ML20072K159
| ML20072K159 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/11/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072K153 | List: |
| References | |
| NUDOCS 8303300401 | |
| Download: ML20072K159 (12) | |
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U'!!TED STATES j#
I'i fiUCLE AR REGULATORY COM.MISSIOTJ fy v.[.~2 r. j n'.' ASHING TO N. D. C. M E55 5 y Q a..!,.i x,.. ;
SAFETY EVALUATIO'4 EY THE OFFICE OF NUCLEAR REACTOR REGULATION SUFPORTING A"ENDME'iT NO. 25 TO FACILITY OPERATING LICENSE NO. D?R-52
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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT,._ UNIT 2 DOCKET NO. 50-260 1.0 Introduction By letter dated October 15,1982 (TVA BFNP TS.179), as supplemented by letters dated November 17, 1982, December 10,1982 and January 7,1983, the Tennessee Valley Authority (licensee) requested changes to the Technical Specificati.ons (Appendix A) appended to Facility Operating License No. DPR-52 for the Browns Ferry Nuclear Plant, Unit 2.
The proposed amendment and~ revised Techncial Speci-fications were to:
(1) incorporate the limiting. conditions for operation associated with fuel -Cycle 5, and (2) reflect changes resulting from design, equipment and procedural modifications made during the current refueling outage.
2.0- Discussion and Evaluation 2.1 Reload Discussion Browns Ferry Unit 2 (BF-2) shutdown for its fourth refueling on July 30, 1982, with a projected restart date in early March 1983.
BF-2 was initially fueled with 764 of the GE 7x7 fuel assemblies containing 49 fuel rods each.
During the first refueling, which began March 18,197E,132 of the 7x7 fuel elements were replaced with one water rod 8x8 fuel assemblies.
In the second refueling, which started April 27,1979, 232 of the 7x7 ~ fuel assemblies were replaced with a like number of two water rod, retrofit 8x8. (2x8R) bundles.
During the second refueling, an additional 36 7x7 fuel asse.blies were also replaced with 8x8 fuel that had origir. ally been procured for fuel Cycle.2 but.not used.
During the third refueling, which began September 5,1953, an additional 240 of the original 7x7 fuel bundles were replaced with prepressurized two water rod SxE retrofit (?Sx8R) fuel J
ass embli es.
The pre;ressurized fuel assemblies are essentially identical from a core physics standpoint to the two water rod fuel assemblies (8x8R) except -that they are prepressurized with about three rather than one atmospheres of helium to minimize fuel clad interaction.
Our evaluation of the F8x8 fuel is discussed in
.the Safety Evaluation attached to our letter of. Ap.ril 16, 1979, to GE approving
-the use of this fuel in Boiling Water Reactor (BWR). reload licensing applicat' ions.
With this reload, the last of the 124 remaining initial 7x7 fuel assemblies were removed from the core. The licensee had also planned to replace 124'of the 8x8 fuel assemblies, making a total of 248 new fuel bundles to be added; During the
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coastdown period for BF-2 that occurred during Cycle 4 operation (from April to July 1982), higher than normal. radioactivity levels were noted in the steam.
During the outage following Cycle 4 operation, the entire core was off-loaded to to be drained for the Mark I torus modi-the spent fuel pool to permit the torusExamination of the removed fuel disclosed that m fications.
The cause was at~tributed assemblies showed evidence of severe waterside corrosion.
to what has been characterized in the industrf 'as '! Crud Induced Localized Corr (CILC)". Detailed inspection of similarly failed fuel at Hatch 1 and Vermont Yankee disclosed that the fuel that failed was mainly gadolinia poisoned fuel rods; the method of failure. as pitting corrosion perforating through the cladding.
w Examination of the fuel removed from BF-2 disclosed 31 fuel assemblies with significant corrosion; these were replaced with new fuel.
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-Thus, some of 'the fuel assemblies that were td be retu'rned'to the ' core for Cycle 5
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were replaced with new fuel and some of the depleted fuel assemblies -that were not The net effect from reus'e of going to be reinserted are now going'to be reused. fuel that was consid weeks less full power capability. The total number of fuel assemblies changed out in the current reload remains at 248.
In support of this' application, the licensee submitted a Supplemental Reload Licensing Report, Y1003J01 A40, (Ref.1) and update to the LOCA analysis report,
-NED0-24088-1, (Ref.2) and a number of proposed changes to the BF-2 Technical The analyses. presented in Ref.1 and 2 were based on adding 248 Specifications.
Both the licensee and the General Electric Company (GE) new fuel assemb. lies.
-reevaluated the analyses in light of the additional new and spent fuel. assemblies
.being added to replace those found corroded and determined that the analyses conservatively bounded the revised coreloading, since the revised coreloading
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plan will have less energy than the loading degcribed in Ref.1.
2.2 Reload Evaluation We reviewed the submittals and eva10ated the nuclear design, the thermal hydraulic design, the transient and accident analyses, and the Technical Specificat The fuel mechanical design Because of our review of a large number of generic considerations related changes.
(Ref.3) to use of 8x8, 8x8R and P8x8R fuels in mixed loadings, and on the basis of the evaluations which have been presented in Reference 3, only a limited number of
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For evalua-additional areas of review have been included in the Safety Evaluation.
tions of areas not specifically addressed in this Safety Evaluation refer to_
Reference 3.
2.2.1 Nuclear Design With the exception of the shutdown margin and standby liquid control system analyses, the nuclear parameters applicable to the Cycle 5 core were obtained by methods and techniques described in Reference 3, which has been approved by the W
staff for this purpose. The results were within the range normally encountered in BWR relcads and are acceptable. The shutdown margin and standby liquid control system analyses were performed by the licensee using.its core simulator and lattice physics methods which have been reviewed and approved by the staff. The shutdown margin wa's 1.4% reactivity change with the strongest rod out. The standby liquid control system is capable of making.the unrodded core subcritical at 200C with a margin of 2.3% reactivity change. These are acceptable margins and
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therefore, we conclude that the nuclear design parameters for the Cycle 5 core are acceptable.
2.2.2 Thennal-Hydraulic Evaluation The thermal-hydraulic review includes the following areas:
(1)safetylimit minimum critical power ratio (MCPR), (2) operating limit MCPR, and (3) thermal hydraulic stability. The objective of this review is to confirm that the thermal hydraulic design of the reload has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal -operation and anticipated operational transients, and is not susceptible to thermal hydraulic instability.
Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 3, for BWR cores which reload with GE's retrofit 8x8 fuel, 4
the safety limit minimum critical power ratio (SLMCPR) resulting from either core-wide or localized abnormal operational transients is equal to 1.07.
When meeting this SLMCPR during a transient, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition. The 1.07 SLMCPR is unchanged from the SLMCPR previously approved. The basis for this safety limit is addressed in Reference 3.
Operating Limit MCPR Various transients could reduce the MCPR below the intended safety limit MCPR l
during Cycle 5 operation. The anticipated operational transients have been analyzed by the licensee to determine which could potentially induce the largest reduction in the initial (CPR). Operating ' cycle MCPR values for this plant specific cycle are as expected for the BWR/4 design with the fuel types that are presented in Cycle 5 of BF-2, and compare favorabl.y'with the MCPR for operating plants such as Brunswick Unit 2 Cycle 5, previously approved.
l Thermal-Hydraulic Stability f
The results of the thermal-hydraulic analysis (Ref.1) show that the maximum.
thermal-hydraulic stability decay ratio is 0.74 for this cycle. Because operation in the natural circulation mode is prohibited by Technical Specifications, there will be added margin to the core stability and therefore, we find the thermal-l hydraulic stability acceptable.
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Sr sry of Thermal-Hydraulic Evalcation The analysis of EF-2 Cycle 5 has been performed with standard technicues described in F.eference I and the values of MCPR and thernal-hydraulic stability decay ratio are as expected for the EWR/4 design and we, therefore, find the thermal-hydraulic design for Cycle 5 acceptable.
2.2.3 Transient and Accident Analyses The transient and accident analyses were performed by'the _ methods and procedures described in GE report NED0-24011-P-A-4, which we previously approved.
It was necessary to perform a cycle ~ specific analysis of the rod drop accident since the accident reactivity shape function for Cycle 5 was not bounded by the. generic shape for the cold startup case described in Reference 3.
The results of this analysis, which are presented in Section 15 of Reference 1. show that the resultant peak enthalpy, cold, was 264.5 cal /gm. The resulting peak enthalpy rise was less than the acceptance criterion of 280 calories per gram and is acceptable.
The GE method for analysis of misoriented and misloaded bundles has been reviewed and approved by the staff and is part of the Reference 3 methodology. Potential fuel loading error's involving misoriented bundles and bundles loaded into incorrect positions have been analyzed by this methodology and. the results are reported in Section 14 of the supplemental reload submittal. The analyses determined that the
!.CPR for a misoriehted fuel bundle was less than the aCPR f6r the ' limiting t'raniisnte and therefore, sie conclude that' the analyses perf'ormed by GE for. fuel loading ' err'org
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1s accepta~ble.
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2.2.4 Core Reload Technical Specification Changes for Cycle 5 During the refueling for Cycle 5 the last of the 7x7 fuel will be removed from Accordingly all references to this fuel are being deleted from' the the core.
This.is acceptable.
In addition, Technical Specifica-Technical Specifications.
tions 2.1.A, Table 3.1.'A, and Table 4.1.A have been changed to reflect the alteration of the flow biased neutron. flux trip to a thermal power monitor and the addition of a separate high neutrdn flux ~ trip at 120% of full power. This change has been found acceptable for several boiling water reactors and therefore, is acceptable for BF-2.
The MAPLHGR values in Table 3.5.1-3 have been extended to burnup values of 40,000 mwd /t and Table 3.5.1-4 has been added.to provide MAPLHGR values for the new fuel type introduced for this reload. The MAPLHGR tables were obtained by standard methods (Ref. 2) and are acceptable)
This is the first reload for BF-2 for which the overpressurization transients were analyzed with the ODYN code. Specification 3.5.k and Surveillance Require-ment 4.5.k have been revised to reflect the new analyses. This procedure, which wag previously reviewed and found acceptable by the staff, is being introduced en all operating EURs and therefore, is acceptable for BF-2
2.2.5 Summary of Core Reload Evaluation On the basis of our review, which has included the nuclear design, thermal-hydraulic design, transient and accident analyses, and Technical Specification changes, we conclude that operation of BF-2 for. Cycle 5 will not endanger the health and safety of the public. This conclusion,is based on the fact that approved methods have been used to perform the various analyses and that the results are consistent with those for other BWR/4 reactors.
3.0 Plant Modifications 3.1 Discussion BF-2' shutdown for the present refueling and maintenance outage on July 30, 1982, and is projected to be down for over seven months. The reason for the extended outage is the time needed to complete a nu' ber of NRC-required modifications, as m
well as the inspections, repairs, surveillance, maintenance, and other_ activities normally associated with a refueling outage. During this shutdown, the licensee expects to complete numerous modifications which NRC has proposed or required for. operating reactors, such as Browns Ferry, in 'various Bulletins, Orders, the TMI-2 Action-Plan -(NUREG-0737), new regulations, revisions to the Security Plan and Emergency Response Plan, resolution of generic issues, etc. Some of these modifications require changes to the Technical Specifications prior to startup and are included in this Safety Evaluation.
3.2 Evaluation Torus Modifications On January, 13, 1981, the Commission issued an Order modifying the BF-2 license to require the licensee to promptly institute a reassessment of the containment design for.. suppression pool hydrodynamic loading conditions and to install any plant modifications needed to conform to the staff's Acceptance Criteria, which are contained in Appendix A to NUREG-0661 (" Safety Evaluation Report, Mark I l
Containment Long-Term Program" dated July 1980) by March 31, 1982. This Order was subsequently modified by an Order dated January 19, 1982, extending the time to complete some of the modifications to the Cycle 6 outage.
These modifications are required by NRC to restore the originally intended margins of safety in the containment design. The structural modifications to the torus.
containment include addition of torus tiedowns, addition of ring' girder reinforce-ment and reinforcing attached piping nozzles. Ven't syitem modifications include shortening the downcomers, adding local' reinforcement to the vent. header, and __
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adding new tie bars to the downcomers. Attached piping is being strengthened l
including modification of the ECCS header support. Many changes are being made I
to the safety relief valve (SRV) piping system including adding quencher arms to the ramshead, adding quencher arm and ramshead supports, adding 10-inch vacuum i
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7 valves, ' reinforcing the ring girder at the SRV hanger attachment,' rerouting of' piping,1and adding new snubbers and suppprts for the piping. These modifications to the torus require changes to the' Technical. Specifications to account-for water
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i displaced by' the-additional structural steel and to reflect the plant unique
. analysis _ which the licensee was required to perform to assure conformance of the
. design to the staff's Acceptance. Criteria in NUREG-0661. The specific changes to the Technical Specifications are discussed below.
Pages '227, 267 and 269 - The minimum torus water level limits in Section 3.7.A.1.a and in the bases for this section are being changed from -7 inches (differential.
pressure control greater than 0 psid) to -6.25 inches and from -8 inches.(0 psid-i differential pressure ' control ) to -7.25 inches; a change in each case.of 0.75 inch.
There are 15-inch by 15-inch sealed bex beams being added as support for.the safety relief valve lines and HPCI-RCIC internal supports. Addition of these supports will result in appreciable water displacement. Calculations indicate.
that the box-beams and HPCI-RCIC supports will increase the torus water level approximately 3/4-inch-due to their presence. This rise in the torus water level is reflected in these revised Technical. Specification values. The changes, which we have reviewed and approved, are necessary to ensure that the minimum water volume is maintained in the torus-for suppression of potential LOCA loads and are acceptable.
(This same change to the Tecanical Specifications was made by Amendment No. 51 to Facility License No. DPR-68 for BF-3 issued March 29, 1982.)
Pages 235a and 269.- In Section 3.7.A. 6.a-(and the bases thereto), the setpoint for.the drywell-suppression chamber (wetwell) differential pressure control (LP) is being changed from 1.3 psid to 1.1 psid.
Downcomer water clearing loads are greatly reduced'by physically shortening the downcomers (by almost one foot) and-imposing a 'drywell-wetwell AP.
The Browns Ferry unique loads were determined'by considering a differential pressure of 1.10 psid at the maximum allowable torus water leve1~. In order to be consistent with.this analysis, the Technical Speciff-cation ' associated with the AP control has been established at 1.10 psid. The changes to the Technical Specifications confonn to the requirements in Section
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2.16, " Differential Pressure Control Requirements," in Appendix A totNUREG-0661 and are therefore, acceptable.
Pages 233, 234, 267. and 268 - The " Bases" section for Specifications 3.7.A and 4.7.A for the suppression pool temperature limits was based on the Humboldt Bay and Bodega Bay tests. Consistent with the long-term torus integrity program of NUREG-0661 and NUREG-0783, the " Bases" require change to~ account for steam mass fluxes through SRV T-quenchers. During the current refueling outage, the e
T-quenchers are being added to the safety-relief valve discharge. device.
In S ction 2.13.8 of Appendix A to NUREG-0661 (" Suppression Pool Temperature Limits")
._the staff 'specified that "the suppression pool ' local temperature shall not exceed 2000F throughout all plant transients involving SRV operations." The licensee's analyses determined that at reactor vessel ' pressures above approximately 555 psig, the bulk pool temperature will not exceed 180 F.
At pressures below approximately 0
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- 0 240 psig, the bulk temperature will not exceed 184 F.
Both temperatures are well below the acceptabl e limits. These temperatures also represent the bounding upper limits that are used in _ suppression pool temperature response analyses for safety relief valve disenarge and LOCA cases. The actions requ. ired by Specifica-tion 3.7.c-f assure the' reactor can be depressurized in a timely manner to avoid exceeding the maximum buik suppression pool water limits. Furthermore, the 184 F limit provides that adequate RHR and core spray putnp NpSH will be available without dependency on containment overpressure. Section 4.7.A.2.k of the present Technical Specifications requires that if extended relief valve operation causes the tempera-ture of the suppression pool to exceed 130 F, the reactor shall be shutdown 'and 0
the torus and drywell visually inspected for signs of distress or displacement.
Since the torus is being extensively upgraded to withstand dynamic loading significantly beyond that originally expected, extended operation of relief valves above' a suppression pool temperature of 1300F is not expected to be a safety con-cern warranting placing the reactor in cold shutdown and performing a torus inspection. Therefore, this requirement is being deleted.
Page 256 - Table 3.7.B has been revised to include penetration X-223. This penetration has been installed to provide another suppression chamber access' hatch to facilitate the torus modifications.
Pace 266 - Table 3.7.H has been revised to include temporary electrical penetration X-200A-SC which is integral to torus access hatch X-200A. This electriccl penetra-tion is designed to accommodate instrumentation for the SRV-torus integrity test program. This penetration is to be removed at the first opportunity following the test progr:m.
Page 273 - The present Technical Specifications in the " Bases" for prim'ary contain-ment, discuss the specific type of protective coatings applied to the drywell and torus surfaces to protect the steel from corrosion and minimize contamination of
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the water. There have been significant developments in protective coating technology since the Brcwns Ferry units were licensed. During the torus modifica-tions, the licensee has thoroughly sandblasted all torus surfaces in each unit and is applying coatings that offer more potential for sealing the surfaces. Therefore, the " Bases" are being generalized so that a technical specification change will not be required if a different protective coating is applied.
Page 145 - Section 4.5.5.1 of the Technical Specifications requires that every three months, the LPCI capability of the RHR pumps shall be demonstrated. In the tests, the pumps take suction from the torus and return the water to the torus. The pumps are required to demonstrate that two pumps in the same loop can deliver at least 15,000 gpa against an indicated system pressure (head) of,200 psig.
The two-pump 15,000 gpm LPCI test surveillance was determined to induce vibrations in the RHR return line o the torus. To eliminate the vibration, an orifice has been installed in the return line. However, installation of this orifice plate also decreases the suppression pool cooling mode of RHR operation from 15,000 gpm
o to approximately 12,000 gpm. A new containment cooling analysis was performed for this configuration, and it was determined that this flow rate induces a long-term suppression pool temperature well within that necessary for stable and complete steam condensation and for adequate RHR and core spray pumps net positive suction head.
The revised test requirement is that the two pumps. demon-strate that they can deliver 12,000 gpm against a hig'her head - 250 psig.
The orifice is in the return line to the torus arid doe's not change the volume of water that would be injected into the reactor during the LPCI mode..The 12,000 gpm at higher pump head pressure is equivalent to 15,000 gpm at lower discharge pressure. We conclude that the change has no adverse impact on the LPCI or containment cooling modes of RHR operation and is acceptable.
480V M0V Boards Tie-In and LPCI M-G Sets Installation Pages 293a, 297b, 298, 300 and 330 - Amendment No. 45 to Facility License No.
OPR-52 for BF-2 dated May 11, 1979 adds a license condition authorizing modifica-tions to the power supply for certain LPCI valves. The modification ensures that the 480V ac reactor M0V boards, with the associated autotransfer feature, will. be isolated from the redundant divisional power supplies. The modifications are designed to eliminate the recirculation loop selection logic and to rewire the accident initiati4n signals to direct both LPCI injection valves to open upon detection of accident conditions. The modifications include installation of qualified Class lE motor-generator (MG) sets to serve as isolation devices be-tween the redundant divisional 480V shutdown boards (power sources) and the swing bus (auto-transfer) of the 480V reactor MOV boards tha'; supply motive power to the LPCI valve operators.
In 1976, the NRC staff requested ths licensee to propose modifications to eliminate the LPCI systems recirculation loop selection logic to eliminate a potential single failure concern. As noted above, the design was approved by Amendment Nos. 51, 45 and 23 for Units 1, 2 and 3, respectively, on May 11, 1979. The modifications require changes to the Technical Specifications which are incorporated herein.
The associated Technical Specifica-tions are consistent with those approved in Amendment No. 75 to Facility License No. DPR-33 for BF-1 dated September 3,1981.
Thermal Power Monitor Pages 8,10, 20, 22, 33, 36a and 37 - During this outage, the licensee has installed a flow-biased simulated thermal power monitor. These monitors are installed on most all BWRs; the justificrtion for these monitors is. discussed in the " Bases" for the APRM' settings in the BWR Standard Technical Specifications (BWR/4 STS, Section 2.2.1, page B2-7). The monitors are installed to have the APRM flow biased neutron flux signal respond to the thermal flux rather than the-neutron flux by accounting for the approximately six-second thermal time constant of the fuel. The proposed changes to the Technical Specifications are acceptable,
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since they are based on previously reviewed and accepted changes for similar BWRs.
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9-Scram Discharge Instrument Volume Pages 37, 39, 40 and 126
.The long-term modifications to the scram discharge instrument volume (SDIV) necessary to resolve problems related to the partial rod insertion event are being implemented during this outage for BF-2.
To upgrade the reliability of the SDIV instrumentation, two of the float-type pressure switches are being replaced by diverse differential pressure swite.hes. Tables 4.1.A and 4.1.B are therefore being revised to add these switches to the list of instruments that require surveillance testing.
Containment Vent and Purge System In response to NRC generic letters of September 27, 1979 and October 22,1979 to "All Light Water Reactors," the licensee is modifying the containment purge system for BF-2 during this outage to satisfy app]icable requirements of NRC Branch Technical Position CSB 6-4 regarding valve closure times and addition of debris Pages 251 and 252 are bein'g revised to reflect the significant reduction screens.
in the maximum allowable operating time. On the nitrogen purge valves, the operating time is being reduced from 10 seconds. to 5 seconds and on the purge inlet and isolation valves, the operating time is being reduced from 90 seconds to only 2.5 seconds. The faster valve closure times significantly reduce potential-offsite doses. The addition of the debris screens provides protection against foreign material entering the purge ducting and interfering with closure of the purge valves.
These same changes to the Units 1 and 3 Technical Specifications were made respectively by Amendment No. 76 to License No. DPR-33, issued September 15, 1981, l
and by Amendment No. 51 to License No. DPR-58 issued March 29, 1982. Since the
. changes tr; the Technical Specifications for BF-2 are those requested by our letter of Decem'>er 17, 1981 and have been previously, approved for BF-1 and BF-3, they l
are acceptable for BF-2.
l Primary Containment Isolation Valves Tables 3.7.A through 3.7.H list the various valves and penetrations associated with primary containment isolation. Specifically, Table 3.7.A lists the primary
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containment isolation valves that must be operable during reactor power operation (in accordance with Section 3.7.0 o,f the Technical Specifications) along with the maximum operating times and normal position. Table 3.7.D lists the primary con-tainment isolation valves on which local leak rate tests must be performed each cycle in accordance with Section 4.7.2.g.
Tables 3.7.E, 3.7.F and 3.7.G list the stop-check and check valves on the torus and drywell influent lines that must be similarly tested. As discussed below, the licensee has proposed revisions sto.
these tables to reflect plant modifications and the requirements in NUREG-0737 l
Item II.E.4.2.
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Tables 3.7.D through 3.7.G have been completely revised to be more consistent with the BWR/4 Standard Technical ~ Specifications. These tables contain a " Test Medium" and " Test Method."' The proposed tables have been revised to contain the " Test Medium" 'within the title of the table and eliminate the " Test Method" altogether. The Standard Technical Specifications do not contain a test method fo'r testing isolation valves.
In addition, the te.st methods for these valves are Lcontained in their specific testing instructions and therefore should.not be contained in the Technical Specifications. The deletion of the test methods from the table does not have any adverse impact on safety. Similar changes were made I
to the same tables for BF-3 in Amendment No. 51 issued March 29, 1982. As part l
of the revisions to-these tables, the licensee has proposed to air test certain isolation valves that were previously water tested. Appendix J to 10 CFR Part 50 specifies air testing as the recommended leak testing method, except for those
. valves that are fluid sealed.
In addition, the staff considers air testing of valves to be a more conservative method than water testing., On the basis of the 5
~information provided by the licensee in the submittal of October 15,1982, a.nd the requirements of 10 CFR Part 50, Appendix J, we conclude that the proposed changes to the Technical Specifications with respect to the test medium are acceptabl e.
i Most of the changes to the tables on isolation valves are to add or delete valves which the inservice pump and valve testing program indicated should be verified L
- for operating time, to correct valve numbers or to correct valve positions. Each proposed change is described in detail in the licensee's ' submittal of October 15,.
1982. We have reviewed each change and concluaed they are acceptable, since they are consistent with modifications either deleting or adding valves.
NUREG-0737, Item II.K.3.15 TMI Action Plan Item II.K.3.15 requires licensees of BWRs to modify pipe-break-detection circuitry so that pressure spikes resulting from HPCI and RCIC initiation will' not cause inadvertent system isolation. The licensee elected to l
employ the BWR Owners Group modification which incorporates a three second time
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delay relay (TDR) to prevent spurious isolation.
In our letter to the licensee of.0cto ber 13,'1981, we requested the licensee to provide certain analyses and to " propose the appropriate Surveillance Requirements and Limiting Conditions of Operation for the HPCI and RCIC systems which address this item." The safety evaluation was provided by the licensee's letter of December 16, 1981. All of the Browns Ferry units have had a three-second TDR on the HPCI systems.
During
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the current outage for BF-2, a 'TDR was added to the RCIC system. The proposed changes to the Technical Specifications requiring calibration and surveillance of the time delay relays was submitted with the licensee's application of October 15, 1982. Table 4.2.B (p99) is being modified to require a logic system functional test, including calibration of the RCIC and HPCI system isolation logic. The changes to.the Technical Specification reflect the surveillance requirements requested in our letter of October 13,1981 'on Item II.K.3.15 and are acceptable.
The same changes were made to the BF-3 Technical Specifications in Amendment No.
51 issued March 29, 1982.
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4.0 Administrative Chances The. licensee has proposed eight administ'rative changes to the BF-2 Technical Specifications; the licensee has described and justified each change in its submittal of October 15, 1982. The changes are to revise the Table of Contents, to reformat one section, to correct or add referen.ces or to delete reference to a table that was removed by a previous amendment. These changes do not affect any actual limiting conditions for operation. 'We conclude that these proposed changes are editorial in nature and do not alter the techncial bases of the.
specifications and therefore, are acceptable.
5.0 Environmental Considerations We ha've determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant.to 10 CFR 51.5(d)(4),
that an environmental impact statement or negat,1ve declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
6.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1). because the amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant reduction in a safety margin, the amendment does not involve a significant hazards consideration (2) there is reasonable assurance that the health and safety of the public will not be en' dangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commissions regulations and the issuance of the amendment will not be inimical to the common defense ar.d security or to the health and safety of the public.
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Dated: March 11,1983 Principal Contributors:
Walt Brooks, Dick Clark, Jim Hall l
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6-o 7.0 ~ References -
~1.
Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant, Unit 2 Reload No. 4 (Cycle 5)," Y10'03J01 A40.~ ~ '-
2.
'EO-24088-1, "LOCA Analysis for Browns Ferry Nuclear Plant Unit 2."
3.
NEDE-24011-P-A-4, " General Electric Standard Application for -
Reactor Fuel," January 1982.
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