ML20126B512
| ML20126B512 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/25/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126B506 | List: |
| References | |
| NUDOCS 8003130326 | |
| Download: ML20126B512 (9) | |
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UNITED STATES 31 8 l D P[l NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D C. 20555 khs, u /
ms SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 59 TO FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE N0. OPR-52 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS NOS.1 AND 2 DOCKET NOS. 50-259 AND 50-260 1.0 Introduction By letter dated October 4,19790) (TVA BFNP TS 131), as supplemented by letters dated January 15, 1980 and January 29, 1980, the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License Nos. DPR-33 and DPR-52 for the Browns Ferry Nuclear Plant, Unit Nos.1 and 2.
The proposed amendments and revised Technical Specifications would:
o (1) incorporate the limiting conditions for operation of Browns Ferry Unit No.1 in the fourth fuel cycle following the current refueling outage, (2) reflect the changes to the low pressure coolant injection (LPCI) system power supply and elimination of the LPCI loop selection logic as requested in our letter of May 11, 1979 authorizing these modifications and (3) clarify the surveillance requirements in Section 4.5.
2.0 Discussion Browns Ferry Unit No.1 (BF-1) shutdown for its third refueling on January 3,1980.
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 rods 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 current refueling outage, an additional 232 of the 7 x 7 fuel bundles will be replaced with P 8 x 8 fuel assemblies, each l
containing 62 fuel rods. The prepressurized fuel assemblies (P 8 X 8R) are essentially identical from a core physics standpoint to the two water rod fuel assemblies {8 X 3R) 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 BR fuel is discussed 8003130
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in the safety evaluation attached to our letter of April 16, 1979 to General Electric approving the use of this fuel in BWR reload j
licensing applications. The larger inventory of helium gas improves the gap conductance between fuel pellets and cladding 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 expected to yield performance benefits in terms of fuel reliability. The increased prepressurization also results in improved margin to MAPLHGR limits by reducing stored energy, although TVA is not proposing to take any credit for these beneficial effects in the subject reload application (i.e., they are not proposing any changes in the existing MAPLHGR vs. Exposure limits in the existing Technical Specifications).
In support of this reload appligion for BF-1, TVA submitted a reload licensing documengrepared by GE and proposed changes to the Technical Specifications.
The first use of P8 x 8R fuel in a' Browns Ferry Unit'was approved for the last reload of Unit No. 3 (Amendment No. 28 to Facility Operating License No. DPR-58 dated November 30,1979).
With this refueling, Browns Ferry Unit 1 will be on an 18 month refueling cycle. Units Nos. 2 and 3 are also on 18 month refueling cycles.
As noted above, this reload involves loading of prepressurized GE 8 x 8 retrofit (P8 x 8R) fuel. The description of the nuclear and mechanical designs of P8 x 8 fuel is contained in Reference 3.
The use and safety implications of prepressurized fuel are presented in i
Reference 3 and have been found acceptable per Reference 4 (enclosed in Appendix C of Reference 3).
Values for plant-specific data such as steady state operating pressure, core flow, safety and safety / relief valve setpoints, rated thermal power, rated steam flow, and other design parameters are provided in Reference 3.
Additional plant and cycle dependent information is provided in the reload application (Reference 2) which closely follows the outline of Appendix A of Reference 3.
Reference 4 includes a description of the staff's review, approval, and conditions of approval for the plant-specific data. The above-mentioned plant-specific data have been used in the transient and accident analysis provided with the reload application in compliance with Reference 4.
Our safety evaluation of the GE generic reload licensing topical report has also concluded that the nuclear, and mechanical design of the 8 x 8R.and P8 x 8R fuels, and GE's analytical methods for nuclear and thermal-hydraulic calculations as applied to mixed cores 1
containing 8 x 8, 8 x 8R and P8 x 8R fuels, are acceptable. Approval of the application of the analytical methods did not include plants incorporating a prompt recirculation pump trip (RPT) or Thermal Power j
Monitor (TPM).
l
Because of our review of a large number of generic considerations related to use of 8 x 8R and P8 x 8R fuels in mixed loadings, and on the basis of the evaluations which have been presented in Reference 3, only a limited number of additional areas of review have been included in this safety evaluation report. For evaluations of areas not specifically addressed in this safety evaluation report, the reader
)
is referred to Reference 3.
3.0 Evaluation 3.1 Core Reload 3.1.1 Nuclear Characteristics For cycle 4 operation, 232 fresh P8 x 8R fuel bundles of type P80RB284 will be loaded into the core (Reference 2). The remainder of the 764 fuel bundles in the core will be previously irradiated bundles as indicated in Reference 2.
Based on the data provided in Reference 2 both the control rod system and the standby liquid control system will have acceptable shutdown capability during cycle 4.
3.1. 2 Thermal Hydraulics 3.1. 2.1 Fuel Claddino Intecrity Safety Limit MCPR As stated in Reference 3, for BWR cores which reload with GE's retrofit 8 x 8 fuel, 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.
3.1. 2. 2 Operatino Limit MCPR Various transient events can reduce the MCPR from its normal operating level. To assure that the fuel cladding integrity SLMCPR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed for this reload by the licensee, in order to determine which event results in the largest reduction in the minimum critical power ratio. Addition of the largest reductions in critical power ratio to the SLMCPR establishes the operating limits for each fuel type.
3.1. 2. 2.1 Transient Analysis Methods The generic methods used for these calculations, including cycle-independent initial conditions and transient input paraneters, are described in Reference 3.
The staff evaluation, included as Appendix C of Reference 3, contains our acceptance of the cycle-independent values. Additionally, Appendix C contains our evaluation of the transient analysis methods, together with a description and summary of the outstanding issues associated with these methods. Supplementary cycle-independent initial conditions and transient input parameters used in the transient analyses appear in the tables in Sections 6 and 7 of Reference 2.
Our evaluation of the methods used to develop these supplementary input values is also included in Appendix C of Reference 3.
3.1. 2. 2. 2 Transient Analysis Results The transients evaluated were the limiting pressure and power increase transients generator load rejection without bypass and the feedwater controller failure (loss of 100'F feedwater heatir.g), and the control rod withdrawal error.
Initial conditions and transient input parameters as specified in Sections 6 and 7 of Reference 2 were assumed.
The results of these analyses are outlined in Reference 2 sections 9 and 10. On this topic, it is acceptable if fuel specific operating limits are established for prepressurized fuel (Appendix C, Reference 3).
On chis basis, the transient analysis results are acceptable for use in the evaluation of the operating limit MCPR. Based on this, the pro-posed Technical Specification modifications to operating limit MCPR are acceptable.
l 3.1.3 Accident Analyses 3.1. 3.1 ECCS Appendix K analysis In our safety evaluation of Reference 3, we concluded that "the continued application of the present GE ECCS-LOCA (" Appendix K")
models to the 8 x 8 retrofit reload fuel is generically acceptable and in our Reference 4 evaluation we extended that conclusion to prepressurized fuel. On these cases, the proposed MAPLHGR limits for the new prepressurized fuel are acceptable.
3.1.3.2 Control Rod Drop Accident' The scram reactivity shape function (cold) does not satisfy the requirements for the bounding analyses described in Reference 3.
Therefore, it was necessary for the licensee to perform a plant and cycle specific analysis for the control rod drop accident. The results of this analysis are well below the acceptance criterion of 280 calories per gram.
3.1. 3. 3 Fuel Loading Error 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 errors involving misoriented bundles and bundles loaded into incorrect positions have been analyzed by this methodology and the results are acceptable.
3.1.3.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accor-dance with the requirements of Reference 3.
As specified in the staff evaluation included in Reference 3, the sensitivity of peak vessel pressure 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. Therefore, the limiting overpressure event as analyzed by the licensee is considered acceptable.
3.1. 4 Thermal Hydraulic Stability The results of the thernal hydraulic stability analysis (Reference 3) show that the channel hydrodynamic and reactor core decay ratios at the natural circulation - 105% rod line intersection (which is the least stable physically attainable point of operation) are below the stability limit. Because operation in the natural circulation mode will be restricted by Technical Specifications, there will be added margin to the stability limit and this is acceptable.
3.1. 5 Startup Test Program The licensee has not changed his startup test program from that approved for the previous cycle. This program therefore remains acceptable.
3.2 Other Changes to Technical Specifications 3.2.1 Reactor Low Water Level On August 2, 1978, we issued Amendments Nos. 40, 38 and 14 to Facility Licenses Nos. DPR-33, DPR-52 and DPR 68 for the Browns Ferry Nuclear Plant, Units Nos.1, 2 and 3.
These amendments changed the Technical Specifications to lower the reactor low water level setrott from 490 inches to 470 inches above vessel zero. The low water level setpoint, l
which is commonly called the L2 setpoint, is that reactor water level below which the main steamline isolation valves close, HPCI and RCIC flows are initiated, and the recirculation pumps trip. We evaluated the ECCS performance with the L2 setpoint at 470 inches and the effect of reduction in L2 on results of anticipated transients and found that these were acceptable. The Amendments changed 4 pages of the Technical Specifications for each unit to reflect the approved value of 470 inches for the L2 setpoint. Subsequently, the licensee found 4 additional l
l pages in the Technical Specifications for Units 1 and 2 (pages 11, 254, 255 and 277). where the 490" was referenced with respect to valve closures. The proposed changes to the Technical Specifications are to correct this error.
(This is an error toward the conservative.)
We conclude that the proposed changes to rectify this oversight are acceptabl e.
3.2.2 Surveillance Requirements in Section 4.5.B In Section 4.5.8 of the present Technical Specification on the Residual Heat Removal System (RHRS) (LPCI and Containment Cooling),
the surveillance requirements in several items do not track the correspondingly numbered limiting condition for operation (LCO) in Section 3.5.B.
For example, surveillance requirement 4.5.B.10 is the surveillance requirement for LCO 3.5.B.11 and surveillance requirement 4.5.B.12 is the requirement for LC0 3.5.B.14 The licensee proposes to change this fcr clarity by having the numbers for the surveillance requirements correspond to the numbered LCOs.
Where no surveillance is indicated, the surveillance requirement will state "No additional surveillance required." As part of their review of this section of the Technical Specifications, the licensee has proposed to increase the surveillance requirements on low pressure ECCS systems when one RHR pump is inoperable. The present Technical Specifications (Section 4.5.B.3, p.145) require that "When it is determined that one RHR pump (LPCI mode) is inocerable at a time when operability is required,
.... the operable RHRS pumps (LPCI mode) shall be demonstrated to be operable 10 days thereafter until the inoperable pump is returned to normal service." The licensee has proposed to change this to require that "the remaining RHR pumps (LPCI mode) and active components in both access paths of the RHRS (LPCI mode) and the CSS and the diesel generators shall be demonstrated to be operable immediately and daily thereafter." While we have not concluded that this increased conser-vatism is necessary, we do find the increased surveillance is acceptable.
Another change in the surveillance requirements (item 4.5.B.12, p.149) is to correct a typographical error in the present Technical Specifications.
3.2.3 LPCI Modifications By letter dated May 11, 1979, we issued Amendments Nos. 51, 45 and 23 to Facility Licenses Nos. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Units Nos.1, 2 and 3.
The Amendments added a condition to the license for each facility authorizing TVA to perform certain modifications (as described in TVA's submittals and the Safety Evaluation related to these Amendments) to change the power supply for certain LPCI valves for Units Nos.1, 2 and 3 and to eliminate the loop selection logic for Unit No. 3.
Our letter of May 11, 1979 noted that TVA had committed to submit proposed Technical Specification changes with the reload amendment request for each unit to reflect l
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(. he changes for Unit No. 3 were submitted with these modifications.
T TVA's reload amendment request of August 6,1979 and approved by Amendment No. 28 to License No. DPR-68 which we issued on November 30, 1979.] Tfte modifications to BF-1 are described in detail in the safety evaluation accompanying our letter of May 11, 1979.
In sumary, the overall modifications encompassed:
a.
Elimination of the Low Pressure Coolant Injection (LPCI) system's recirculation loop selection logic, revision of the logic and closure of the Residual Heat Removal (RKR) cross-tie valve and a recirculation equalizer valve; and b.
Changing the power supply to the reactor MOV boards that feed the motor operators of the LPCI injection valves, the recirculation pump discharge valves, and the RHR pump minimum flow bypass valves.
The change involves the use of' Class lE motor-generator (M-G) sets as isolation devices between the auto-transfer feature of the 480V reactor M0V boards and the divisional 480V shutdown boards. The auto-transfer feature will be eliminated from all 480V reactor MOV boards not protected by M-G sets.
The proposed changes on pages 97,111,112,182 and 221 reflect the above modification.
Each proposed change is discussed in detail below.
a.
The change to Table 4.2.B, p. 97 (Surveillance Requirements for Instrumentation that Initiate or Control the CSCS) removes the surveillance requirements on four reactor pressure sensors (PS 186A&B, and PS-3-187A&B) whose sole function was that of a per-missive in the LPCI recirculation loop selection logic.
Since the logic no longer exists, the sensors have been removed and deleting them from the instruments to be surveillance tested is appropriate. We find the proposed change acceptable, b.
The proposed change in Section 3.2 " BASES" at the bottom of page 111 and top of page 112 is to delete the words "provides input to the LPCI loop selection logic." This sentence discussed the J
bases for the reactor pressure sensors in "a" above.
The change is to remove the low reactor water level instrumentation as the source of a LPCI loop selection logic initiation signal, since the latter function no longer exists. We find the proposed change acceptable.
c.
The present Technical Specifications (Section 3.6.F.1 and 3.6.F.2, p.182) require that the speeds of the two recirculation pumps be maintained within 122% and 135% of one another when the core power is above 80% or below 80% of rated power, respectively.
As explained in the bases for these requirements ('p. 221, " Jet Pump Flow Mismatch), this was necessary when there was automatic l
l loop selection logic. The purpose of this limitation was to prevent the LPCI loop selection logic from selecting the wrong loop for injection which was possible for certain low probability accidents with the recirculation loop operating at large speed di fferences. Since the LPCI loop selection logic has been removed from the Browns Ferry Nuclear Plant, Unit Nos.1 and 2, there is no longer the need for surveillance requirements relating to this logic nor the need to limit the variation of recirculation pump speed for purposes associated with this logic. We find the proposed changes to the Technical Specifications to be acceptable.
3.2.4 Single Loop Operation On September 19, 1978 and September 29, 1978, we issued Amendments Nos. 41 and 43, respectively, to Facility License No. DPR-33 which authorized operation of BF-1 with one recirculation loop for the duration of cycle 2.
Cycle 2 for BF-1 ended in November 1978. During the period of single loop operation, there was a reduced limit on core maximum fraction of limiting power density (Section 2.1.B page 10) that applied only during this period. The proposed change on page 10 is to remove the limit for one recirculation loop operation since it is no longer applicable.
We find the proposed change to be desirable and acceptable.
4.0 Environmental Considerations We have determined that the amendments do 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 amendments involve 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 negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendments.
5.0 Conclusion s
We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the l
amendments do not involve a significant hazards consideration, (2) there i
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 Comission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
February 25, 1980
a
-9 References 1.
Letter, L. M. Mills (3VA). to H. R. Denton (NRC), dated October 4,1979.
2.
" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload No. 3," NEDO-24209, August 1979.
3.
" General Electric Boiling Water Reactor Generic Reload Application,"
NEDE-24011-P-A, August 1979.
4 Letter, T. A. Ippolito ('USNRC) to R. Gridley (;GE), April 16,1979 and enclosed SER.
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