ML18036B274
| ML18036B274 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/30/1993 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18036B273 | List: |
| References | |
| GL-88-16, NUDOCS 9305110278 | |
| Download: ML18036B274 (17) | |
Text
2.1 BASES (Cont'd)F.(Deleted)G.&H.Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.Operation of the reactor at pressures lower than 825 psig requires that the reactor, mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams.Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.With the scrams set at 10 percent of valve closure, neutron flux does'not increase.I.J.&K.Reactor Low Water Level Set oint for Initiation of HPCI and RCIC Closin Main Steam Isolation Valves and Startin LPCI and Core S ra Pum s.These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive,clad temperatures.
The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints.
Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.L.References 1.Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 1 (applicable cycle-specific document).
2.GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version).3."gualification of the One-Dimensional Core Transient Model for Boiling Water Reactor," NED0-24154-P, October 1978.4.Letter from R.H.Buchholz (GE)to P.S.Check (NRC),"Response to NRC Request For Information On ODIN Computer Model," September 5, 1980.BFN Unit 1 1.1/2.1-16 9305110278 930430 PDR ADOCK 0500025'P P PDR 6.9.1.7 CORE OPERATING LIMITS REPORT a.Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:
(1)The APLHGR for Specification 3.5.I (2)The LHGR for Specification
.3.5.J (3)The MCPR Operating Limit for Specification 3.5.K/4.5.K b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel" (latest approved version).c.The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits)of the safety analysis are met.~d.The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.BFN Unit 1 6.0-26a
PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 331i SUPPLEMENT
- 1)
2.1 BASES
(Cont'd)F.(Deleted)G.&H.Main Steam line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.The scram feature that occurs when the main steamline isolation valves close shuts down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams.Thus, the combination of main steamline low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.With the scrams set at 10 percent of valve closure, neutron flux does not increase.I.J.&K.Reactor Low Mater Level Set oint for Initiation of HPCI and RCIC Closin Main Steam Isolation Valves and Startin LPCI and Core S ra Pum s.These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures.
The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints.
Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.References 1.Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 2 (applicable cycle-specific document).
2.GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version).BFN Unit 2 1.1/2.1-16
~6.9.1.7 CORE OPERATING LIMITS REPORT a.Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any re'maining portion of an operating cycle, for the following:
(1)The APLHGR for Specification 3.5.I (2)The LHGR for Specification 3.5.J (3)The MCPR Operating Limit for Specification 3.5.K/4.5.K b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel" (latest approved version).c.The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.BFN Unit 2 6.0-26a PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 3 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 331~SUPPLEMENT
- 1)
2.1 BASES
(Cont'd)\F.(Deleted)Q.E H.Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 850 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRN and APRN high neutron flux scrams.Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.Mith the scrams set at 10 percent of.valve closure, neutron flux does not increase.I.J.E K.Reactor Low Mater Level Set oint for Initiation of HPCI and RCIC Closin Hain Steam Isolation Valves and Startin LPCI and Core These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive.clad temperatures.
The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints.
Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.L.References 1.Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 3 (applicable cycle-specific document).
2.GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version).BFN Unit 3 1.1/2.1-16
.6.9.1.7 CORE OPERATING LIMITS REPORT a.Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:
(1)The APLHGR for Specification 3.5.I (2)The LHGR for Specification
'3.5.J (3)The MCPR Operating Limit for Specification 3.5.K/4.5.K b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel" (latest approved version).c.The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The CORE OPERATING LIMITS REPORT, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.BFN Unit 3 6.O-26a ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)UNITS 1p 2g AND 3 (TVA BFN TECHNICAL SPECIFICATION AMENDMENT 331'UPPLEMENT 1)REASON FOR THE CHANGE DESCRIPTION AND JUSTIFICATION REASON FOR THE CHANGE These proposed changes to the BFN technical Specifications are administrative in nature.The changes are to revise Specifications 6.9.1.7.b and 6.9.1.7.d and Bases 2.1, Reference 2 to agree with the guidelines of Generic Letter (GL)88-16,"Removal of Cycle-Specific Limits from Technical Specifications.<<
DESCRIPTION OF THE PROPOSED CHANGE 1.For Units 1, 2, and 3, Reference 2 to Bases 2.1, page 1.1/2.1-16 reads: "GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (applicable amendment specified in the CORE OPERATING LIMITS REPORT)" The revised Reference 2 reads: "GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version)2.For Units 1, 2, and 3, the proposed Specification 6.9.1.7.b reads: "The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel" (applicable amendment specified in the CORE OPERATING LIMITS REPORT).The revised specification reads: "The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A,"General Electric Standard Application for Reactor Fuel" (latest approved version)"
ENCLOSURE 2 (Continued)
Page 2 oS 2 3.For Units 1, 2, and 3, proposed Specification 6.9.1.7.d reads: "The CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle STARTUP, for each reload cycle or within 30 days of issuance of any mid-cycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector".
The revised specification reads: "The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC".JUSTIFICATION FOR THE PROPOSED CHANGE The proposed changes to Reference 2 to Bases 2.1 and proposed Specification 6.9.1.7.b clarify the requirement to use the latest approved version of NEDE-24011-P-A.
The proposed change to proposed Specification 6.9.1.7.d clarifies requirement to submit the Core Operating Limits Report upon its issuance.These proposed changes are to make the BFN TS consistent with the guidelines of GL 88-16.These clarifications apply to TS-309 submitted on August 20, 1992 and revisions submitted on March 18, 1993, as part of TS-331.The evaluation and no significant hazards considerations provided by the August 20, 1992 and March 18, 1993, submittals are applicable to this change.
ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION DESCRIPTION OF THE PROPOSED TECHNICAL SPECIFICATION CHANGE The BFN technical specifications are being revised as follows: 1.For Units 1, 2, and 3, revise Reference 2 to Bases to include the requirement to use the latest approved version of NEDE-24011-P-A.
2.For Units 1, 2, and 3, revise Specification 6.9.1.7.b to include the requirement to use the latest approved version of NEDE-24011-P-A.
3.For Units 1, 2, and 3, revise Specification 6.9.1.7.d to include the requirement to submit the Core Operating Limits Report upon issuance to the NRC.BASES FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION 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 if operation of the facility in accordance with the proposed 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.
The proposed changes are administrative in nature.They are being made to clarify administrative requirements in previous proposed technical specifications submittals.
These changes do not affect any of the-design basis accidents.
They do not involve an increase in the probability or consequences of an accident previously evaluated.
lIt ENCLOSURE 3 (Continued)
Page 2 oS 2 2.Th'e proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes are administrative in nature.They are being made to clarify administrative requirements in previous proposed technical specifications submittals.
No modifications to any plant equipment are involved.There are no effects on system interactions made by these changes.The changes will revise the technical specifications so that they are more accurately reflect the guidelines of GL 88-16.They do not create the possibility of a new or different kind of accident from an accident previously evaluated.
3.The proposed amendment does not involve a significant reduction in the margin of safety.The proposed changes are administrative in nature.They include the clarification of some requirements to ensure consistent application of the GL 88-16 guidelines.
No safety margins are affected by these changes.CONCLUSZON 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.
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