ML18025B936
| ML18025B936 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/14/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18025B935 | List: |
| References | |
| NUDOCS 8303300234 | |
| Download: ML18025B936 (8) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 1
EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. DPR-33 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT NO.
1 DOCKET NO. 50-259
, 1. 0 Introducti on By letter dated February 1, 1983, (TVA BFNP TS 184), the Tennessee Valley Authority (the licensee) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License No.
DPR-33 for the Browns Ferry Nuclear Plant, Unit 1.
The proposed Technical Specifications
'ould allow operation of Browns Ferry Unit 1 (BF-1) with increased core flow during the 'remainder of Cycle 5.
In support of this application, the licensee submitted a safety evaluation performed by the General Electric Company (GE),
(NED0-22135, "Safety Review of Browns Ferry Nuclear Plant Unit No.
1 at Core Flow Conditions Above Rated Flow During Cycle 5").
2.0 Disucssion BF-1 is presently in a coastdown mode of operation and is scheduled to be shutdown about mid April 1983 for a four to five-month refueling and maintenance outage.
At present, the maximum attainable power is 992 MWe,. about 93K of the normal electrical output, operating at 100Ã,of rated flow.
The proposed changes to the Technical Specifications are to permit BF-1 to operate with core flows up to 105% of rated f)ow for the rest of the fuel cycle.
The increased core flow would permit the unit to generate about 3Ã more power than would otherwise be attainable during the current coast-down mode of operation.
This amendment does not authorize BF-1 to exceed the thermal power limit authorized by License No.
BF-1 is operating in a coastdown mode because of the delayed restart of Browns Ferry Unit 2 (BF-2).
BF-2 shutdown on July 30, 1982 for refueling'and major modifications (e.g., the Mark I torus modifications).
'BF-2 was originally scheduled to return to service by mid January 1983 an) BF-1 was scheduled to shutdown about March 1, 1983.
The projected startup date for BF-2 is now about mid March 1983.
To avoid having two units down at the same time, the shutdown date for BF-1 has been postponed until mid-April 1983.
3.0 Evaluation
- 3. 1 Thermal and Hydraulic Desi n
The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable
- methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, and is not susceptible to thermal-hydraulic instability.
8303300234 830314 PDR ADQCK 05000259 P
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The review includes the following areas:
(1) safety limit minimum critical power ratio (MCPR), (2) operating limit MCPR, (3) thermal-hydraulic stability, and (4) changes to Figures 3.5. K-1 and 3.5.2 of the Technical Specifications.
The licensee has submitted the analysis report for Cycle 5 operation at core flow conditions above rated flow (Ref. 2).
This report relies on a generic document (Ref. 3), which has been reviewed and approved (Ref. 4) by the staff.
We conclude that additional staff review of this portion of Reference 2
concerning the standard thermal-hydraulic design is not required for Cycle 5
operation at core flow conditions above rated flow since it has been previously reviewed and found acceptable.
Discussion of the review concerning the thermal-hydraulic design for Cycle 5 operation follows:
3.1.1 Safet Limit MCPR As stated in Reference 3, for BWR cores which reload with GE's retrofit 8x8 fuel, the safety limit minimum critical power ratio (SLNCPR) resulting from either core-wide or localized abnormal operational transients is equal to 1.07.
When meeting this SLMCPR during a transient, at least 99.9X of the fuel rods in the core are expected to avoid boiling 'transition.
The 1.07 SLMCPR is unchanged from the SLMCPR previously approved for Cycle 5.
The basis for this safety limit is addressees in Reference 3.
3.1.2 0 eratin Limit MCPR The most limiting events have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (aCPR).
The hCPR values given in Table 2-1 of Reference 2 are plant specific va'Iues calculated by using the ODYN methods.
The calculated aCPRs are adjusted to reflect either Option A or Option B ACPRs by employing the conversion method described in Reference 7.
The MCPR values are determined by adding the adjusted aCPRs to the safety limit MCPR.
Table 6. 1 of Reference 2
presents both the cycle MCPR values for the non-pressurization and pressurization events.
The maximum cycle MCPR values (Options A and B) in Table 6.1 are specified as the operating limit MCPRs and incorporated into the Technical Specifications.
Since the approved method was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated operation transients, we conclude that these limits are acceptable.
- 3. 1.3 Thermal-H draulic Stabilit The results of the thermal-hydraulic analysis (Ref. 2) show that the maximum reactor core stability decay ratio in increased core flow operation during Cycle 5 is bounded by the Reload-4 licen'sing submittal's which have been previously approved (Ref. 6).
Therefore, we conclude that the thermal-hydraulic stability results are acceptable for increased core flow operation during Cycle 5...'
3.1.4 Chan es to Fi ures 3.5.K-l and 3.5.2 of the Technical S ecifications Figure 3,5.$ -1 of the Technical Specifications has been modified to include the operating limit MCPR for Cycle 5 extended flow operation.
Using Option A, the operating limit MCPRs shall be 1.35 for PBXBR fuel, and 1.34 for 8X8 and 8XBR fuel types.
.Using Option B, the operating limit MCPR shall be 1. 27, 1.26 and 1.25 for P8XBR, 8X8 and BXBR fuel types respectively.
Figure 3.5.2 has been changed to include a note to reflect that the Kf factor is equal to rated core flow.
3.1.5 Fuel Bundle Liftoff GE re-evaluated the bundle liftoffmargin for 105 percent core flow.
The method used was described in a letter from R. Gridley (GE) to D. Eisenhut (NRC) dated July 11, 1977.
The new analysis yielded a bundle liftoffmargin of 132 lbs, which is 15 lbs less than the old analysis using 100 percent core flow.
We conclude that this is a small variation and an adequate liftoffmargin is maintained for the increased core flow during Cycle 5 operation.
- 3. 2
~N1 Il The rod block monitor is programmed to block rod withdrawal when its output is 106 percent of full power.
If the program were not changed, at 105 percent flow the block would occur at 109.3 percent of full power.
This would result in a change in CPR of 0.31 for 8XB fuel - an unacceptably high value.
Accordingly the RBM upscale flow biased setpoint is clipped at 106 percent rated power.
The change in CPR would then be 0. 19 for this event for the 8X8 fuel.
This is an acceptable procedure and result.
Table 3.2.C of the Technical 'Specificatioris "
has been modified to show this change.
The rod drop accident is a low flow startup event that is sot affected by the change in flow except for end-of-cycle operation where the initial conditions are slightly altered.
However, end-of-cycle conditions are not limiting for this event and the previous analysis is still valid.
3.3 Summar of Evaluation We find that approved thermal hydraulic methods have been used and the the'esults of analyses support the proposed MCPR limits, which avoid.violation of the safety limit MCPR for design transients.
We conclude that the changes approved by this amendment will not adversely affect the capability to operate BF-1 safely during Cycle 5 extended flow operation and that the proposed changes to Figures 3.5. K-1 and 3.5.2 of the Technical Specifications discussed above are acceptable..'ased on the discussion in Section 3 above we conclude that clipping the Rod Block Monitor at 106 percent of rated power will permit the plant to be operated within the limits shown on Figure 3.5.K. l.
'In summary we conclude that operating during the remainder of Cycle 5 with extended flow will not endanger the health and safety of.the public.
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4.0 Environmental Considerations He have determined that the amendment does not authorize a
Change
$ n 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 insignificart from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environnental impact appraisal need not be prepared in connection. with the issuance of this amendment.
- 5. 0 Concl us ion Me 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 margin of safety, 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 Comaission'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: March 14, 1983 Principal Contributors:
W. Brooks, S. Sun, S.
Wu
References:
j 1.
Letter from L. Miller (TVA) and attachments to H. Denton (NRC) dated February 1, 1983.
2.
NED0-22135, "Safety Review of Browns Ferry Nuclear Plant Unit No.
1 at Core Flow Conditions above Rated Flow During Cycle 5," dated May 1982.
3.
NED0-24011-A-4, "General Electric Boiling Mater Reactor Generic Reload Fuel Applications," January 1982.
4.
Letter from D.
G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978.
5.
Y1003JOlA19, Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1, Reload No. 4 (Cycle 5) dated March 1981.
6.
Letter from T. Ippolito (NRC) to H. Parris (TVA) dated September 15, 1981.
7.
Letter from R. Buchholz (GE) to P.
Check (NRC), Response to NRC Request for Information on ODYN Computer Model, September 5, 1980.
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