ML042020194
| ML042020194 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/19/2004 |
| From: | Marshall M NRC/NRR/DLPM/LPD2 |
| To: | Singer K Tennessee Valley Authority |
| Pascarelli,R J,NRC/NRR/DIPM,415-1245 | |
| References | |
| TAC MB9085, TAC MB9086 | |
| Download: ML042020194 (13) | |
Text
July 19, 2004 Mr. K. 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 RELIEF FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILERS AND PRESSURE VESSEL CODE, SECTION XI, REQUIREMENTS FOR SECOND 10-YEAR INSERVICE INSPECTION PROGRAM (TAC NOS. MB9085 AND MB9086)
Dear Mr. Singer:
By a letter dated May 13, 2003 (ADAMS Accession No. ML040120680) and, as supplemented by letter dated November 6, 2003 (ADAMS Accession No. ML033230347), the Tennessee Valley Authority (TVA) submitted Relief Request (RR) 1-ISI-23 for Unit 1 and RRs 1/2-ISI-24 and 1/2-ISI-25 for Units 1 and 2, requesting relief from American Society of Mechanical Engineers (ASME) Boilers and Pressure Vessel Code inservice inspection for Class 1 and Class 3 welds due to design configuration constraints. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(5)(iii), TVAs request proposes alternative methods, best effort ultrasonic volumetric examinations for Class 1 welds and Visual Test-1 on accessible areas for the Class 3 welds, prescribed in the 1989 Edition of the ASME Boiler and Pressure Vessel Code,Section III.
Based on our review of your submittals, we have concluded that the alternatives proposed in RRs 1-ISI-23,RRs 1/2-ISI-24, and 1/2-ISI-25 provide an acceptable level of quality and safety, and, therefore, are authorized pursuant to 10 CFR 50.55a(g)(6)(i) and 10 CFR 50.55a(a)(3)(i) and (ii). Additionally we have concluded that TVAs method of record is impractical pursuant to 10 CFR 50.55a(g)(5)(iii) under the conditions and the alternative proposed in RRs 1-ISI-23, 1/2-ISI-24, and 1/2-ISI-25 provides an acceptable level of quality and safety, and is therefore approved.
K. W. Singer This relief is authorized for the remainder of the second 10-year inservice inspection interval for Sequoyah Units 1 and 2, which began December 16, 1995, and ends May 31, 2006.
Sincerely,
/RA/
Michael L. Marshall, Jr., Acting Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328
Enclosure:
Safety Evaluation cc w/enclosure: See next page
ML042020194 NRR-028 OFFICE PDII-2/PM PDII-2/LA OGC PDII-2/SC(A)
NAME RPascarelli BClayton SLewis MMarshall DATE 07/15/04 07/15/04 07/06/04 07/19/04
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NOS. 1-ISI-23,1/2-ISI-24, AND 1/2-ISI-25 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By a letter dated May 13, 2003 (ADAMS Accession No. ML040120680) and, as supplemented by letter dated November 6, 2003 (ML033230347), the Tennessee Valley Authority (TVA, the licensee) submitted Relief Requests (RRs) 1-ISI-23 for Unit 1 and RRs 1/2-ISI-24, and 1/2-ISI-25 for Units 1 and 2, requesting relief from American Society of Mechanical Engineers (ASME)
Boilers and Pressure Vessel Code (Code) inservice inspection (ISI) for Class 1 and Class 3 welds due to design configuration constraints. In accordance with Title 10 of the Code of the Federal Regulations (10 CFR) Section 50.55a(g)(iii), TVAs request proposes alternative methods, best effort ultrasonic volumetric examinations for the Class 1 welds, and Visual Test-1 on accessible areas for the Class 3 welds, prescribed in the 1989 Editions of the ASME Boiler and Pressure Vessel Code,Section III.
The ISI of the ASME Code Class 1, Class 2, and Class 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda, as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). In 10 CFR 50.55a(a)(3), it states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2 and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ISI code of record for Sequoyah Nuclear Plant (SQN), Units 1 and 2, second 10-year ISI interval, which began December 16, 1995, is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to commission approval.
2.0 REGULATORY EVALUATION
The Office of Nuclear Reactor Regulation has reviewed the information concerning second 10-year ISI program requests for SQN, Units 1 and 2, in the TVA letters dated May 13, and November 6, 2003. The information provided by the licensee in support of the requests for relief from Code requirements has been evaluated and the basis for disposition is documented below.
2.1 Relief Request No. 1-ISI-23 Component for Which Relief is Requested The component was identified as Weld No. RHRF-107. It is a Class 1 (Equivalent) pressure retaining piping weld in RHR (Residual Heat Removal) System 74.
Code Requirements The licensee requested relief from the requirements of the WCAP-14572, Table 4.1-1, Examination Category R-A, Item Nos. R1.11 and R1.16, which require 100-percent volumetric examination.
License Proposed Alternative In lieu of the code-required 100-percent volumetric examination, a best effort ultrasonic testing (UT) was performed on accessible areas to the maximum, extent practical, given the physical limitations of the subject weld. The examination data report was attached to the submittal.
Basis for Alternative The design configuration of the RHR piping and piping support location precludes UT of the required volume of the subject weld RHRF-107 because of the interference of a permanent piping support. The design configuration limits UT to approximately 72 percent of weld RHRF-107 for the required examination areas as calculated in accordance with TVA Procedure N-GP-28.
The design configuration used in the fabrication of the subject piping in conjunction with the permanent rigid pipe support (1-SIH-801) precludes UT of essentially 100 percent of the required examination volume. In order to examine the weld in accordance with the Code requirements, the RHR system pipe support would require extensive modification. RHRF-107 is limited at two locations due to a permanent rigid support (1-SIH-801) plate which limits the access from the elbow side for a total of 8 inches. One plate is located on the inside radius of the elbow and is approximately 5 inches in length. The other plate is located on the outside radius of the elbow and is approximately 3 inches in length. The design configuration limits the UT to approximately 72 percent for weld RHRF-107. The subject weld was examined using an application of the best available ultrasonic technology. Radiographic examination as an alternate volumetric examination method was determined to be impractical because the same plates that limit UT also obstruct film placement. Examination of similar locations provides reasonable assurance that significant degradation, if present, would have been detected.
Performance of an ultrasonic volumetric examination of essentially 100 percent of the required volume of pressure retaining circumferential weld RHRF-107 (pipe to elbow) in the RHR piping is impractical. The licensee determined that it would be impractical to attempt other volumetric examinations in order to increase examination coverage. The maximum extent practical for UT of the weld and adjacent material and the Code required Visual Examination-2 (VT-2) for leakage will provide reasonable assurance of an acceptable level of quality and safety Significant degradation, if present, would be detected during the UT and VT-2 examinations that are performed on the subject weld. As a result, reasonable assurance of structural integrity for this weld is provided by the examinations that were performed. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), it is requested that relief be granted for the second inspection interval.
Staff Evaluation For compliance with the Code-required examination, the licensee must volumetrically examine essentially 100 percent (more than 90 percent) of the subject weld in the RHR piping.
However, the required examination was limited due to physical restrictions to scanning caused by the RHR system pipe support. Two permanent support plates, one on the outside radius of the elbow and one on the inside radius of the elbow, limit the access for examination. To perform the required examination would require extensive modifications to the existing pipe support, which would cause significant hardship. Another volumetric method, Radiographic Testing (RT), was considered by the licensee but was determined to be impractical because the same restriction that limits UT also obstructs film placement. It is not feasible using either UT or RT to increase examination coverage from either the internal or external surfaces of the pipe because of the burden associated with the modification of the components and the high-radiation levels. The additional examination coverage that would be achieved would not be significant compared with the 72 percent already examined. To require the licensee to modify the subject areas to perform the examinations would be a burden on the licensee.
Therefore, compliance with the Code coverage requirements is impractical.
The volumetric examinations of the subject weld were performed to the extent practical. The examinations performed on the weld resulted in an overall coverage of 72 percent. As part of the Code requirements, a VT-2 examination for leakage was performed on the same component. The staff has determined that the 72 percent examination coverage obtained for the weld combined with a VT-2 examination for leakage is sufficient to detect existing patterns of degradation and, therefore, provides reasonable assurance of structural integrity. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the staff grants relief from the specified coverage requirements of the subject weld for the second inspection interval.
2.2 Relief Requests No. 1-ISI-24 and No. 2-ISI-24 Components for Which Relief is Requested Class 1 (Equivalent) Integral Attachments for Vessels (Pressurizer Support Skirt). Items associated with the RRs are Pressurizer Support Skirt Weld Nos. RCW-23A (Unit 1) and RCW-23 (Unit 2).
Code Requirements The licensee requested relief from the requirements of the 1989 Edition,Section XI of the ASME Code, Table IWB-2500-1, Examination Category B-H, Item B8.20 (Code Case N-509, Category B-K, Item No. B10.10) for the surface examination as applicable based on the configuration of the support skirt to vessel welds: RCW-23A (Unit 1) and RCW-23 (Unit 2).
The Code requires a surface examination of areas A-B and C-D of the pressurizer support skirt to vessel weld. The weld configuration and examination area are illustrated in the ASME Code Section XI, Figure IWB-2500-13. This corresponds to Figure 1 attached to this safety evaluation.
Licensee Proposed Alternative and Its Basis for Use In lieu of the Code-required 100-percent surface examination on both sides of the pressurizer support skirt integral attachment welds, the licensee proposed to perform a 100-percent surface examination on surface A-B as illustrated in Figure 1 and a best effort ultrasonic volumetric examination for surface C-D, as illustrated in Figure 1, on both welds. The examination reports were attached to the submittal.
The design configuration limits surface examination to approximately 50 percent of weld RCW-23 for Unit 1 and 50 percent of weld RCW-23A for Unit 2 of the required examination areas as calculated in accordance with TVA Procedure N-GP-28.
The design configuration of the pressurizer support skirt precludes surface examination of essentially 100 percent of the required examination area. In order to examine the welds in accordance with the Code requirements, the pressurizer would require extensive modification.
The weld configuration for SQN Units 1 and 2 pressurizer support skirts is illustrated in Figure 1, and receives a surface examination of areas A-B and C-D (the outside and inside surfaces, respectively).
Physical access by the examiner is limited because of high-radiation levels and obstructions due to the immersion heaters and the related wiring. The magnetic particle examination (yoke) cannot be used due to the space restrictions. The dye penetrant examination would require a very thorough cleaning of the weld and adjacent base material. The preparation of the weld would potentially require using techniques such as manual wire brushing since power tools may not fit into the limited space area.
Radiological dose rate in these areas (bottom head) may be approximately 40 to 45 millirem/hour. It is conservatively estimated that approximately 32-man hours would be required (two people and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to remove/assemble scaffolding, two people and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to remove/assemble insulation and two people and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform the examination, if it was accessible). A total of 1.28 roentgen equivalent-man could be received by all involved personnel.
The ultrasonic volumetric examination of the inaccessible area was performed using a 45-and 60-degree shear wave examination resulting in 100-percent examination coverage of the peressurizer support skirt weld examination area C-D, as illustrated in Figure 1 (inaccessible surface area). Performance of a surface examination of essentially 100 percent of the required surface of the pressurizer support skirt weld is impractical. A surface examination on surface A-B as illustrated in Figure 1 and an ultrasonic volumetric examination for the inaccessible surface was performed on the subject weld. These examinations provide reasonable assurance of an acceptable level of quality and safety because the weld surface and the weld volume examined provide sufficient information to judge the overall integrity of the weld.
The proposed alternative volumetric examination of the inaccessible area has been evaluated and judged technically acceptable by the NRC staff for use at TVAs Sequoyah Nuclear Plant, Units 2 and 3, for the reactor pressure vessel support skirt weld. Therefore, an acceptable level of quality and safety will be achieved and public health and safety will not be endangered by utilizing the proposed alternative volumetric examination of the inaccessible areas in lieu of the prescribed surface examination requirements of the ASME Code. Compliance with the ASME Code will result in unusual difficulty and radiation exposure to personnel without any compensating increase in the level of quality or safety. The proposed alternative will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), it is requested that relief be granted for the second inspection interval.
Staff Evaluation The ASME Code specifies examination requirements for the skirt weld configurations depicted in Figure 1. The Code requires surface examinations of both outside and inside welded surfaces for the Figure 1 configuration. The inside examination requires personnel entry to the area inside the skirt.
The licensee has requested relief from the Code requirement. The relief is necessitated by the narrow access through the skirt and the obstruction in the confined area inside the skirt under the bottom head. The design configuration of the pressurizer support skirt precludes surface examination of essentially 100 percent of the required examination area. In order to examine the welds in accordance with the code requirements, the pressurizer would require extensive modification. Therefore, compliance with the Code-specified requirements would result in hardship.
The skirt weld configuration at SQN is similar to Figure 1. Instead of performing surface examinations from both sides of the welds in accordance with the Code requirements applicable to the Figure 1 configuration, the licensee proposes to perform (a) a surface examination on the outside surface (surface A-B as illustrated in Figure 1) of the welds and, (b) a best-effort UT of the volume defined along the inside surface (surface C-D as illustrated in Figure 1) of the welds, from the outside surface. This does not necessitate personnel entry to the area inside the skirt.
This licensees proposed alternative for its skirt welds is similar to the Code requirements applicable to the Figure 1 configuration. The Figure 1 configuration can, to a significant extent, be volumetrically examined from the outside surface. The weld volume capable of being examined is through-wall and at or near the inside weld surface. Any crack propagating from the inside weld surface would be detectable. The staff has determined that a best-effort examination, in combination with the outside surface examination of the skirt weld will provide reasonable assurance of structural integrity and an acceptable level of quality and safety.
The staff concludes that compliance with the Code requirements would result in hardship without a compensating increase in the level of quality and safety, and that the proposed alternative provides reasonable assurance of the structural integrity of the component.
Therefore, the licensees request is authorized pursuant to 10 CFR 50.55a(a)(3)(i) and (ii) for the second 10-year ISI interval.
2.3 Relief Requests No. 1-ISI-25 and No. 2-ISI-25 Components For Which Relief Is Requested ASME Code Case 3 (Equivalent) essential raw cooling water (ERCW) strainer support integral attachment weld.
Code Requirements ASME Section XI, Table IWD-2500-I, Examination Category D-A, Item No. D1.20, Requires VT-3 Examination.
Licensee Proposed Alternative and Its Basis for Use The licensee requested relief from the requirements of the VT-3 examination coverage of essentially 100 percent of the ERCW strainer support integral attachment weld. In lieu of the Code required 100-percent VT-3 examination, a VT-1 examination of the ERCW strainer integral attachment weld was performed on accessible areas to the maximum extent practical given the physical limitations. Examination reports were attached to the submittal.
The design configuration of the subject weld (partial penetration), Item No. 5 (base plate) to Item No. 1 (strainer) precludes visual examination of essentially 100 percent of the required examination area. In order to visually examine the weld in accordance with the Code requirement, the ERCW strainer would require extensive disassembly in order to access the inside surface of the strainer. In order to disassemble the subject strainer, a temporary monorail is required to be installed due to limited overhead clearance.
The physical arrangement of the ERCW Strainer is composed of 16 integral lug (Item No. 30) attachments that support that strainer body (Item No. 1) to a base plate connection (Item No. 5). As part of the integral attachment, the base plate assembly (Item No. 5) is attached to the strainer body (Item No. 1) by a partial penetration weld from the inside surface for 360 degrees and a fillet weld 360 degrees around the outer circumference for 360 degrees.
Each of the 16 attachment lugs and outer base plate to strainer body welds are accessible from the outside surface. The inside surface partial penetration weld does not allow for visual examination. The amount of examination coverage when considered in total weld length examined was determined to be approximately 75 percent.
Radiographic examination from the outside surface as an alternative volumetric examination method was determined to be impractical due to the support attachments affecting radiographic quality. Performing a radiographic examination from the inside surface of the ERCW strainer would require placing a radiographic source near the center of the component. The ERCW strainer would require extensive disassembly and being taken out of service for an extended period of time. Radiographic quality would be compromised due to the support attachments and the subject contrast effects from conditions associated with a partial penetration weld (i.e.,
butt joint match lines would mask indication due to the partial penetration condition). Thus, additional radiography from the inner surface to gain any additional coverage, is also impractical.
Performance of an ultrasonic volumetric examination to supplement the required coverage was also deemed to be impractical. The design configuration of the partial penetration weld is not amenable for the detection of circumferential flaws due to the partial penetration condition (i.e.,
discrimination of weld flaws is masked due to the partial penetration weld geometry). In addition, the fillet weld impedes the search unit position from the outside surface such that the ultrasonic beam cannot be oriented for detection of radial oriented flaws. The support attachments are 1-inch thick welded lugs equally stationed at 16 positions around the strainer.
The lug attachments, along with five equally spaced bolt holes, would in addition interfere with the sound beam from being detected to the inner-weld region.
Performing a surface examination from the inside was determined to be impractical due to the same reasons as the visual examination. The more detailed VT-1 examination was performed in lieu of the VT-3 examination.
The high percentage (75 percent) of the VT-1 examination of the subject weld areas and adjacent base metal was obtained. This examination coverage provides reasonable assurance of an acceptable level of quality and safety. The weld areas examined provides sufficient information to judge the overall integrity of the weld. Significant degradation, if present, would be detected during the VT-1 examination of the subject weld.
As a result, assurance of structural integrity for these welds is provided by the examinations that were performed.
Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), it is requested that relief be granted for the second 10-year ISI interval.
Staff Evaluation The Code requires 100-percent VT-3 examination of the ERCW strainer support integral attachment welds. The licensee was unable to obtain the Code required coverage of 100-percent VT-3 examination of the subject welds. The examinations were limited due to the support configuration or permanent obstructions on the outside surface. To achieve the examination requirements would require major disassembly of the components and modifications to the plant which would be a significant burden on the licensee. Therefore, compliance with the Code requirements would result in significant hardship.
The licensee proposes to perform VT-1, a more detailed and closer visual examination than VT-3, on the subject welds to the extent possible. During the performance of the examination, the examiner is to be within 2 feet of the examination surface for the VT-1 instead of 4 feet as required for the VT-3 examination. This closer distance would enable the examiner to better detect discontinuities and imperfections on the surface. The inspection obtained approximately 75 percent of the examination coverage using the VT-1. The achieved coverage using a more detailed, closer visual examination is sufficient to detect any patterns of degradation even though 100-percent coverage could not be achieved. The staff finds that the licensees proposed alternative provides reasonable assurance of structural integrity of the subject welds.
The staff concludes that to examine the subject welds as required by the Code, the welds would have to be redesigned and modified, which would result in considerable hardship on the licensee without a compensating increase in the level of quality and safety. The licensees alternative obtained significant coverage of the subject welds with a more detailed and closer examination method, and provided reasonable assurance of structural integrity of the component. Therefore, the licensees request is authorized pursuant to 10 CFR 50.55a(a)(3)(i) and (ii) for the second 10-year ISI interval.
3.0 CONCLUSION
The staff has reviewed the information in the licensees letter dated May 13, 2003, and November 6, 2003, RRs 1-ISI-23, 1-ISI-24, and 1-ISI-25 for SQN, Unit 1, and 2-ISI-24, and 2-ISI-25 for SQN, Unit 2.
For RR 1-ISI-23, the staff concludes that the licensee has maximized examination coverage to the extent practical for the welds and that licensees request for relief is granted pursuant to 10 CFR 50.55a(a)(6)(i) for the second 10-year ISI interval.
For RRs 1-ISI-24 and 2-ISI-24, the staff concludes that compliance with the Code requirements would result in hardship without compensating increase in the level of quality and safety, and the licensees proposed alternative provides reasonable assurance of the structural integrity of the component. Therefore, the licensees request is authorized pursuant to 10 CFR 50.55a(a)(3)(i) and (ii) for the second 10-year ISI interval.
For RRs 1-ISI-25 and 2-ISI-25, the NRC staff concludes that to examine the subject welds as required by the Code would result in hardship without compensating increase in the level of quality and safety, and that the licensees proposed alternative provides reasonable assurance of the structural integrity of the component. Therefore, the licensees request is authorized pursuant to 10 CFR 50.55a(a)(3)(i) and (ii) for the second 10-year ISI interval.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Z. Bart Fu Date: July 19, 2004
Attachment:
Figure 1
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. James E. Maddox, Vice President Engineering & Technical Services 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 37379 General Counsel Tennessee Valley Authority ET 11A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. T. J. Niessen, Acting General Manager Nuclear Assurance Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Mark J. Burzynski, Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. Pedro Salas, Manager Licensing and Industry Affairs Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37379 Mr. David A. Kulisek, Plant Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37379 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