ML20010J237
| ML20010J237 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/15/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010J236 | List: |
| References | |
| NUDOCS 8109300010 | |
| Download: ML20010J237 (9) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO.76 TO FACILITY LICENSE NO. DPR-33 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-259 1.0 Introduction By letter dated April 29,1981 (TVA BFNP TS 161), which was supplemented by letters dated June 12, 1981 and July 13, 1981, tha Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Speci?ications (Appendix A). appended to Facility Operating License No.
OPR-33 for the Browns Ferry Nuclear Plant, Unit No.1.
The proposed amendment and revised Technical Specifications would (1) incorporate the limiting conditions for operation of the facility in the fifth fuel cycle following the fourth refueling of the reactor and (2) reflect new primary containment atmospheric hydrogen monitoring instrumentation being installed during the current refueling outage.
In support of this reig document \\pq application, TVA submitted a supplemental reload licensing
> prepared by the General Electric Company (GE), errata and Nuclear Plant Unit il2(1 (originally issued September 1977) also preparedadd by GE and proposed changes to the Technical Specifications.
2.0 Dir,cussion Browns Ferry Unit No, I (BF-1) shutdown for its fourth refueling on April 11, 1981. BF-1 was initially fueled with 764 of the General Electric Co. (GE) 7 x 7 fuel assemblies containing 49 fuel rods each.
During the first refueling,166 of the 7 x 7 fuel assemblies were replaced with a like number of one water rod 8 x 8 fuel assemblies containing 63 fuel reds each.
During the second refueling, an additional 156 of the original fuel assemblies were replaced with two water rod retrofit 8 x 8R fuel bundles containing 62 fuel rods each. During the
. third refueling outage, another 232 of the 7 x 7 fuel bundles were replaced with P S x 3 fuel assemblies, each containing 62 fuel rods.
The prepressurized fuel assemblies (P 8 x 8R) are essantially identical from a core physics standpoint to the two water rod fuel assemblies (8 x 8R) except that they are prepressurized with about three rather than one atmospheres of helium to minimize fuel clad interaction. Our evaluation of the P 8 x 8R fuel is discussed in the safety evaluation attached to our letter of April 16, 1979 to General Electric approving the use of this fuel in BWR reload licensing applications. The larger es109Jooo10 810915'.
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. t inventory of. helium gas improves the gap conductance between fuel pellets'and r.ladding resulting in reductions in fuel temperatures, thermal expansion and fission gas release. The pressurized rods operate at. effectively lower linear heat generation rates and are therefore 6tpected to yield performarce benefits in terms of fuel reliability. The increased prepressurization also results in improved margin to MAPLHGR limits by reducing stored energy.
During the current refueling outage, all of thGmaining 214 original 7 x 7 fuel bundles will be replaced.along with 46 of the 8 x 8 fuel assemblies.
Thus, a total of 260 new fuel assemblies will be loaded in the core, consisting of 256 of the P8 x 8R fuel bundles and 4 lead test assemblies (two GLTA-1 and two GLTA-2). The four lead test assemblies (LTAs)
'are exactly the same as the standard P90RB284L {P8 x SR). reload bundle fuel except for a small axial section of increased Gadolinia content in some rods. Test measurements will be performed on these bundles during Cycle 5 to benchmark the effect of this increased Gadolinia content.
All approved thermal-mechanical and reload methods described in NEDE-240ll-P-A', " General Electric Standard Application for Roload," will hold for these LTAs.
With this refueling, Browns Ferry Unit 1 will continue to be on an 18 month refueling cycle. Units Nos. 2 and 3 are also on 18 month refueling cycles.
7 As noted above, this reload involves loading of prepressurized GE 8 x 8 retrofit (P8 x 8R) fuel. This is the same type of fuel as was loaded during the last reloads for all three Bro:vns Ferry Units. The description of the nuclear and mechanical designs of 8 x 8 retrofit fuel is contained in References 3 and 4.
Reference 3 also contains a complete set of refer-ences to topical reports which describe GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and information regaraing the applicability of these methods to cores containing a mixture of fuel. The use and safety implications of pre-
. pressurized fuel have been found acceptable per Reference 4 The conclusions of Reference 5, which was cited above, found that the methods of Reference 3 were generally applicable to prepressurized fuel. Therefore, unless otherwise specified, Reference 3, as supported by Reference 5, i; adeqt. ate justification for the current application of prepressurized fuel 3.0 Evaluation 3.1 Reactor physics The reload application follows the procedure described in NEDE-240ll-P,
" Generic Relcad Fuel Application." We have reviewed this application and the consequent Technical Specification changes. The transient analysis input parameters are typical #or BWRs and are acceptable. Core wide transient analysis results are given for the limiting transients and the required operating limit values for MCPR are qiven for each fuel type. The revised MCPR' limits are required b-9e reload and they are acceptable.
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.3-3.2 Thermal-Hydraulics As stated in Reference 3. for BWR cores which reload with GE's retrofit 8x8R fuel, the safety limit minimum critical power ratio (SLMCPR) resulting from either core-wide or localized abnormal operational transier.ts 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.
To assure that the fuel cladding integrity SLMCPR will not be violated during any abnormal operational transient or fuel misloading, the most limiting events have been reanalyzed for this reload by the licensee, in order to determine which-event resdits in the largest reduction in
'the minimum critical power ratio. These events have been analyzed for the. exposed fuel and fresh fuel. Addition of the largest reductions in s
critical power ratio to the SLMCPR was used to establish the operating limits for each fuel type.
We have found the methods used fer this analysis consistent with previously approved past practice (Reference 3). We have found the results of this analysis and the correspcnding Technical Specification changes acceptable.
3.3 ECCS Appendix K input data and results for ECCS analysis have been given in References J
l and 2.
The information presented fulfills the requirements for each analyses outlined in Reference 3.
We have reviewed th2 analyses and information submitted for the reload and conclude that BF-1 will be in conformance with all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 when it is operated in accor-
. dance with the Technical Specifications we are issuing with this an.endment.
Supplemental calculations that address the issues of NUREG-0630 have also beers given f r. Reference 2.
3.4 Chr.ges to Technical Specification
-Our evaluation of the specific changes to the Technical Specifications resulting from the current reload is presented below:
Pgs. 9,16,131 and 160 - Since this reload removes the last of the original 7-x 7 fuel elements, the linear heat generation rate limit on these fuel elements is no lonjer pertinent and is b.ng removed from the Technical Specifications and the. bases, r-
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I' x U' Pgs.19, 25 and 169 - This is the first reload for BF-1 in which the transients were analyzed by General Electric's ODYN Code as required by the staff. An additional citation is being added to the Technical
' Specifications to reference NRC's approval of this Code for core reloads.
Pgs.19 and 221 - Section 2.1 of the Technical Specifications contains the bases for the " limiting safety system settings related to fuel cladding integrity." At the bottom of page 19 there is presently a paragraph relating to operatien in the natural circulation mode. This paragraph is being moved, v:rbatim, to the bases for recirculation pump operation on-p. 221, which is a more appropriate location. There is no safety significance to this reformating of the Technical Specifications.
Pgs. 30 and 219 - In Sections 2.2 (. bases for reactor coolant system integrity) and 3.6.0/4.6.0 (bases for relief valves), the value for the total capacit*, c.' the 13 relief valves' is being ircreased from 82.6%
to 83.9*..
The value of 83.9 percent total relief capacity is derived from the values of 77.46 percent for 12 SRV's operable out of a total of i3 SRV_'s.
The capacity of 77,46 percent of nuclear boiler rated steam flow, as listed in the BF 1 Relcad 4 Supplemental Reload Licensing sub-mittal, was calculated based on certified valve capacity for a 5.125-inch throat-diameter valve -(869,000 lbs/ hour at 1,090 +3 psig) issued by the ASME National Board of Boiler and Pressure Vessel Inspectors.
-The certified values are obtained by testing and are listed as 90 percent of the measured capacity values for conservatism. The proposed change is supported by the reload submittal and is acceptable.
Pgs.122,123,124 and 129 - As described in the discussion section of
.this safety evaluation, the reload for BF-1 will contain four LTAs. In order to obtain additional physics data, special cold criticality tests t
have been planned far this cycle. These criticality tests require suspension of the rud sequence control system (RSCSl constraints by means of the indivMual rod bypass switches. This testing is planned as part of the Lead Test Assembly program in which TVA and GE are participating.
We'have been kept appraised of this program through di<cussions and meetings, such as the meeting between TVA, GE and NRC staff in Bethesda, Md. on July 14, 1981 The naxt aspect of the program will include loading of four LTA's in the October 1981 refueling of Browns Ferry Unit No. 3.
An analysis was performed to show that a postulated rod drop accident involving control rods withdrawn during the cold critical test would not
. exceed the peak fuel enthalpy design limit of 280 cal /gm.
The rod worth minimizer (RWM) will be programmed t.o ensure adherence to the withdrawal sequence specified in the cold critic.al test procedure. The RWM must be operable for this test; a second licensed operator may not be used in lieu of the RWM for this testing. The proposed changes in the RSCS below 20% rated power - in conjunction with the compensatory measures -
is acceptable.
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. 4 Pgs 143 and 145 - These changes are administrative changes that remove e
references to nonapplicable technica' specification requirements. These changts do not affect any actual limiting conditions for operation; therefore, p4rt safety is not affected.
Pgs.158 and 159 - As a result of previous changes to the Technical Specifications, sections 3.5.H and 4.5.H (Maintenance of Filled Discharge Pipe) is now located on two pages with haif a page in between the lead sentence and the requirements. The proposed change is to relocate the parts of sections 3.5.H and 4.5.H now on page 159 to page 158 without any change in the wording. This reformating will improve.clarit/ and has no safety significance.
Pgs.159 and 169 - As supported by the reload submittal, the power spiking pe* -Ity is being removed from the linear heat generation rate (LHGR). limits for the 8 x 8, 8 x 8R and P8 x 8R fuel assemblies. This same change was previously made for BF-2 and BF-3 by Amendment No. 67 to Facility License No. DPR-52 on June 12, 1981 and by Amendment No. 37 to Facility License No. DPR-68 on January 12, 1981. The proposed change is supported by the reload submittal and is acceptable.
Pgs 160 and 172b - As supported by the reload submittal, the operating limit MCPR's are being changed. Since the MCPR's were determined by the ODYN Code (rather than the REDY Code), OLMCPR's are now calculated from two curves rathar than being a single value (or a ramp change with fuel ex posu re).
Pg.160a - Whenever the reactor power is equal to or greater than 25%
thermal power, section 4.1.8 of the Technical Specifications requires that the ratio of Fraction of Rated Power (FRP) to Core Maximum Fraction of Limiting Power Density (CMFLPD) shall be checked daily and the APRM scram trip setpoint(s) and the rod block trip setpoint (SRB) recalculated and adjusted if the ratio is lest tnan one (1). Unlike the BWR Standard Technical Specifications (NUREG-0123, Rev. 3), the Browns Ferry Technical Specifications do not provide a specified time to initiate corrective action or a time period to adjust the setpoints. Also, any excursion above this limit is now subject to the reporting requirements of Section 6.7.2.b(2). Under the old MCHFR correlations, the peaking factor (MFLPD/
FRP) adjus: ment to the flow biased scram and rod block equations had relevance to maintaining core limits in certain flow excursion transients.
Since adoption of CPR correlations, this is no longer the case and the flow biased equations now serve as a backup to the fixed (120%) scram and the RBM system, and provides additional conservatism for transients.
Credit is-not taken for-the flow biased trips.in the Browns Ferry transient aralyses. Therefore, there is sufficient justification for relaxing the corrective action and time allowances in comparison to the standard core limits (MCPR, LHGR, etc.). Section 3.5.L is being modified to incorporate language similar to the BWR Standard Technical Specifications on the time permitted to initiate corrective action and to bring the factor within limits.
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. Pgs.171,172 and 172a - These revised pages present the new MAPLHGR versus average planar exposure limits determined by the supplemental reload analysis.
3.5 Plant Modifications r
During this refueling outage, 67 significant modifications are being performed in addition to refueling, inservice inspection, surveillance and calibration tests, equipment overhaul and other maintenance performed during a refueling outage. These modifications are described in TVA's letter to us of May 22, 1981 and in the montaly operating reports. The most significant of the modifications are the torus integrity modifi-cations being performed as part of the Mark I Containment Program.
Another major modification is the changes being made to the BF-1 and BF-2 electrical systems; these electrical modifications are described and evaluated in a separate amendmert.
3.5.1 Hydrogen Monitoring System One of the modifications being performed, which requires changes to the Technical Specifications, is replacement of the containment hydrogen-oxygen monitoring system. This is the same type of monitoring system installed at the last refueling outa es in Units 2 and 3.
A complete description and evaluation of the new monitoring system is included in Amendment No. 58 to Facility Operating License No. OPR-52 issued November 12. 1980 and in Amendment No. 37 to Facility Operating License No. DPR-68 issued January 12, 1981 for Browns Ferry Unit Nos. 2 and 3, respectively. The evaluations contained therein are incorporated herein by reference. We conclude that the monitoring system meets the require-monts in NUREG-0737 (" Clarification of TMI Action Plan Requirements")
and that the proposed changes to the Technical Specifications are accepta ble.
3.5.2 Torus Modificationi Numerous modifications are being implemented in the Unit I torus during the current refueling outage as part of the Mark I Containment Program.
These modifications are required by NRC to restore the originally intended margins of safety in the containment design. The structural modifications to the torus contair. ment include addition af torus tiedowns, addition-of ring girder reinforcement and reinforcing attached piping nozzles.
Vent System modifications include shortening the downcomers, adding local reinforcement to the vent header and adding new tie bars to the downcomers. Attached piping is being strengthened including modification of the ECCS header support. Many changes are being made to the safety relief valve piping system including adding quencher arms to the ramshead, adding quencher arm and ramshead suoports, adding ten-inch vacuum valves, reinforcing the ring girder at the SRV hanger attachment, rerouting of piping and adding new snubbers and supports for the piping. These modifi-cations have taken much longer to implement than originally estimated and have considerably extended the Unit 1 outage. When Unit 1 shutdown on April ll,1981, the scheduled restart date was July 23, 1981.
The projected startup date has slipped to about mid-September 1981 - almost two months longer than estimated.
. TM modifications to the torus and piping systems requires some changes to the Technical Specifications, as discussed Below:
Pgs. 227 and 267 - 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" (differential pressure control greater than 0 psid) to -6.25" and from -8" (0 psid differential pressure control) to -7.25" - a change in each case of 0.75".
There 'ar? 15-inch 5y 15-inch sealed box 6eams being added as support for the safet/ 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 tor 9s water level is reflected in these revised technical specification values.
Pgs. 235a and 269 - In Section 3.7.A.6.a (and the bases therefore), the setpoint for the drywell-suppression chamber (wetwell) differential pressure control (aP) is being changed from 1.3 psid to 1,1 psid.
Downcomer water clear 1ng loads are the downcomers (by almost one foot) greatly reduced by physically shortening 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 level.
In order to be consistent with this analysis the technical specification associated with the AP control has been established at 1.10 psid.
Pg. 268 - In the bases for the limits established fer primary containment, there is a discussion of steam condensing loads associated with relief valve operation. The peak temperature of the torus water used in the evaluation is being changed from 160*F to 200*F local temperature.
During the current refuel outage the T-quenchers are being tdded to the safety-relief valve discharge device. The NRC licensed value for the T-quencher is 200*F local water temperature (to avoid excessive steam condensingloads). This technical specification change is needed to reflect that T-quencher licensed value of temperature.
3.5.3 Containment Purca System In response to our generic letters of September 27, 1979 and Cctober 22, 1979 to "All Light Water Reactors," TVA is modifying the containment purge system for Unit 1 during this outage to satisfy applicabl: require-ments of NRC Branch Technical Position CSB 6-4 regarding valve closure times and addition of debris screens. Table 3.7.A (pages 251 and 252) is being revised to reflect the significant reduction in the maximum allowable operating time for the purge valves. 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
tu-N4 1
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.
In their letter of June 2,1981, TVA provided the data and analysis to demonstrate that the purge valves-are adequate for closure against the
- design. basis loss-of-coolant accident forces. We have concluded that the plant modifications and changes to the Technical Specifications are significan_t improvements in plant safety and should be approved.
3.5.4 HPCI Bypass Valve Ouring-this refueling outage, a one-inch bypass valve is being added around the HPCI steam supply outboard isolatic., valve, FCV73-3. During quarterly surveillance testing on HPCI isolation valve FCY 73-3, in which the valve is closed' and reopened, the 'steamline downstream from FCV 73-3 is subjected to thermal stresses from the. closure and subsequent reopening.
Additim of FCV 73-81 will relieve those stresses..This is a one-inch valve.
It is an isolation group 4 valve with a maximum closing time of 10 seconds. Since this is an isolation valve, !c is being added to the list _ of valves in Tables 3.7.A (p. 251) and 3.7.0 (p. 260).
4.0 Environmental' Considerations We have 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 standp9t ::f anvironmental impact, and pursuant to 10 CFR Section 51.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance'of the amendnent.
5.0
-Concluston We have concluced,' based on the considerations discussed above, that:
.(.1) because the amendment does not involve a significant increase.in the probability or consequences of accidents previously considered and
.does not involve a significant decrease 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 endangered by operation in the proposed manner, and (3) such activities will -be, conducted in compliance with the Commission's regulations and the. issuance of this ' amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: September 15, 1981 f -y
%, t.
y References -
-l.
-Supplemental Reload Licensing)Sub:nittal for Browns Ferry Nuclear Plant
- Unit 1,; Reload No._4 (Cycle 5 Y1003J01 A19 dated March 1981.
2.-
- Errata and Addenda Sheet No. 2 dated April 1981 to NEDO-24056, " Loss-of-Coolant-Accident Analysis for Browns Ferry Nuclear Plant Unit 1 issued September 1977" 3.
" General Electric Boiling Water Reactor Generic Reload Application,"
H NEDE-240ll-P-A, May 1977.
- 4 ~.
-Letter, R. E. Engel (GE) to V. S. Nuclear Regulatory Commission, dated i -
January 30, 1979.-
5.
Letter, T. A.-Ippolito (USNRC) to R. Gridley'(GE), April 16, 1979, and enclosed SER.
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