ML19269C556
| ML19269C556 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/17/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19269C555 | List: |
| References | |
| NUDOCS 7902060018 | |
| Download: ML19269C556 (10) | |
Text
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4r UNITED STATES NUCLEAR REGULATORY COMMisslON g V.
- t WASHINGTON, D. C. 20555 I
5 y
- .o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 47 TO FACILITY LICENSE NO. OPR-33 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-259 1.0 Introduction By letter dated September 8, 1978 (TVA BFNP TSil5), as supplemented by letters dated October 5, 1978, November 30, 1978, December 5, 1978, December 14, 1978, January 8,1979 and January 9,1979, the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant, Unit No.1.
The proposed amendment and revised Technical Specifications would incorporate the limiting conditions for operation of the facility in the third fuel cycle following the second refueling of the reactor.
In support of this reload application for Browns Ferry gt No.1 (BF-1), the licensee has submitted a reload licensing document prepared by the General Electric Company (GE), a supplemental reload licensing document (5) also prepared by GE and proposed changes to the Technical Specifications (1,2,5,22).
2.0 Discussion Browns Ferry Unit No.1 (BF-1) shutdown for its sec'd refueling on November 26, 1978. During the refueling, 156 irradiated 7x7 fuel assemblies were replaced with a like number of new, two water rod, retrofit 8x8 (8x8R) fuel assemblies designed and fabricated by the General Electric Company (GE). During initial operation in fuel cycle 2 (January to November 1978), an increase in fission product activity was noted in the off-gas.
During the outage, all of the irradiated fuel was
" sipped" to check for possible leakage of fission products through the cladding. As a result of this operation, it was found that two of '.he 168 8x8 fuel assemblies that had been installed during the previous refuel-ing evidenced a slight amount of leakage and were replaced with two 7x7 fuel assemblies irradiated during the initial fuel cycle.
The development of minor leakage in two 8x8 fuel assemblies is not considered significant.
The fact that all of the fuel was inspected (sipped) provided confidence that the 8x8 fuel is acceptable for use in the forthcoming fuel cycle.
79020600\\%
This reload (Reload 2) is the first for BF-1 to incorporate GE's 8x8R fuel design on a batch basis. The description of the nuclear and mech-anical design of the Reload 2 8x8R fuel and the exposed fuel designs used for initial core and Reload 1 is contained in GE's generic licen-sing topical report for BWR reloads.(6)
Reference 6 also contains a complete set of references to GE's topical reports which describe GE's BWR reload analysis methods for the nuclear, mechanical, themal-hydraulic, transient and accident calculations, together wi th infomation on the applicability of these methods to cores containing a mixture of different fuel designs.
Portions of the plant-specific data, such as operating con-ditions and design parameters which are used in transient and accident calculations, have also been included in the topical report.
Our safety evaluation (l) of GE's generic reload licensing topical report concluded that the nuclear and mechanical design of the 8x8R fuel and GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, as applied to cores containing mixtures of 7x7, 8x8, and 8x8R fuel, are acceptable. Our acceptance of the nuclear and mechanical design of the standard 8x8 fuel was expressed in the staff's evaluation (8) of the infomation in Reference 9.
As part of our evaluation (7) of Reference 6 we found the cycle-independent input data for the reload transient and accident analyses for BF-1 to be acceptable. The supplementary cycle-dependent infor-mation and input data are provided in Reference 5, which follows the format and content of Appendix A of Reference 6.
As a result of the staff's generic evaluation (7) of a substantial number of safety considerations related to use of 8x8R fuel in mixed core loadings with 8x8 and 7x7 fuel, only a limited number of addi-tional review items are included in this evaluation.
These include the plant and cycle-specific input data and results presented in Ref-erence 5, the LOCA-ECCS analysis results for the reload fuel design, eference 7 as requiring special atten-and those items identified ir o tion during reload reviews.
3.0 Evaluation 3.1 Nuclear Characteristics For Cycle 3,156 fresh 8x8R fuel bundles, with a bundle average en-richment of 2.65 wt/% U-235 will be loaded into the core, replacing a like number of exposed 7x7 assemblies.
The remainder of the 764 fuel assembly reload core will consist of the irradiated 7x7 and 8x8 fuel assemblies exposed during the first two fuel cycles. The reference core loading for Cycle 3 will result in eighth core symmetry, which is consistent with previous cycles.
. The information provided in Section 6 of Reference 5 indicates that the fuel temperature and void dependent behavior of the reconstituted core is not significantly different from that of previous cycles.
Additionally, scram effectiveness, Figure 2 of Reference 5, is also similar to earlier cycles. The 1.7%Ak/k calculated shutdown margin for the reconstituted core meets the Technical Specificatien require-ment that the core be subtritical by at least 1 3 % V k :n the most reactive operating state with the single most re.tM control rod fully withdrawn and all other rods fullv insarted.
Finally, Pefer-ence 5 indicates that a bynn concentratm af 600 ppm in the Jioderator has been calculated to make th? reactor subcritical by at least 3.0%Ak at 20 C, and xenon fre'e conditions. Therefore, the alternate shutdown requireraent of the General Design Criteria can be achieved by the St.indt Liquid Control System. We have reviewed these analyses and conipared them to the Technical Specification requirements and find them acceptable.
3.2 Thermal-Hydraulics-3.2.1 Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 7, for BWR cores which reload with GE's retrofit 8x8R fuel, the allowable minimum critical power ratio (MCPR), resulting f rom either core-wide or localized abnormal operational transients, is equal to 1.07.
With this MCPR safety limit, at least 99.97, of the fuel rods in the core are expected to avoid Doiling transition.
The 1.07 safety limit minimum critical power ratio (SLMCPR) proposed by the licensee for Cycle 3 represents a.01 increase from the 1.06 SLMCPR applicable during Cycle 2.
The basis for the revised safety limit is addressed in Reference 6, while our generic apmval of the new limit is given in Reference 7.
This change continu i to meet the reconnenda-tion of Standard Review Plan 4.4 and on that oasis has been found acceptable in Reference 7.
3.2.2 Operating Limit MCPR Various transient events will reduce the MCPR from its normal operating value. To assure..iat the fuel cladding integrity safety limit MCPR will not be violated during any abnormal operational transient, the most limiting transients hTve been reanalyzed by the licensee to determine which event results in the largest reduction in critical power ratio.
Each of the events has been conservatively analyzed for each of the several fuel types (i.e., 7x7, 8x8, 8x8R) and for the full range of exposure through the cycle.
N. In the transient analyses of Reference 7, credit was taken for an end-of-cycle (E0C) recirculation pump trip (RPT). ~(The E0C RPT is different from and should not be confused with the ATWS RPT). We have reviewed the usign of the E0C RPT and conclude that it is unacceptable for reasons as ; ven in our January 16 letter (Reference 23).
Since there is no avail-able analysis which is specific to this core, we require a conservative bound on operating limit MCPR. The previous cycle transieat analyses (Reference 24) were evaluated from this standpoint. The input parameters for that cycle analyses are conservative when compared to this cycle input paramters at the EOC. This includes comparisons of void reactivity coefficient, scram reactivity insertion, and Doppler reactivity coefficient which are the key parameters for core-wide transient behavior. The key parameters for CPR evaluations, which are also conservative for last cycle's analysis, are power peaking factor, bundle flow rate and initial CPR. With these conservative input para-meters, the transient results for last cycle are bounding for this cycle at E0C.
Therefore, we have proposed and the licensee has agreed to operating limit MCPRs of 1.34 for 7x7 fuel and 1.43 for 8x8 and 8x8R fuel. These were derived from a safety limit MCPR of 1.07 and ACPR of G.27 for 7x7 fuel and 0.36 for 8x8 fuel. This assures that an abnormal operational transient will~ result in a CPR no lower than the 1.07 safety limit which we find acceptable as discussed in the previous section.
3.3 Accident Analysis 3.3.1 ECCS Appendix K Analysis The licensee has reevaluated the adequacy of ECCS performance in connec-tion with the new reload fuel design, using methods previously approved by the staff. The results of these plant-specific analyses are given in Reference 5.
We have reviewed the information submitted by the licensee and conclude that all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 will be met when the reactor is operated in accordance with the MAPLHGR varsus Average Planar Exposure values given in Section 6 of Reference 5 and which have bten incorporated in the revised Technical Specifications.
Continuous operation with four of six automatic depressurization system ( ADS) valves operable (instead of the previous five out of six requirement) has been found acceptable in References 16 and 17.
We have reviewed this acceptance and its applicability to BF-1.
On th: bases of our review and the conclusions reached in References 1
id 17, we find the proposed change from five to four to be ac-
.able.
. 3.3.2 Control Rod Drop Accident Because the characteristic accident analysis input parameters for the worst case CRDA did not satisfy all of the assurr.ptions of the bounding analysis, the licensee reanalyzed this event on a plant-specific basis. The results showed the peak fuel enthalpy to be less than the 280 cal /gm limit which is acceptable.
3.3.3 Failure of Trip Inputs from Turbine Building to Reactor Protection System During our review of the reactor protection system, we noted that the trip inputs for the recirculation pump trip and reactor scram following load rejection or turbine trip originate in the turbine building. The turbine building, as is the case of most boiling water reactor plants, is not seismically qualified, henca, its integrity and functions cannot be assured in the event of an earthquake.
For these reasons, the licensee was requested to analyze the conse-quences of a safe shutdown earthquake concurrent with the limiting transient event without taking credit for reactor scram or recircu-lation p:.mp trip from the turbine building inputs.
Br owns Ferry Unit 1 has referenced a Hatch Unit 2 analysis.
We have compared the significant parameters for these two plants (bundle power level and critical power ratio change) and have concluded that the Hatch 2 analysis conservatively bounds the Browns Ferry Unit 1 conditions. We agree with the licensee previous staff findings on this analysis, gry and on the basis of that this analysis is applicable to Browns we find the results acceptable.
3.3.4 Fuel Loading Error The licensee has also considered the effect of a possible fuel loading error on bundle CPR. An analysis of the most severe misoriented fuel loading error usin GE's new methodology,(13,14) which as modified, has been approved ( 5) by the staff, shows that the worst possible rotation of a fuel bundle will not cause a violation of the 1.07 safety limit MCPR.
Additionally, an analysis of the most severe mislocated fuel bundle with GE's new, approved methodology shows that the worst potential mislocation will not violate the MCPR safety limit. We find the results of these analysis acceptable.
. 3.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, ha-been perfomed in accordance with the requirements of Reference 7 As specified in Reference 20, the sensitivity of peak vessel i ' essure to failure of one safety valve has also been evaluated.
We agree that there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one valve. Therefore, the limiting overpressure event as analyzed by the licensee is considered acceptable.
3.5 Thermal-Hydraulic Stability A themal-hydraulic stability analysis was performed with the methods described in Reference 6.
The results show that the channel hydro-dynamic and reactor core decay ratios at the least stable operating state (corresponding to the intersection of the natural circulation curve and 105% rod line on the power-flow map) are below the 1.0 Ultimate Perfomance Limit decay ratio proposed by GE.
The staff Las expressed generic concerns regarding reactor core themal-hy Iraulic stability at the least stable reactor condition.
This condition could be reached during an operational transient from high power if the plant were to sustain a trip of both recircu-lation pumps without a reactor trip. The concerns are motivated by increasing decay ratios as equilibrium fuel cycles are approached and as reload fuel designs change.
The staff concerns relate to both the consequences of operating at a decay ratio of 1.0 and the capability of the analytical methods to accurately predict decay ratios.
The General Electric Company is aao,. sing these staff concerns through meetings, topical reports and a stability test program.
Although a final test report has not as yet been received by the staff for review, it is expected that the test results will aid considerably in resolving the staff concerns.
For the previous operating cycle, the staff, as an interim measure, added a requirement to the Technical Specifications which restricted planned operation in the natural circulation mode.
Continuation of this restriction will also provide a significant increase in the reactor core stability operating margins for the current cycle so that the decay ratio is <1.0 in all operating modes. On the basis of the foregoing, the staff considers the plant themal-hydraulic stability characteristics to be acceptable.
4 4.0 Physics Startup Testing The licensee will perfonn a series of physics startup tests ano procedures to provide assurance that the conditions assumed for the transient and accident analysis calculations will be met during Cycle 2.
The tests will check that the core is loaded as intended, that the incore monitoring system is functioning as expected, and that the process computer has been reprogrammed to properly reflect changes associated with the reload.
The test program is consistent with that previously found acceptable for Browns Ferry Unit 3.(11)
We find this test program to be acceptable.
5.0 Technical Specification Change The proposed Technical Spec cation changes include a revised fuel cladding integrity safety 1 Nt MCPR, a revised operating limit minimum critical power rat, (MCPR) for each fuel type, addition of a MAPLHGR vs average planar exposure table and addition of a design maximum total peaking factor for the reload 8x8R fuel assemblies.
The revised 1.07 safety limit MCPR results in a.01 increase from the 1.06 safety limit MCPR (SLMCPR) of the previous cycle.
Based on our generic review.(7) we find the use of r 1.07 SLMCPR to be acceptable (Section 3.2.1, herein). Also, bas d on the discussions appearing in Section 3.2.2 herein, the staff finds the proposed oper-ating limit MCPRs, as modified to reflect analysis uncertainties, to be acceptable. We find that the proposed MAPLHGR vs average planar exposure table is adequate to assure conformance with the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 for the reload 8x8R fuel assemblies per Section 3.3.1, herein.
The proposed flow biased APRN upscale rod block has been revised. The revision reduces the setpoint for this rod block by 1% of rated power.
This reduction will result in a less severe rod withdrawal error, because the transient will be terminated earlier.
The rod withdrawal error analysis utilized this revision. Our evaluation of these results showed that the safety limit MCPR criteria was met and therefore, the revision is acceptable.
The Technical Specifications have been modified to adjust the number of operable ADS valves based on the findings, as discussed in Section 3.3.1, herein.
A calculational constant (Total Core Length) for the 8x8R LHGR evaluation has been added to the Technical Specifications. This has been previously found acceptable in Reference 5.
Since the fuel for this reload is identical to that of the Reference 5 evalu-ation, we find this addition acceptable.
_8-Finally. the Technical Specifications, which are associated with safety / relief valve number and operability, are being revised. The revisions allow replacement of two safety valves with two safety /
relief valves which will be aligned identically to the present safety / relief valves.
Section 3.4, herein, has found that acceptable overpressuritation protection is provided by these specifications.
Therefore, the modification is acceptable.
Environmental Considerations The revised operating limit minimum critical power ratios (0LMCPR) discussed in Section 3.2.2 may result in a restriction on the attain-able power generation.
The reduction in rated power level is estimated to be minimal (a few percent) for the first part of the fuel cycle.
It is expected that the present OLMCPRs will be revised to be less restric-tive when satisfactory documentation is received relating to the testing of the E0C RPT.
Thus, the reduction in rated power level, if any, will be for a limited period of time. The small reduction in power from one unit of the BFNP for a limited period does not affect the environ-mental evaluation contained in the Final Environmental Statement (FES) related to operation of the Browns Ferry Nuclear Plant, Units 1, 2 and 3, issued September 1,1972, There will be no significant change in the other environmental impacts identified in the FES. This amendment does not authorize a change in effluent types or total amounts. We conclude that this amendment will not result in any significant environmental impact.
Having made this determination, we have further concluded that this amendment involves an action which is insignificant from the standpoint of environmental 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 this amendment.
Conclusion We have concluded 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 comon defense and security or to the health and safety of the public.
Dated: January 17, 1979
References:
Tennessee Valley Authority (TVA) letter (Gilleland) to USNRC (Denton) 1.
dated September 8, 1978.
2.
TVA letter (Gilleland) to USNRC (Denton) dated October 5,1978.
3.
TVA letter (Gilleland) to USNRC (Denton) dated November 30, 1978.
TVA letter (Gilleland) to USNRC (Denton) dated December 5, 1978.
4.
" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear 5.
Plant Unit 1 Reload 2," NED0-24136, Rev.1, November 1978.
6.
" Generic Reload Fuel Application," General Electric Report, NEDE-24011-P-3, dated March 1978.
USNRC letter (Eisenhut) to General Electric (Gridley) dated May 12, 7.
1978, transmitting " Safety Evaluation for the General Electric Topical Report, ' Generic Reload Fuel Application,' (NEDE-24011-P)."
" Status Report on the Licensing Topical Report, General Electric 8.
Boiling Water Reactor Gencric Reload Application for 8x8 Fuel,"
NED0-20360, Revision 1 and Supplement 1 by the Division of Technical Review, Office of Nuclear Reactor Regulation, USNRC, April 1975.
" General Electric Boiling Water Reactor Generic Reload Application 9.
for 8x8 Fuel," NED0-20360 Revision 1, Supplement 4, April 1,1976.
- 10. Memo from P. S. C" ck (RSB-NRC) to T. A. Ippolito (0RB#3-NRC),
" Browns Ferry 3
,ycle 2 Reload' (TACS #8026)," November 8,1978.
- 11. NRC letter (Ippolito) to TVA (Hughes), Amendment Nos. 45, 41, and 18 to Facility License No. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant Units Nos.1, 2, and 3, dated November 18, 1978.
- 12. GE letter (Fuller) to NRC (Ross) dated January 13, 1978.
GE letter (Engle) to NRC (Eisenhut), " Fuel Assembly Loading Error" 13.
dated June 1,1977.
- 14. GE letter (Engle) to NRC (Eisenhut) dated November 30, 1977.
- 15. NRC letter (Eisenhut) to GE (Engle) dated May 8, 1978.
- 16. NRC Safety Evaluation Report on Operation of Browns Ferry Units 1 and 2 with Four of the Six ADS Valves Operable, May 7,1978.
. 17. NRC letter (Ippolito) to TVA (Hughes), Amendment No. 35 to Facility License No. DPR-52 for Browns Ferry Nuclear Plant Unit No. 2 Cycle 2, dated June 21, 1978.
- 18. Edwin I. Hatch Nuclear Plant, Unit 2, FSAR Question 212.64 (15.1.1),
(15.2.2).
19.
Safety Evaluation Report related to operation of Edwin 1. Hatch Nuclear Plant, Unit No. 2, Georgia Power Company, et. al., USNRC, Office of Nuclear Reactor Regulation, Docket No. 50-366, NUREG-0411, June 1978.
- 20. Letter, E. D. Fuller (GE) to USNRC (Ross), " Impact of One-Dimensional Transient Model on Plant Operations Limits," June 26, 1978.
- 21. Carmichael, L.
A., and Niemi, R. O., " Transient and Stability Tests at Peach Bottom Atomic Power Station Unit 2 at End of Cycle 2,"
EPRI-NP-564, June 1978.
Letter, T. A. Ippolito (NRC) to H. G. Parris (TVA), January 16, 1979.
- 24. General Electric Boiling Water Reactor Reload-l Licensing Amendment for Browns Ferry Nuclear Plant Unit 1, NE00-24020, May 1977.