ML20098C513
| ML20098C513 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/02/1995 |
| From: | Stetz J CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20098C501 | List: |
| References | |
| NUDOCS 9510100182 | |
| Download: ML20098C513 (14) | |
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Dockst Number 50-346 Licensa Number NPF-3 Serial Number 2323 Enclosure Page.1 APPLICATION FOR AMENDMENT TO r
FACILITY OPERATING LICENSE NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPR-3. Also included is the Safety Assessment and Significant Hazards Consideration.
The proposed changes (submitted under cover letter Serial Number 2323) concern:
Appendix A, Technical Specification 5.0, Design Features By: 2 J. P. SM tz, Vice Wed@ent - Nuclear Sworn and subscribed before me this 2nd day of October,1995.
0ssts/ 1/A Notary PublicI 7t' ate of Ohio EVELYN L DRESS Notary Public. Stane of ONe My Commiseson Empires 7/28/90 i
9510100192 951002 PDR ADOCK 05000346 P, . , , _ _. ..PDR
-Dockst Number 50-346 Licensa Number NPF-3~
LSarial. Number 2323 Enclosure.
Page.2 The:following information is provided to' support issuance of the requested-
-changes'to Davis-Besse Nuclear Power Station, Unit Number 1 Operating License'NPF-3, Appendix A, Technical Specification (TS) Section 5.0 Design Features.
A. Time Required.to Implement: This change is to be implemented within 90-days after NRC, issuance of the License Amendment.
B. Reason for Change.(License Amendment Request 95 0009, Revision 0):
The proposed. changes would add a site. location description,' remove site area maps, remove containment'and reactor. coolant system design parame-ters, remove the description of the meteorological tower location, re-move component cyclic-or transient limits, and revise the fuel assembly description to include the use of Z1RLO clad fuel rods. The revision to'TS Section 5.0 is being proposed as line item TS improvements consistent with the improved " Standard Technical Specifications for Babcock and Wilcox Plants" (NUREG-1430,. Revision 1) and a lead plant change approved by the NRC for the'Calvert Cliffs Nuclear Power Plant, Units 1 & 2.
.C. Safety Assessment and Significant Hazards Consideration: See
' Attachment.
l Docket Numbar 50-346 License Number NPF-3 Serial Number 2323 Attachment '
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 95-0009 (26 pages follow) i 4
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LAR 95-0009 Page 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR i LICENSE AMENDMENT REQUEST NUMBER 95-0009 l l
TITLE. ,
Proposed Revision of Section 5.0, Design Features, of the Operating License Appendix A, Technical Specifications.
DESCRIPTION:
The purpose of the proposed changes is to modify the following Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications (TS) of Section 5.0, Design Features: TS 5.1 (including Figures 5.1-1 through 5.1-4), Site; TS 5.2, Containment; TS 5.3, Reactor Core; TS 5.4, Reactor Coolant System; TS 5.5, Meteorological Tower Location; and TS 5.7, Component Cyclic or Transient Limit (including Table 5.7-1). These changes propose to add a site location description, remove site area maps, remove containment and reactor coolant system design parameters, remove the description of the meteorological tower location, remove component cyclic or transient limits, and revise the fuel assembly description to allow the use of ZIRLO clad fuel rods. Aside from minor format changes, changes are not proposed for TS 5.6, Fuel Storage, because the current content reflects the specific DBNPS design and is generally consistent with the type of information in NUREG-1430, Revision 1.
These changes, described in detail below, are being proposed as line item TS improvements similar in content to the improved " Standard Technical Specifica-tions for Babcock and Wilcox Plants" (NUREG-1430, Revision 1) and a lead plant change approved by the NRC for the Calvert Cliffs Nuclear Power Plant, Unit 1, Cp-rating License No. DPR-53 (Amendment 204, dated March 14, 1995, TAC No.
M88429) and Unit 2, Operating Licere No. DPR-69 (Amendment 182, dated March 14, 1995, TAC No. M88430). These changes revise Section 5.0, Design Features, and reformat the content to be generally consistent with NUREG-1430, Revision 1.
Technical Specification 5.1, Site, presently describes thu Exclusion Area (Figure 5.1-1), the Low Population Zone (Figure 5.1-2), and the unrestricted area and site boundaries for radioactive liquid effluents (Figure 5.1-3) and for radioactive gaseous effluents (Figure 5.1-4). Consistent wit'.' NUREG-1430, and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant, it is proposed that these four figures be deleted, and that a site location description be added to TS 5.1 as follows:
The Davis-Besse Nuclear Power Station, Unit Number 1, site is located on Lake Erie in Ottawa County, Ohio, approximately six miles northeast from Oak Harbor, Ohio and 21 miles east from Toledo, Ohio. The exclusion area boundary has a minimum radius of 2400 feet from the center of the plant.
Technical Specification 5.2, Containment, presently describes the containment vessel and shield building. It is proposed that TS 5.2 be deleted, consistent
- ith NUREG-1430 and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant.
LAR 95'-0009 Page 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 95-0009 TITLE:
Proposed Revision of Section 5.0, Design Featurer, cf the Operating License Appendix A, Technical Specifications.
DESCRIPTION:
The purpose of the proposed changes is to modify the following Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications (TS) of Section 5.0, Design Features: TS 5.1 (including Figures 5.1-1 through 5.1-4), Site; TS 5.2, Containment; TS 5.3, Reactor Core; TS 5.4, Reactor Coolant System; TS 5.5, Meteorological Tower Location; and TS 5.7, Component Cyclic or Transient Limit (including Table 5.7-1). These changes propose to add a site location description, remove site area maps, remove containment and reactor coolant system design parameters, remove the description of the meteorological tower location, remove component cyclic or transient limits, and revise the fuel assembly description to allow the use of ZIRLO clad fuel rods. Aside from minor format changes, changes are not proposed for TS 5.6, Fuel Storage, because the current content reflects the specific DBNPS design and is generally consistent with the type of information in NUREG-1430, Revision 1.
These changes, described in detail below, are being proposed as line item TS improvements similar in content to the improved " Standard Technical Specifica-tions for Babcock and Wilcox Plants" (NUREG-1430, Revision 1) and a lead plant change approved by the NRC for the Calvert Cliffs Nuclear Power Plant, Unit 1, Operating License No. DPR-53 (Amendment 204, dated March 14, 1995, TAC No.
M88429) and Unit 2, Operating License No. DPR-69 (Amendment 182, dated March 14, 1995, TAC No. M88430). These changes revise Section 5.0, Design Features, and reformat the content to be genertlly consistent with NUREG-1430, Revision 1.
Technical Specification 5.1, Site, presently describes the Exclusion Area (Figure 5.1-1), the Low Population Zone (Figure 5.1-2), and the unrestricted area and site boundaries for radioactive liquid effluents (Figure 5.1-3) and for radioactive gaseous effluents (Figure 5.1-4). Consistent with NUREG-1430, and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant, it is proposed that these four figures be deleted, and that a site location description be added to TS 5.1 as follows:
The Davis-Besse Nuclear Power Station, Unit Number 1, site is located on Lake Erie in Ottawa County, Ohio, approximately six miles northeast from !
Oak Harbor, Ohio and 21 miles east from Toledo, Ohio. The exclusion area boundary has a minimum radius of 2400 feet from the center of the plant. ;
Technical Specification 5.2, Containment, presently describes the containment vessel and shield building. It ia proposed that TS 5.2 be deleted, consistent with NUREG-1430 and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant.
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t Technical Specification 5.3, Reactor Core, presently describes fuel a'ssemblies and control rods. The proposed change would revise the fuel assembly descrip-tion in TS 5.3.1 to allow the use of ZIRLO clad fuel rods in addition to zirca-loy clad fuel rods. The use of ZIRLO as fuel rod cladding has been previously accepted by the NRC in 10 CFR 50.46 and would facilitate potential future use by the DBNPS. The proposed change would also remove from TS 5.3.2 the description of the control rods and the discussions of the length of the control rod absor-ber material, the cladding material for the control rods and axial power shaping rod (APSR) assemblies, and the length of the APSR absorber material. These changes are consistent with NUREG-1430 and the referenced similar change approv-ed by the NRC for the Calvert Cliffs Nuclear Power Plant.
Technical Specification 5.4, Reactor Coolant System, presently describes the reactor coolant system, including operating pressure and temperature, and total water and steam volume. It is proposed that TS 5.4 be deleted, consistent with NUREG-1430 and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant.
Technical Specification 5.5, Meteorological Tower Location, presently co.'tains a description for the location of the meteorological tower (Figure 5.1-1). It is proposed that TS 5.5 be deleted, consistent with NUREG-1430 and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant.
Technical Specification 5.7, Component Cyclic or Transient Limit, presently contains component cyclic or transient limits (Table 5.7-1). It is proposed that TS 5.7 and Table 5.7-1 be deleted, consistent with NUREG-1430 and the referenced similar change approved by the NRC for the Calvert Cliffs Nuclear Power Plant.
Various editorial changes are also proposed to make the format of TS 5.0 gener-ally consistent with the format of NUREG-1430.
SYSTEM, COMPONENTS AND ACTIVITIES AFFECTED:
Description of Design Features in Technical Specification Section 5.0 FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS AND ACTIVITIES:
Site Location The DBNPS site location is described in Updated Safety Analysis Report (USAR) ;
Sections 1.2.1.1, Site, and 2.1.1, Site Location. The Davis-Besse Nuclear Power Station (DBNPS) occupies a site on the southwestern shore of Lake Erie and is located in Ottawa County in Northwestern Ohio. As shown on USAR Figure 2.1-1, DBNPS Site Location & Geographic Features - 20 Mile Radius, the DBNPS site is located approximately raix (6) miles northeast from Oak Harbor, Ohio, and twenty, one (21) ndles east from Toledo, Ohio. The site boundary is depicted on USAR Figure 2.1-3, DBNPS Site Arrangement. As stated in USAR Section 2.1.2, Site '
Description, the site boundary shown on this figure is also the limit of,the exclusion area. As stated in USAR Section 2.1.2.2, Boundaries for Establishing-Effluent Release Limits, the site boundary is the boundary line which determines dI
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LAR 95-0009 Page 3 the limits of gaseous releases for purposes of the Offsite Dose Calculation Manual limits to fulfill the requirements of 10 CFR 20 pertaining to radioactive releases to unrestricted areas. As further stated in USAR Section 2.1.2.2, the distance from the vent stack for gaseous releases, located on the side of the shield building, to the nearest site boundary is two-thousand and four hundred (2,400) ft. The shield building is at approximately the center of the plant structures. USAR Section 2.1.2.2 also describes the discharge points for effluents from the station area for liquid discharges. USAR Section 2.1.3.3, Low Population Zone, describes the low population zone as the area outside of the site boundary within a radius of two (2) miles from the center of the containment structure. This area is depicted on USAR Figure 2.1-7, DBNPS Low Population Zone.
Containment The containment system is described in USAR Sections 1.2.10, Containment Sys-tems, 3.8.2.1, Containment Vessel, 3.8.2.2, Shield Building and 6.2.1, Containment Vessel Functional Design. The containment is composed of a steel containment vessel and a reinforced concrete shield building. The containment vessel is a low leakage cylindrical steel pressure vessel with a hemispherical dome and ellipsoidal bottom. It is designed to withstand a postulated loss-of-coolant accident (LOCA) and to confine a postulated release of radioactive material. As described in USAR Section 1.2.10.1, Principal Design ,
Criteria, the containment vessel design maximum internal pressure is 40 psig at a coincident temperature of 2640F. The shield building is a reinforced concrete structure having a cylindrical shape with a shallow dome roof. It completely surrounds the containment vessel and is designed to provide biological shielding during normal operation and from hypothetical accident conditions. An annular space is provided between the shield building and the containment vessel. The shield building provides a means for collection and filtration of fission product leakage from the containment vessel following a hypothetical accident.
In addition, the building provides environmental prote'ction for the containment vessel from adverse atmospheric conditions and external missiles.
Reactor Core The reactor core is described in USAR Section 4.2.1, Puel and USAR Appendix 4.B, Reload Report. The function of the reactor core is to generate power for an analyzed period. The complete core consists of one-hundred and seventy-seven (177) fuel aseemblies arranged in a square lattice to apprcximate a cylinder. A fuel assembly is no-> ally composed of two hundred and eight (208) fuel rods, sixteen (16) control rod guide tubes, one (1) instrument tube assembly, eight (8) spacer grids, and two (2) end fittings. The guide tubes, spacer grids, and end fittings form a structural cage to arrange the rods and tubes in a 15x15 array.
i The fuel rods consist c' zircaloy cladding containing fuel pellets in a columnar +
arrangement. Zircaloy is a zirconium-based alloy. Each fuel pellet consists of
- a high density, low enrichmenc, uranium dioxide material. The safety function of the fuel rod cladding is to provide a primary barrier to prevent the release i of fission products.
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i LAR 95-0009 Page 4 The control rod assemblies, including extended life control rod assemblies, are described in USAR Section 4.2.3.1.1.1, Control Rod Assembly (CRA) and Extended Life Control hod Assembly (ELCRA). There are fifty-three (53) CRAs in the reactor core. The safety function of the control rod assemblies is to shut down the reactor. In addition, the control rods control fast reactivity changes, and, in conjunction with the axial power shaping rods (APSRs), burnable poison rods, and soluble boron, control lifetime reactivity and power distribution changes.
The APSRs are described in USAR Section 4.2.3.1.1.2, Axial Power Shaping Rod Assembly (APSRA) Black and Gray Type. There are 8 APSRs in the reactor core.
The function of the APSRs is to maintain an acceptable axial distribution of power.
Reactor Coolant System The Reactor Coolant System is described in detail in USAR Section 5.0, Reactor Coolant System. The Reactor Coolant (RC) System (RCS) consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed reactor coolant pumps, an electrically heated pressurizer, and interconnecting piping.
The system, located entirely within the containment Vessel (CV), is arranged in two heat transport loops, each with two RC pumps and one steam generator.
Reactor coolant is transported through piping connecting the reactor vessel to the steam generators and flows downward through the steam generator tubes
, transferring heat to the steam and water on the shell side of the steam genera-tor. In each loop, the RC is returned to the reactor through two lines, each containing a RC pump. In addition to serving as a heat transport medium, the coolant also serves as a neutron moderator and reflector and as a solvent for the soluble poison (boron in the form of boric acid) utilized in chemical shim reactivity control. Design pressure and temperature for RC pressure boundcry (RCPB) components are provided in USAR Table 5.1-la, Design Conditions for RCPB, and USAR Table 5.1-1b, Components Within RCPB. Volumes of the various RCS components are shown on USAR Figure 5.1-1, DBNPS Reactor Coolant System Flow Diagram at Full Power Steady State Conditions.
As described in USAR Section 1.2.2.3, Safety Considerations, the RCS is designed to maintain its integrity under all operating conditions, thereby minimizing the release to the containment vessel of fission products that escape the primary ,
barrier (the fuel rod cladding). l Meteorological Tower Location The location of the meteorological towers is described in USAR Section 2.3.3, On-Site Meteorological Measurements Program, and depicted on USAR Figure 1.2-12, Site Plan and USAR Figure 2.1-3, DBNPS Site Arrangement. As described in USAR Section 2.3.3, the two towers, a 100 meter tower and a nearly 10 meter tower, are located within a fenced compound in the southwest corner of the site. The on-site Meteorological Monitoring System (MMS) complies with the requirements of Regulatory Guide 1.23, " Meteorological Programs in Support of Nuclear Power Plants," proposed Revision 1, dated September 1980. As described in Section 7.5.5 of the DBNPS Emergency Plan, meteorological instrumentation readouts are located in the DBNPS Control Room. The Emergency Control Center and the
LAR 95-0009 Page 5 ]
i Technical Support Center can also obtain this instrumentation data through the )
Data Acquisition and Display System. The data can be used to estimate potential j radiation doses to the public resulting from actual routine or accidental 1 releases of radioactive materials to the atmosphere, or to evaluate the potential dose to the public as a result of hypothetical reactor accidents.
Component Cyclic or Transient Limits The component cyclic or transient limits for RC system components are described in USAR Sections 5.1.3, Reaction Loads, and 5.1.4, Service. Each component of the RC system is designed to withstand the e ffects of cyclic loads due to '
temperature and pressure changes in the RC System. These cyclic loads are introduced by normal unit load transients, reactor trip, and startup and shut- -
down operation. The design service life is 40 years. Design cycles are shown in-USAR Table 5.1-8, Transient Cycles - 40 Year Design Life. (This USAR table
.will be reviewed and modified as deemed necessary.)
EFFECTS ON SAFETY:
Section 182.a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to state TSs to be included as part of the license. Section 182.a of the Act states, in part:
In connection with applications for licenses to operate production or =
utilization facilities, the applicant shall state such technical specifica-tions, including information of the amount, kind, and source of special nuclear materials required, the place of the use, the specific character-istics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that utilization or production of special nuclear material will be in accord with .
the common defense and security and will provide adequate protection to the '
health and safety of the public. Such technical specifications shall be a
- part of any license issued.
The Nuclear Regulatory Commission's regulatory requirements related to the content of TSs are set forth in 10 CFR Section 50.36. This regulation requires
- that the TSs include items in five specific categories, including: (1) safety
- limits, limiting safety system settings and limiting control settings; (2) l limiting conditions for operation; (3) surveillance requirements; (4) design
- features; and (5) administrative controls. However, the regulation does not a specify the particular requirements to be included in a plant's TS.
10 CFR 50.36(c)(4) states, in part, " Design features to be included are those i
features of the facility such as materials of construction and geometric
, arrangements, which, if altered or modified, would have a significant effect on !
l safety and are not covered in categories described in paragraphs (c)(1), (2), I and (3) of this section (Safety Limits, Limiting conditions for Operations, and Surveillance Requirements)."
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1 10 CFR 50.59, ' Changes, tests and experiments," provides the criteria for determining if a proposed change to the features of a facility, as described in the' safety analysis report, is an unreviewed safety question. The criteria in 10 CFR 50.59 for making this determination are: (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may increase; (2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any TS is reduced. ;
With the above citations as references, the following criteria is used to determine what information should be placed in the design features section of the TSs:
- 1. The amount, kind, and source of special nuclear material required;
- 2. The place of the use of the special nuclear materials; and
- 3. Those features of the facility, such as materials of construction and geo-metric arrangements, which if altered or modified would have a significant effect on safety and are not covered in the safety limits, limiting condi-tions for operation (LCO), or surveillance requirements of the TSs.
Based on the above criteria, the proposed changes to the DBNPS TS Section 5.0, Design Features, will remove some subsections and modify others to generally conform to the examples provided in the Design Features section of NUREG-1430 and the changes approved for the Calvert Cliffs Nuclear Power Plant, as follows:
Site Location i
Section 182.a of the Act requires that the pli.e of the use of the special nuclear material be specified. Presently, TS 5.1.1 meets this requirement by reference to TS Figure 5.1-1, which is a map of the site. In the proposed change to TS 5.1, the place of use would be specified by a description rather than a map. This is consistent with NUREG-1430, which requires a text descrip-tion of the site location. In addition, the proposed change would delete TS i Figures 5.1-1, 5.1-2, 5.1-3, and 5.1-4, however, as described above, this i information is presently included in the Updated Safety Analysis Report (USAk). j In accordance with 10 CFR Part 100, the site description includes a minimum j distance to the Exclusion Area Boundary to ensure that the area, for which the licensee has the authority to determine all activities including the exclusion or removal of personnel and property from the area, is clearly associated with the " place of use" referred to in Section 182.a of the Act. The inclusion of the maps in the USAR ensures that any change to either the boundaries or the zone will have to be evaluated using the 10 CFR 50.59 process.
Containment The change proposes that TS 5.2 be removed consistent with NUREG -1430. Al-
, though certain modifications or alterations to the containment could have a significant impact on plant safety, adequate control of the containment systems i
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LAR 95-0009 Page 7 limiting conditions for operations are included in TS Section 3/4.6. Therefore, l this information need not be specified in the Design Features section. Further, the containment vessel and shield building are described in USAR Chapters 1, 3 ;
and 6 and changes to these would require evaluation using the 10 CFF 50.59 )
process.
Reactor Core Since the Act requires that the amount, type, and source of special nuclear material be specified in the TSs, TS 5.3 will be retained, however, several modifications are proposed.
As described above, a change is proposed to revise the fuel assembly description to allow the use of ZIRLO clad fuel rods, consistent with NUREG-1430 and 10 CFR 50.46. ZIRLO is a zirconium-based alloy. As described in USAR Section 4.2.1, and USAR Appendix 4.B, Reload Report, zircaloy is presently used for fuel rod cladding, hence a change to the USAR (and a 10 CFR 50.59 evaluation) would be required in order to use ZIRLO cladding. In addition, the present TS 5.3.1 requirement that fuel designs be ' analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases" will ensure that any future use of ZIRLO cladding will not have an adverse effect on safety.
As also described above, consistent in general with NUREG-1430, a change is proposed to remove much of the detail in TS 5.3.2 regarding the description of control rods and APSRs, including the length of absorber material and cladding material. The information being removed is presently contained in USAR Sections 4.2.3.1.1.1 (Table 4.2-6) and 4.2.3.1.1.2 (Table 4.2-7).
Reactor Coolant System The change proposes that TS 5.4 be removed, consistent with NUREG-1430. Other sections of the TSs (such as TS Section 3/4.4, Reactor Coolant System) adequate-ly control the Reactor Coolant System parameters such as temperature, pressure, and boundary degradation, which could have a significant impact on safety. l Further, the information being removed from TS 5.4 is presently contained in !
USAR Chapter 5.
Meteorological Tower Location The change proposes that TS 5.5 be eliminated,. consistent with NUREG-1430. As j described above, the meteorological tower location is presently contained in USAR Section 2.3.3.
1 Component Cyclic or Transient Limits i The change proposes that TS 5.7 be removed, consistent with NUREG-1430, Revision I The information does not meet the criteria for inclusion in the TSs. Fur-1.
ther, the information contained in TS 5.7, which is to be removed by this proposed change, is either already in the USAR or will be added to the USAR, and will be controlled by the requirements of 10 CFR 50.59, 1
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.Under the proposed changes, TS Section 5.0 would continue to satisfy the appli- I cable requirements of Section 182.a of the Act and 10 CFR 50.36 (c)(4). Furth- i er, the changes are consistent with NUREG-1430. The information being removed !
is presently included in the USAR or is being added to the USAR, hence suffici-ent controls exist under 10 CFR 50.59 to ensure that future changes to these items are acceptable. Therefore, the proposed changes will have no adverse effect on safety, The various editorial changes which are proposed to make the format of TS 5.0 more consistent with the format of NUREG-1430, are administrative, and will have no adverse effect on safety.
SIGNIFICANT HAZARDS CCNSIDERATION:
.The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no signifi-cant hazards consideration if operation of the facility in accordance with the proposed changes would: (1) Not involve a significant increase in the proba-bility or consequences of an accident previously evaluated; (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety.
Toledo Edison has reviewed the proposed changes and determined that a signifi-cant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station Unit Number 1, in accordance with these changes would:
la. Not involve a significant increase in the probability of an accident pre-viously evaluated because no accident initiators, conditions or assumptions are affected by the proposed changes to Section 5.0, Design Features, of the Technical Specifications. Theae changes are proposed to add a site location description, remove site area maps, remove containment and reactor coolant system design parameters, remove the description of the meteorolog-ical tower location, remove component cyclic or transient limits, and revise the fuel assembly description to include the use of ZIRLO clad fuel rods.
Under the proposed changes, Technical Specifications (TS) Section 5.0 would continue to satisfy the applicable requirements of Section 132.a of the Atomic Energy Act. of 1954, and 10 CFR 50.36(c) (4) . Further, the proposed changes are consistent with NUREG-1430, " Standard Technical Specifications for Babcock and Wilcox Plants," Revision 1. The information proposed for removal from existing TS 5.0 is presently included in the Updated Safety Analysis Report (USAR) or is being proposed to be added to the USAR, hence sufficient controls exist under 10 CFR 50.59 to ensure that future changes to these items are acceptable.
LAR 95-0009 Page 9 lb. Not involve a significant increase in the consequences of an accident previously evaluated because no accident conditions or assumptions are affected by the proposed changes. As described above, these changes are consistent with the " Standard Technical Specifications for Babcock and Wilcox Plants" (NUREG-1430) and are adrdnistrative changes. The proposed changes do not alter the source term, containment isolation, or allowable releases. The proposed changes, therefore, will not increase the radio-logical consequences of a previously evaluated accident.
- 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assump-tions are introduced by the proposed changes, which involve only edmini-strative controls. As described above, these changes are consiste.it with the " Standard Technical Specifications for Babcock and Wilcox Plants" (NUREG-1430) and are administrative changes. The proposed changes do not alter any accident scenarios.
- 3. Not involve a significant reduct. ten in a margin of safety because the proposed changes are administrative and do act reduce or adversely affect the capabilities of any plant structures, systems or components.
CONCLUSION:
On the basis of the above, Toledo Edison has determined that the License Amend-ment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifica-tions that must be reviewed by.the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.
ATTACHMENT:
Attached is a copy of the present Technical Specifications Section 5.0 pages and index, marked up to show the proposed changes, and a copy of the proposed re-vised Technical Specifications Section 5.0 pages and index.
REFERENCES:
- 1. NUREG-1430, Revision 1, " Standard Technical Specifications for Babcock and Wilcox Plants," dated April, 1995.
- 2. " Proposed Policy Statement on Technical Specification improvements for Nuclear Power Reactors," Volume 52, Number 25, dated February 6, 1987.
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- 3. " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," Federal Register, Volume 58, Number 139, dated July 22, 1993.
- 4. 10 CFR 50.36, " Technical Specifications."
- 5. 10 CFR 50.54, " Conditions of Licenses."
- 6. 10 CFR 50.42, " Additional Standards for Class 103 Licenses."
7, 10 CFR 50.55, " Conditions of Construction Permits."
- 8. 10 CFR 50.59, " Changes, Tests, and Experiments."
- 9. 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
- 11. DBNPS Operating License, Appendix A, Technical Specifications, through Amendment 199.
- 12. DBNPS Emergency Plan, Revision 18.
- 13. Calvert Cliffs Nuclear Power Plant Operating License Numbers DPR-53 and DPR-69, Amendment Numbers 204 and 182, dated March 14, 1995 (TAC Numbers M88429 and M88430).
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Present Index Page XIV and Technical Specification 5.0 Pages 5-1 through 5-10 (marked up to show proposed changes) i e
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