ML20248C884
| ML20248C884 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/13/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248C873 | List: |
| References | |
| NUDOCS 8910040051 | |
| Download: ML20248C884 (4) | |
Text
. - _ _ - _ _ _ - _ _
?[>R"*%.
o UNITED STATES NUCLEAR REGULATORY COMMISSION n
{
E, WASHINGTON, D. C. 20555 R....+f j.
ENCLOSURE 2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.172 TO FACILITY OPERATING LICENSE NO. DPR-52 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-260
1.0 INTRODUCTION
By letter dated August 26,1988 (Reference 1), the Tennessee Valley Authority (the licensee or TVA) requested an amendment to Facility Operating License No.
DPR-52 for the Browns Ferry Nuclear Plant Unit 2 (BFN2).
The proposed amend-ment would change the Technical Specifications (TS) of the operating license to modify the operating thermal limits to be consistent with the reanalysis associated with' Cycle 6 operation.
The staff review included those aspects of the reload related to the BFN Fuel Inspection and Reconstitution Program. A suninary report on this program was submitted by the licensee by Reference 2.
2.0 EVALUATION The original application for Technical Specification changes for operation of BFN Unit 2 in Cycle 6 was submitted by Reference 3 and was reviewed by the NRC in connection with Amendment 125 to Facility Operating License No. DPR-52 in 1986 (Reference 4).
The present proposal reflects Cycle 6 fuel loading changes made as a result of the fuel inspection and reconstitution program which was completed in July 1988.
In support of the application TVA submitted Revi-sion 2 to TVA-RLR-002 (dated July 1988) which is an update of the current licensed design reviewed by the staff for Amendment 125.
The revision was pret ired with consideration of the reconstituted fuel and reanalysis by TVA witn input from General Electric Company and reported in Reference 2.
Reference 2 was reviewed by the staff to the extent necessary to confirm that the modeling assumptions used in the reload analyses are valid for the recon-stituted core. The inspection and reconstitution process was necessary because of fuel reliability problems as a result of a corrosion mechanism which can cause fuel rod cladding degradation. The objective of the program was to provide a sufficient number of reload fuel assemblies to ensure reliable operation of the BFN2 core within its licensing basis.
Those considerations relative to the proposed Amendment included core nuclear design characteris-tics, the transient and accident safety analysis results, and the proposed operating thermal limits.
$$k00$$0bk b 0
P
l 2
The reconstitution process for assemblies designated for Cycle 6 reload involved exchanging fuel rods found unacceptable by visual observation with rods meeting acceptance criteria as established in the inspection plan.
The designated fuel assemblies were twice-and thrice-burned bundles from prior operation of BFN Plant Unit 2.
A total of 212 reconstituted assemblies will be used in Cycle 6.
Those aspects of the fuel reconstitution relative to the Amendment request review are addressed in the following Safety Evaluation (SE) sections.
2.1 Reload Description For Cycle 6, 304 1rradiated fuel assemblies will be removed from the reactor core and replaced by 300 new General Electric pressurized P8x8R assemblies and four Westinghouse QUAD + Demonstration assemblies. These new assemblies were previously reviewed and found acceptable in License Amendment 125 (Reference 4).
The safety analyses for Cycle 6 were redone with the 212 reconstituted assem-
~
blies modeled as origindi assemblies. This modeling assumption was verified in support of the fuel reconstitution project using a methodology previously reviewed and approved by the NRC (Reference 5).
The fuel (P8x8R) to be inserted into the core for Cycle 6 is similar to that customarily used for BWR reloads.
This fuel and the four QUAD + Demonstration assemblies were previously found acceptable in License Amendment 125.
The fuel reconstitution effort does not affect this conclusion.
2.2 Nuclear Design The nuclear design and analysis for the Cycle 6 reload was performed with methods 'and techniques previously reviewed and approved by the staff for use in such analyses.
The reanalyses with consideration of the fuel reconstitution effort were reported in TVA-RLR-002, Revision 2 (enclosed with Reference 1).
The shutdown margin is calculated to be 1.0 percent (delta K)/K at the point in the cycle at which it is a minimum.
This value exceeds the Technical Specifi cation requirement of 0.38 percent and is acceptable.
The Standby Liquid Control System provides a shutcown margin of 2.9 percent (delta K)/K with a boron concentration of 600 ppm boron. This is greater than the design criter-ion of 1.8 percent and is acceptable.
The modeling of the reconstituted fuel in the reanalysis had a minimel affect on the margin.
The conclusion of the staff in Amendment 125 is unchanged by the reanalyses.
2.3 Thermal-Hydraulic Design The thermal-hydraulic reanalysis of the BFN Plant Unit 2 Cycle 6 reload did not result in significant differences in results previously reported in Amendment 125. The staff conclusions are therefore unchanged.
The analyses of core-wide pressurization transient, non-pressurization events and the loss-of-coolant accident did not require any changes in the transient models since the thermal, mechanical, and hydraulic characteristics of the reconstituted assemblies are equivalent to those used in the previous analyses.
i
3 L'
+
3 The Operating Limit Minimum Critical Power Ratio-(MCPR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) analyses previously accepted in-Amendment 125 remain acceptable.
2.4 Thermal-Hydraulic Stability The' li:ensee's submittal included the results of analyses using a methodology which predicted core and channel stability by way of a calculated decay ratio.
Based on:recent staff consideration of power oscillations in boiling water reactors (NRC Bulletin 88-07, Reference 6), the staff has taken the position, in part, that:past licensing calculations are not a reliable indicator that a core will be stable under all operating conditions during a fuel cycle and instrumentation for detection and suppression of neutron flux oscillations and l
recording instrumentation for evaluation of limit cycle flux oscillations may not be adequate. This raises a question of-compliance of Browns Ferry Plants with General Design Criterion 12, " Suppression of. Reactor Power Oscillations,"
The licensee has responded to NRC Bulletin 88-07 in Reference 7.
The staff evaluation of the licensee's response will be provided in a separate report prior to restart.
2.5 Technical Specification Chanaes The Technical Specification (TS) changes proposed by the licensee reflect the new fuel to be loaded in the BFN Plant Unit 2 for Cycle 6 operation.
These changes include core related changes for Linear Heat Generation Rate (LHGR) Limit.
MCPR operating limits and MAPLHGR curves for the new fuel.
Specifically, Tables 3.5.I.1 and 3.5.I.2 contained revised MAPLHGR limits based upon the licensee's acceptable reanalyses as discussed above. Tables ~3.5.I-3 and 3.5.I-4 added new MAPLHGR limits for the two new fuel types resulting from the fuel reconstitution effort.
Figure 3.5.K-1 provides a revised curve for calculating MCPR limits again based upon the licensee's revised analysis.
2.6 Summa ry The licensee has proposed the above described TS changes which reflect the reanalyses required by the Inspection and Reconstitution Program for the Cycle 6 reload.
The changes are acceptable since they are based on analyses using approved methodology and use fuel model:ng assumptions for fuel verified t
to be acceptably reconstituted.
L
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation expo-
- sure, lhe Commission has previously issued a proposed finding that this L_______
(
[~
4 amendment involves no significant hazards consideration and there has been no I
'public comment on such finding. Accordingly, the amendment meets the eligibi-lity criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environ-mental assessment need be prepared in connection with the issuance of the amendment.
4.0' CONCLUSION The Connission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (53 FR 48336) on November 30, 1988 and consulted with the State of Alabama. No public connents were received and the State of Alabama did not have any l.
connents.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense and security nor to the health and safety of the public.
5.0 REFERENCES
1.
Letter, M. J. Ray (TVA) to Document Control Desk (USNRC) dated August 26, 1988 (TVA BFN TS-254).
2.
Letter, R. Gridley (TVA) to Document Control Desk (USNRC) dated October 26,1988, " Summary Report for BFN Plant Unit 2 Cycle 6 Inspection and Reconstitution Program."
3.
Browns Ferry Nuclear Plant Reload Licensing Report, Unit 2, Cycle 6, TVA-RLR-002, July 1984, as supplemented.
4.
Amendment No. 125 to Facility Operating License No. DPR-52 for Browns Ferry Nuclear Plant, Unit 2, August 19, 1986.
5.
TVA-EG-047, "TVA Reload Core Design and Analysis Methodology for the Browns Ferry Nuclear Plant," January 1982.
6.
NRC Bulletin 88-07, " Power Oscillations in Boiling Water Reactors," June 15, 1988 and Supplement 1, dated December 30, 1988.
7.
Letters, R. Gridley (TVA) to Document Control Desk (USNRC) dated November 4, 1988 and March 6, 1989.
Principal Contributors: Michael McCoy Dated:
September 13, 1989
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _