC staff used the LTR NEDO-31400A and its safety evaluation, as well as SRP Chapter 15.4.9, Appendix A, Radiological Consequences of Control Rod Drop Accident (BWR), to aid in its review of the proposed changes.
3.0 TECHNICAL EVALUATION
In a boiling water reactor (BWR), control rods are positioned into the core from beneath the reactor vessel by the control rod drive mechanisms (CRDMs) and associated systems and circuitry. A CRDA is the result of an unlikely failure in this equipment that allows a control rod to drop out of the core. This event would cause rapid increase in the local core power in the fuel adjacent to the dropped rod. As a result of this power increase, some fuel damage is postulated to occur. The radiological consequences of such an event are analyzed to ensure that the postulated offsite doses are well within the 10 CFR Part 100 dose limits. Well within is defined in SRP 15.4.9 as 25 percent of the limits.
3.1 Comparison of Vermont Yankee Site-Specific Parameters to NEDO-31400A Safety Assessment Assumptions The NRC staff reviewed the licensees evaluation of the applicability of the NEDO-31400A safety assessment to Vermont Yankee. With the exception of the atmospheric dispersion factor values (X/Qs), the NRC staff was able to confirm that the values used in the NEDO-31400A analysis do bound the values for VY by using the information provided by the licensee in its submittal and in the current VY Updated Final Safety Analysis Report (UFSAR).
The staff requested additional information on the licensees determination of the X/Q values and by letter dated July 24, 2002, the licensee provided this information.
The LTR assumed a power level of 0.12 MW/rod, which includes a peaking factor of 1.5. The licensees analyses of record assume 0.113 MW/rod for 8 x 8 fuel and 0.0917 MW/rod for 9 x 9 fuel, both of which are bounded by the LTR assumption. The LTR assumed that 850 rods of 8 x 8 fuel fail and release the fission product inventory in the fuel rod gap to the reactor coolant.
The VY analyses assumed 850 rods of 8 x 8 fuel or 1,000 rods of 9 x 9 fuel fail. The fission product inventory from 1,000 rods of 9 x 9 fuel is slightly lower than that from 850 rods of 8 x 8 fuel, so the LTR analysis assumption for fuel failure is the same or higher than that for VY. The LTR analyses assumed that a fraction of the failed rods reached temperatures sufficient to cause fuel melt. VYs current UFSAR analysis shows that this fuel melt does not occur, and the VY radiological analysis of the CRDA does not assume this fuel melt source term. The LTR assumption bounds the VY assumption for fuel melt.
Both the LTR and VY analyses assume 10 percent of the iodines and noble gases in the core are in the fuel rod gap and are available for release to the coolant. Of the radioactive material released to the reactor coolant, 100 percent of the noble gases and 10 percent of the iodines are assumed to reach the main condenser in both sets of analyses. The LTR assumed 90 percent of the iodines that reach the condenser are retained in the condenser, while the VY analyses assume that 50 percent are retained. This is more conservative than the LTR assumption. VY has shown that the assumptions used in the NEDO-31400A safety assessment bound the conditions at VY for the CRDA with the MSLRM scram and isolation functions removed.
3.1.1 Atmospheric Dispersion Factors addressed by NEDO-31400A In the NEDO-31400A bounding safety assessment of the CRDA, two scenarios were addressed. The first scenario assumed MSIV closure with an X/Q value of 2.5x10-3 s/m3 (seconds/meter3). The second scenario assumed no MSIV closure and a value of 3.0x10-4 s/m3. Both values were applied for the 2-hour duration of the dose assessment. To demonstrate that these values were bounding for VY, the licensee calculated 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> values for the exclusion area boundary (EAB) using site-specific inputs. Since only the EAB dose was calculated in the LTR, only the X/Q estimates for the EAB were evaluated.
When making the confirmatory calculations for the first scenario, the licensee used onsite meteorological data collected during calendar year 1989. In response to a request from the staff to provide evidence that the 1989 data were representative of long-term conditions, the licensee provided 4 years of additional data from 1990 through 1993. Wind direction and speed were measured at approximately 35 and 295 feet above grade and the temperature difference intervals used to estimate atmospheric stability were measured between approximately 200 and
35 feet and 295 and 35 feet above grade. The UFSAR states that the meteorological measurement program meets the intent of Regulatory Guide (RG) 1.23, Onsite Meteorological Programs.
The NRC staff performed a review of the 5 years of meteorological data measured at the three levels using the methodology described in NUREG-0917, Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data. Further review was performed using a computer spreadsheet with staff noting the following. Data recovery during 1989 was greater than 95 percent for each measured parameter and well above the 90 percentile recommended by RG 1.23 throughout the 5-year period. Examination of the data revealed some occurrence of wind speed or direction remaining unchanged for 2 or more consecutive hours, but such occurrences were of short duration, occurred relatively infrequently, and should have an insignificant effect on the calculations for this dose assessment. With respect to atmospheric stability measurements, the length and time of occurrence of stable and unstable atmospheric conditions appeared very good. Stable and neutral conditions were consistently reported to occur at night and unstable and neutral conditions during the day. The longest continuous occurrence of a single unstable category was less than 6 consecutive hours, with two exceptions, which is consistent with expected meteorological conditions. Wind direction frequency occurrence at both the 35-foot and 295-foot levels were very similar from year to year throughout the 5-year period. While the 295-foot level showed more distinctive bimodal flow along the axis of the river valley in which VY is located, winds at both heights predominated from the north northwest and generally south southeast sectors. The lower level experienced secondary winds from the generally westerly quadrants toward the river to the east of the site.
Based on this review, the data appear to be of high quality with little variability during the 5-year period. Thus, the 1989 data appear to be representative of long-term conditions and to provide a suitable base for the ground level release X/Q estimate used in this dose assessment.
For the first scenario the licensee calculated X/Q values assuming a ground level release from the turbine building using site-specific inputs and a computer code based upon the RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, methodology. The EAB distance was varied by direction. The resultant limiting 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> X/Q value of 1.7x10-3 s/m3 occurred in the south-south-west (SSW) sector. For the second scenario the licensee made calculations assuming an effluent release from the plant stack, with an effective stack height of 91.5 meters. Using the fumigation assumptions described in RG 1.145, the licensee calculated an X/Q of 2.04x10-4 s/m3.
Staff review of the data and inputs to the X/Q calculations has found them to be consistent with expected meteorological conditions, NRC guidance, and staff practice. Based on this review, the staff agrees with the licensees conclusion that the NEDO-31400A topical report X/Q values of 2.5x10-3 s/m3 and 3.0x10-4 s/m3 bound the X/Q values calculated for the VY EAB for the 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> period.
3.2 Radiological Consequences Not Addressed by NEDO-31400A The licensee used additional X/Q values not addressed by the NEDO-31400A topical report in the dose assessment discussed below. Calculations were made for an elevated release from the plant stack to the most limiting offsite location, low population zone (LPZ), and control room factoring in terrain, and fumigation when it was the limiting case. Calculations were also made for a postulated release from the turbine building as a ground level release. Both sets of
calculations were made using a computer code based upon the PAVAN methodology (NUREG/CR-2858, PAVAN: An Atmospheric Dispersion Program for Evaluating Design-basis Accidental Releases of Radioactive Materials from Nuclear Power Plants). The calculations had been made for previous licensing activities, using 1978 meteorological data for one set and 1985 data for the other sets of calculations. Staff reviewed the inputs and made comparison estimates using hand calculations, and the PAVAN methodology for the offsite calculations and PAVAN, Murphy-Campe and ARCON96 (NUREG/CR-6331, Rev.1, Atmospheric Relative Concentrations in Building Wakes,) methodologies for the control room calculations using the 1989 through 1993 data set previously discussed. The staff has found the calculations for the postulated elevated release to the offsite locations to be acceptable as part of this review. With regard to the control room intake X/Q values, the methodology used by the licensee to estimate the turbine building X/Q values at the control room intake did not impose a limit on the building wake correction which is inconsistent with guidance and current staff practice. However, X/Q values for postulated releases from both the plant stack and the turbine building to the control room intake utilize the same methodology as that in the licensees January 12, 1981, response to NUREG 0737, Item III.D.3.4. This response was approved by letter dated February 24, 1982. The licensee has subsequently recalculated the X/Q values using more recent meteorological data. Thus, the X/Q values in the January 12, 1981, submittal are not the same as in the submittals supporting this amendment. The X/Q values used in the dose assessment described below are listed in Table 1.
The LTR did not account for an additional release pathway that is assumed to occur at VY.
This is a release of 32 gph (gallons per hour) of reactor coolant for 30 days through the reactor coolant system (RCS) recirculation sampling lines in the reactor building as a result of the elimination of the MSLRM RCS recirculation sampling line automatic closure trip function in Group 1. A summary of the licensees radiological consequences analysis of the release through this additional pathway was provided by letter dated July 16, 2002. The staff reviewed this analysis and determined that the assumptions and inputs are consistent with the SRP method, current VY UFSAR analysis, and NEDO-31400A and, are acceptable. The staff also performed an independent calculation of the additional CRDA doses from a release through this pathway and was able to confirm the licensees results as summarized in Table 2. When the doses resulting from the release through the RCS recirculation sampling lines were added to the doses which result from either LTR scenario (i.e., the release from the condenser or the release through the alternate offgas system), the total CRDA radiological consequences remain below the SRP 15.4.9 dose acceptance criteria of well within the values given in 10 CFR Part 100, or 75 rem for the thyroid and 6 rem for the whole body.
The staff also asked about the impact of the removal of the MSLRM scram/isolation functions on control room habitability. This aspect was not directly addressed by the LTR. In response to the staffs request for additional information, the licensee stated by letter dated July 16, 2002, that they analyzed the impact of the proposed change on control room habitability and provided a summary of the analysis. The licensee analyzed radiological consequences in the control room of the CRDA for the following release pathways:
- a. Leakage from the main condenser, based on the assumption of manual isolation of the main steamlines and/or of the advanced off gas (AOG) system,
- b. Release from the AOG system, based on the assumption that the main steamline remain open, and
- c. Release from the RCS recirculation sampling lines in the reactor building.
The release from the RCS recirculation sampling lines is additive to each of the other two pathways, which are mutually exclusive of each other. The licensees analysis also assumed the control room would not be isolated and would continue to draw in 3,700 cfm of unfiltered air for the duration of the accident.
The staff determined that the licensees assumptions, inputs, and methodologies are consistent with the guidance in the SRP, and are acceptable. The staff performed independent calculations based on the licensees assumptions and confirmed the licensees dose results, which are given in Table 3. The licensees analyses show that the radiological consequences in the control room of a CRDA remain within the dose limits given in GDC-19.
4.0 EVALUATION RESULTS Based on the preceding discussion, the staff finds the removal of the MSLRM scram and main steamline isolation valve closure functions do not increase the offsite radiological consequences of the design-basis CRDA, which remain within the SRP Section 15.4.9 acceptance criterion of well within the dose limits given in 10 CFR Part 100. The staff also finds that the proposed changes do not increase the radiological consequences in the control room, which remain within the dose limits given in GDC-19 for control room habitability. Therefore, the staff finds that the proposed changes are acceptable with respect to the radiological consequences of design-basis accidents. The licensee also revised applicable bases sections and the staff has no objections to these changes.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Vermont State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 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 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 (67 FR 45573).
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.
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 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: M. Hart L. Brown Date: September 18, 2002