ML18036B291
| ML18036B291 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/17/1993 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18036B290 | List: |
| References | |
| NUDOCS 9305240330 | |
| Download: ML18036B291 (20) | |
Text
ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE BROWSES FERRY NUCLEAR PLANT UNITS 1~ 2~ AND 3 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 309'UPPLEMENT 1)
F'DR 9305240330 930517
.PDR ADOCK 05000259
PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 1 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 309i SUPPLEMENT 1)
5.0 MAJOR DESIG FEATURES
- 5. 1 SITE FEATURES Browns Ferry unit 1 is located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA. The site shall consist of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.
5.2 REACTOR A. The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by'ests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
B. The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide and/or hafnium metal as approved by the NRC.
5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.
5.4 CO AINME A. The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR.
The applicable design codes shall be as described in Section 5.2 of the FSAR.
B. The secondary containment shall be as described in Section 5.3 of the FSAR.
BFN 5.0-1 Unit 1
5.0 MAJOR DESIG FEATURES (Continued)
C. Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.
5.5 FUEL STORAGE A. The arrangement of fuel in the new-fuel storage facility shall be such that kaffir for dry conditions, is less than 0.90 and flooded is less than 0.95 (Section 10.2 of FSAR).
B. The keff of the spent fuel storage pool shall be less than or equal to 0.95.
C. Loads greater than 1000 pounds shall not be carried over spent fuel assemblies stored in the spent fuel pool.
The station class I structures and systems have been designed to withstand a design basis earthquake with ground acceleration of 0.2g. The operational basis earthquake used in the plant design assumed a ground acceleration of O.lg (see Section 2.5 of the FSAR).
BFN 5.0-2 Unit 1
PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 309'UPPLEMENT 1)
5.0 MAJOR DESIG FEATURES 5.1 SITE FEATURES Browns Ferry unit 2 is located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA. The site shall'onsist of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.
5.2 REACTOR A. The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
B. The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide and/or hafnium metal as approved by the NRC.
5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.
5.4 CONTAINMENT A. The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR.
The applicable design codes shall be as described in Section 5.2 of the FSAR.
B. The secondary containment shall be as described in Section 5.3 of the FSAR.
BFN 5.0-1 Unit 2
5.0 MAJOR DESIGN FEATURES (Continued)
MEUUEHE (
C. Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.
5.5 FUEL STORAGE A. The arrangement of fuel in the new-fuel storage facility shall be such that keffp for dry conditions, is less than 0.90 and flooded is less than 0.95 (Section 10.2 of FSAR).
B. The keff of the spent fuel storage pool shall be less than or equal to 0.95.
C. Loads greater than 1000 pounds shall not be carried over spent fuel assemblies stored in the spent fuel pool.
5.6 SEISMIC DESIGN The station class I structures and systems have been designed to withstand a design basis earthquake with ground acceleration of 0.2g. The operational basis earthquake used in the plant design assumed a ground acceleration of O.lg (see Section 2.5 of the FSAR).
BFN 5. 0-2 Unit 2
PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 3 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 309'UPPLEMENT 1)
5.0 MAJOR DESIG FEATURES S'il SITE FEATURES Browns Ferry units 1, 2, and 3 are located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA. The site shall consis't of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.
5.2 REACTOR A. The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
B. The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide and/or hafnium metal as approved by the NRC.
5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.
A. The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR.
The applicable design codes shall be as described in Section 5.2 of the FSAR.
B. The secondary containment shall be as described in Section 5.3 of the FSAR.
BFN 5. 0-1 Unit 3
5.0 JOR DESIG FEATURES (Continued)
- MEAA C. Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.
5.5 FUEL STORAGE A. The arrangement of the fuel in the new-fuel storage facility shall be such that keffp for dry conditions, is less than 0.90 and flooded is less than 0.95 (Section 10.2 of FSAR).
B. The keff of the spent fuel storage pool shall be less than or equal to 0.95.
C. Loads greater than 1000 pounds shall not be carried over spent fuel assemblies stored in the spent fuel pool.
The station class I structures and systems have been designed to withstand a design basis earthquake with ground acceleration of 0.2g. The operational basis earthquake used in the plant design assumed a ground acceleration of O.lg (see Section 2.5 of the FSAR).
BFN 5. 0-2 Unit 3
ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1q 2p AND 3 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 309I SUPPLEMENT 1)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION REASON FOR THE CHANGE The current technical specifications (TS), Section 5.2, Reactor, explicitly describe the fuel types utilized in the reactor core.
Progressing to advanced fuel assembly designs offers potential for improved fuel cycle economics and plant operations. Under the current TS, a license amendment is required in order to implement improved fuel assembly designs. The proposed TS change utilizes the guidance provided by Generic Letter 90-02, Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications, eliminates the requirement for cycle-specific TS changes when the fuel design improvements have been analyzed with NRC approved methodology.
DESCRIPTION OF THE PROPOSED CHANGE The proposed change revises TS Section 5.2, Reactor, to read as follows for all three units:
A. The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO ) as fuel material, and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide and/or hafnium metal as approved by the NRC.
11 ENCLOSURE 2 (CONTINUED) 2 of 3 BROGANS FERRY NUCLEAR PLANT (BFN)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION JUSTIFICATION FOR THE PROPOSED CHANGE
References:
- 1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, Revision (latest approved version).
- 2. Letter from A. C. Thadani (NRC) to J. S.
Charnley (GE), "Acceptance for Referencing of Ammendment 18 to General Electric Licensing Topical Report NEDE-24011-P-A, 'General Electric Standard Application for Reactor Fuel,'"
May 12, 1988.
- 3. Letter from H. H. Berkow (NRC) to J. S. Charnley (GE), "Acceptance for Approval of Fuel Designs Described in Licensing Topical Report NEDE-24011-P-A-6, Amendment 10, For Extended Burnup Operation," December 3I 1985.
- 4. Letter from O.J. Zeringue (TVA) to NRC, "TVA BFN Technical Specification (TS) No. 309 Removal of Cycle-Specific Core Operating Limits (Generic Letter 88-16)," August 20, 1992.
- 5. Letter from C.O. Thomas (NRC) to J.F. Klapproth (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-22290, "Safety Evaluation of the General Electric Hybrid I Control Rod Assembly," August 22, 1983.
- 6. Letter from A.C. Thadani (NRC) to J.F. Klapproth (GE), "Acceptance as a Licensing Topical Report NEDE-22290-P, Supplement 3, 'Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly,'" May 5, 1988.
The proposed revision to TS Section 5.2, Reactor, adopts language similar to that proposed by Generic Letter (GL) 90-02, (and Supplement 1) Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications. The GL guidance was issued to provide additional flexibility for the use of reconstructed fuel assembly designs which are based upon NRC-approved design methodologies. The proposed revision is also consistent with the BWR improved standard technical specifications language for fuel assembly and control rod assembly design features. The current fuel assemblies in use at BFN are designed and analyzed in accordance with NRC-approved methodologies. The design bases and criteria are described in Reference l.
ENCLOSURE 2 (CONTINUED) 3of3 BROWNS FERRY NUCLEAR PLANT (BFN)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION NRC has reviewed and approved this design methodology in References 2 and 3. NRC has also reviewed and approved the use of the control rod assemblies currently in use at BFN (References 5 and 6). The proposed TS ensures that future changes to fuel assembly and control rod design are in accordance with NRC-approved methodology. Changes to these designs are made in accordance with BFN's program for compliance with 10 CFR 50.59.
In the Reference 4 letter, BFN submitted a change to Surveillance Requirements (SR) 4.5.J for Unit 1 only and SR 4.5.K for all three units which removes the reference to the specific fuel types for which the Linear Hear Generation Rate (LHGR) and the Minimum Critical Power Ratio (MCPR) shall be checked. In addition to that provided in Reference 4, this change is further justified by the fact that the values for LHGR and MCPR for the fuel types used in each operating cycle will be contained in the Core Operating Limits Report (COLR). As stated in Reference 4, these limits are determined in accordance with NRC-approved methodologies, but the specific numerical values are relocated to the COLR.
In the Reference 4 letter BFN submitted a change to Definition 1.N for all three units which removes the reference to "design power" from the definition. This deletion was proposed for clarity and will ensure that the maximum thermal power authorized by the operating license is not confused with the power level to which the safety analysis applies.
ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)
PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERM1NATZON DESCRIPTION OF THE PROPOSED TECHNICAL SPECIFICATION CHANGE The proposed revision to TS Section 5.2, Reactor, adopts language similar to that proposed by Generic Letter (GL) 90-02, (and Supplement 1)," Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications". The GL guidance was issued to provide additional flexibility for the use of reconstructed fuel assembly designs which are based upon NRC-approved design methodologies. The proposed revision is also consistent with the BWR improved standard technical specifications language for fuel assembly and control rod assembly design features.
BASES FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATZON DETERMINATION NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license involves no significant hazards considerations in accordance with the proposed if operation of the facility amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed TS change is judged to involve no significant hazards considerations based on the following:
- 1. The proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.
Core reload design analyses which address the applicable safety criteria are performed prior to each cycle of operation. These safety criteria require that the radiological consequences of any design basis accident not exceed the guidelines set forth in 10 CFR 100. The analytical methodologies used to satisfy these guidelines are approved by NRC. The proposed amendment requires that these analytical methodologies continue to be approved by NRC. The anticipated operational transients and design basis accidents described in the FSAR are initiated by operator errors or equipment malfunctions other than the fuel or control rod assemblies.
Thus, the adoption of this amendment will not increase the probability or consequences of any accident previously evaluated.
ENCLOSURE 3 (CONTINUED) Page 2 of 2 BROWNS FERRY NUCLEAR PLANT (BFN)
REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION
- 2. The proposed amendment does not. create the possibility of a new or different kind of accident from any accident previously evaluated.
Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed amendment will not result in any changes to any safety-related equipment or its functions.
The proposed amendment will not cause any changes to be made to equipment, systems, or setpoints designed to prevent or mitigate accidents. No changes in the design bases will be made.
- 3. The proposed amendment does not involve a significant reduction in the margin of safety.
Operation of the facility, in accordance with the proposed amendment, would not involve a significant reduction in the margin of safety because an adequate margin of safety is ensured by performing analyses using NRC-approved methodologies to verify compliance with the conditions and acceptance criteria assumed in the FSAR. As these analyses are performed, specific numerical values for core operating limits/restrictions are appropriately set to insure that adequate margin to safety is maintained should an event occur. These limits/restrictions will be contained in the TS and the Core Operating Limits Report. The TS will continue to require compliance with and operation within the bounds of these limits/restrictions and no changes will be made to actions required by the TS in the event of noncompliance.
Development of limits/restrictions and core reload design for future cycles will conform to the NRC-approved methods.
CONCLUS ION TVA has evaluated the proposed amendment described above against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has determined that the proposed amendment will not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility for a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Thus, TVA has concluded that the proposed amendment does not involve a significant hazards consideration.
16:26 I IAY-17-1993 FRQI I VA L ICEI IS II IG TO
~ NRC ROSS P. 819/819 The reactor shall contain 764 fuel assembli s. Each assembly shall consist of a matrix of zirco ium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO>) as fuel mater 'al, and water rods. Limited substitutions of zirconium a loy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that 'have not completed representative testing may be placed in nonlimiting core regions.
k The reactor core shall contain 185 cruciform shaped control rod assemblies. The control mat ria shall be boron carbide and/or hafnium eral as approved by the NRC.
TOTAL P.819
V