ML19290B750

Safety Evaluation Supporting Amend 28 to License DPR-68

ML19290B750
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Site:	Browns Ferry
Issue date:	11/30/1979
From:	Office of Nuclear Reactor Regulation
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ML19290B749	List: ML19290B748 ML19290B750 ML19290B751
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NUDOCS 7912130717
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%.....E SAFETY EVALUATI0tl BY THE OFFICE OF NUCLEAR REACTOR REGULATION

, SUPPORTING AMENDMENT N0. 28 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 COCKET NO. 50-296 1.0 Introduction By letter dated August 6,1979 (TVA BFNP TS 127), and supplemented by two letters dated September 26, 1979 and letters dated October 10, 1979 and October 25, 1979, the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operathg License No. DPR-68 for the Browns Ferry fluclear Plant, Unit No. 3.

The proposed amendment and revised Technical Specifications (1) incorporate the limiting conditions for operation associated were to:

with the third fuel cycle, (2) reflect tacility modifications made during the current refueling outage to eliminate the low pressure coolant injec-tion (LPCI) loop selection logic, (3) add a check valve in the reactor water cleanup (RWCU) system piping as a result of rerouting this piping so that the return flow is distributed equally among the feedwater lines and (4) reflect replacement of two of the eleven safety-relief valves with valves set to relieve at 1150 psig rather than 1125 psig.

Browns Ferry Unit No. 3 (BF-3) shutdown for refueling on August 24, 1979.

Besides routine maintenance and equipment overhaul, several significant modifications were completed, including main steam relief valve (MSRV) tailpipe routing, core spray piping modifications, feedwater sparger modifications and LPCI modifications. Because of these modifications, all of the fuel was removed from the reactor vessel and stored in the spent fuel pool (SFP) while the vork was in progress.

1.1 Reload The initial core loading for Browns Ferry Unit No. 3 consisted of 764 of the single water rod 8 X 8 fuel assemblies, each containing 63 fuel rods.

During the first refueling in September 1978, 208 of the fuel assemblies were replaced with 8 X 8 fuel assemblies containing 62 fuel rods in each.

During the present refueling ' outage, an additional 144 of the initial fuel bundles were replaced with P 8 X 8 fuel assemblies, each containing 62 fuel rods. The prepressurized fuel assemblies (P 8 X8R) are essentially ident-ical 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 Our evalua-one atmospheres of helium to minimize fuel clad interaction.

tion of the P 8 X 8 R 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 inventory of 79121307/ 7 1553 109

. h ?luim gas improves the gap conductance between fuel pellets and rimidinin resulting in reductions in fuel temperatures, thermal expansion and tission 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 upport of this reload application for BF-3, TVA Specifications).

I dated August 6,1979, and supplemented by letter (2)

I submitted by letter dated October 25, 1979, a supplemental reload licensing document (31 pre-paredbyGeneralElectricCompany(pl..E.) for TVA and proposed changes to the BF-3 Technical Specifications 1.2 LPCI Modification By letter dated May 11, 1979, we isseed Amendments Nos. 51, 45 and 23 to Facility Licenses Nos. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant Uni t s Nos.1, 2 and 3.

The Amendments added a condition to the license for each facility authorizing lVA to perform certain mod-ifications (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 tne loop selection logic for Unit No. 3.

Our letter of May 11, 1979 noted that TVA had committed to complete the modifications for BF-3 by the end of the second refueling outage (the curre it outage) and to submit proposed Technical Specification changes with the reload amendment request for each unit.

For BF-3, the modifications consisted of the following:

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 (RHR) cross-tie valve and a recirculation equalizer valve; and b.

Changing the power supply to the recirculation pump discharge valves, LPCI injection valves, RHR pump minimum flow bypass valves, and RHR test isolation valves.

The change also modifies independent valve a.c. power supplies and modifies d.c. power supplies to 4kV shutdown board control power to provide adequate independence such that a

• tation battery failure does not jeopardize core cooling capabilities.

} dated August 6 1979, TVA submitted proposed changes By their letter to the Technical Specifications (4} associated with the above modifica-tions. 'Since this modification constitut9s a change to the Emergency Core Cooling System (ECCS), TVA by lettertS) dated September 26, 1979 also transmitted a revised " Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3."

The modifications to the BF-3 1553 110

. ECCS make it functionally identical to the ECCS system currently installed at Browns Ferry Units 1 and 2.

The purpose of the changes is to upgrade overall performance of the BF-3 ECCS by assuring delivery of LPCI injection flow to the core in the event of a postulated break in the suction side of the recirculation system piping. By improving ECCS performance for this currently limiting break, additional margin to the 2200 F peak cladding temperature limit can be achieved. Our review of the BF-3 Loss of Coolant Accident reanalysis results together with those Technical Specifications required to implement the analysis results and assumptions is contained within this safety evaluation.

1.3 Mo_dification of React _or_ Water Cleanup System Piping In the past, cracks have been detected in some BWR piping systems. The staff's investigation and evaluation of the causes of these cracks and recommende getions to minimize cracking potential has been reported in 71 a revision to which was issued in October 1979, and NUREG-NUREG-Q313 0531t8 The cracks have generally been attributed to stress corrosion 1

cracking. For this to occur, two elements must be present - a corrosive environment and stress. High purity water is corrosive to any metal.

Since the concentration of ions such as iron, chromium and nickel in demineralized water is below the solubility limit and the water is not buffered, the water tends to dissolve or corrode the metal surface. This condition can be aggravated by crevices (such as might exist at fittings or welds) since there is the potential for oxygen concentration cells, and by other conditions in the piping systems (such as stagnant flow conditions).

The other causitive element - stress - can result from residual stresses left in the piping during manufacture, stresses induced during fabrication (particularly stresses created by weld joints) and stresses created by operating conditions, such as those caused by thermal shock, vibration, water hammer, etc. The objective is to reduce either the stresses or the corrosivity of the environment - and preferably both - below the threshold required to initiate and propagate stress corrosion cracking.

One of the facility modi ations recommended by the staff and by the General Electric Company 1 is to modify the Reactor Water Cleanup (RWCU)

System return piping so that the return flow is distributed equally among the feedwater lines. TVA performed this modification on BF-3 during the current refueling outage. This modification allows feedwater to be mixed with the higher temperature RWCU return water at low flow rates thereby lessening the thermal cycling on the feedwater nozzle and the consequent thermal fatigue. Because this modification entailed the addition of a check valve, by letter (10) dated September 26, 1979, TVA requested a change to the Technical Specifications to revise Table 3.7.G to include the required check valve.

(This letter is separate from the letter of the same date in reference 5.)

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. 1.4 Replacement of Two Safety-Relief Valves Prior to the refueling outage, BF-3 had 11 Target Rock safety-relief valves. Four of these valves were set to relieve at 1105 psig, 4 were set to relieve at 1115 psig and 3 at 1125 psig. There have been some problems noteg! y th the Target Rock valves as discussed in I&E Circular No. 79-18 tl' <, and I&E Bulletin 74-4 and IE Bulletin 74-4a(12,g During the current refueling outage, TVA has instalied two 6R10 Crosby Safety Relief Valves (SRVs) at BF-3 to obtain performance experience with these valves for possible future use at Browns Ferry, Hartsville and Phipps Bend. At the latter plants, the safety-relief valves will also be grouped as at Browns Ferry with respect to set-point pressure; however, whereas the highest setpoint at Browns Ferry is presently 1125 psig, at Hartsville and Phipps Bend the lowest setpoints will be 1165 psig. To obtain experience at a more prote-typical pressure, TVA proposed that the two replacement Crosby relief valves be set at 1150 psig. The two Crosby SRVs set at 1150 psig will replace two Target Rock valves set at 1125 psig in locations G and H which are not automatic depressurization system (ADS) locations.

The Crosby SRV is a simple, direct-acting, spring-loaded valve with an external pneunatic piston. Safety valve action occurs when the inlet pressure forces exceed the spring load and force the valve disc off of its seat.

For manual actuation, the external pneumatic piston is capable of opening the valve agairst the force of the spring at any steam pressure down to O psig. The pneumatic operator is so arranged that if it malfunctioned it would not prevent the valve disc from lifting if steam inlet pressure reached the spring set pressure.

Since the Target Rock valves on Browns Ferry Unit No. 3 have had their throats enlarged to provide increased capacity, the capacity of each of the two Crosby replacement valves is 94.3% of each of the modified Target Rock valves when compared at the same inlet pressure.

By letter (I ) dated October 10, 1979, TVA submitted proposed changes to the Technical Specifications associated dth replacement of 2 of the 11 safety-relief valves and a revised anal, 314) for the limiting transients to evaluate the impact of using tne 2 Crosby SRVs set at 1150 psig in place of 2 of the high set (1125 psig) Target Rock SRVs.

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. 2.0 Discussion 2.1 Reload This refueling (Reload 2) is the first for BF-3 to incorporate GE's P8xbR fuel aesign on a Daten Dasis. The description of the nuclear ano mechanical cesign of the Reload 2 PdxoR fuel and the exposea un-pressurized 8x6 and 8x8R fuels, used in the initial and first reload is contained in GE's generic licensing topical report for owr cores, fl5) Reference 15 also contains a complete set of references reloads to topical reports which describe GE's analytical methods for the nuclear, thermal-hydraulic, transient and accident calculations per-formed for this reload together witn information on the applicability of these methods to cores containing a mixture of different fuel de-signs. Portions of the plant-specific data, such as operating conditions and design parameters, which are used in transient and accident calculations, have also been included in the topical report.

Ob, 17)of GE's generic reload licensing Our safety evaluations topical report and report amendment concluded that the nuclear ano mechanical design of P8x8R fuel used in this reload and GE's analytical methods for nuclear, thermal-hyaraulic, transient and accident calcu-lations, as applied to cores containing a mixture of fuel types, are c.cceptabl e.

Our acceptance of the nuclear and mechanical design of the standard 8x8 (one water rod) fuel was expressed in the staff's evaluation (18)of the information in Reference 19.

As part of our evaluation (16) of Reference 15,we found the cycle-independent input data to be used for the reload transient and accident. analyses for BF-3 to be acceptable. The supplementary cycle-dependent information and input data are provided in Reference 3. which follows the format and content of Appendix A of Reference 15.

As a result of the staff's generic evaluations 06,17) of a substantial number of safety considerations related to the use of P8x8R fuel in mixed core loadings with 8xbR and 8x8 fuel, only a limited numoer of additional review items are included in this evaluation. These in-clude the plant and cycle-specific analysis input data and analysis results presented in Reference 3, and those items identified in Reference 16 as requiring special attention during BWR reload reviews.

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t

. 2.2 LPCI Modification The most severe pipe break locations for a boiling water reactor are in the recirculation system discharge and suction line piping. Large breaks occurring in these locations result in the most rapid reactor system depressurization rates and the earliest boiling transition times and uncovery times. For plants with LPCI loop-selection-logic a break in either the recirculation suction line or discharge line, when coupled with a postulated failure of the LPCI injection valve in the intact loop, results in no LPCI flow reaching the core. That is, since all flow is directed to the intact loop through a single injection point, failure (to open) of a single LPCI injection valve results in no LPCI flow reaching the core. Thus for these plants reflood times for recirculation line breaks and a postulated LPCI injection valve failure result in the longest hot node uncovery times since only the two operable core spray systems are available to provide core cooling and to reflood the core.

The worst break size, break location and single failure conoition for a plant with LPCI loop selection logic is generally the complete severance of the F r BF-3(with largest (suction) line, with LPCI injection valve failure.

Q LPCI loop-selection-logic), the suction line break results(2911n the most rapid jet pump uncovery, boiling transition and hot node uncovery times, with the most delayed core reflooding time due to the unavailability of LPCI.

Accordingly, for plants with LPCI loop selection logic, the suction Dreak generally results in the highest peak cladding temperature and establishes the basis for the MAPLHGR limits for the plant.

In order to lessen the severity (PCT) of this limiting (suction) brea_K, with assumed LPCI injection valve failure condition, the licensee modified the BF-3 ECC system during the second refueling outage. The LPCI modification consists of eliminating the LPCI loop-selection-logic system and permanently piping the aischarge flow from two LPCI system pumps to one recirculation system discharge line and pennanently piping the discharge flow from the other two LPCI system pumps to the second recirculation discharge line. Adaitionally, the modification will result in both recirculation line discharge valves closing af ter blowdown following a LOCA. These valves are located between the LPCI in-jection point on the recirculation discharge line ana any potential break location on the suction line. The flow from the LPCI system pumps connected to the broken recirculation line is therefore isolated from any suction line break while the injection flow from the other system is also isolated because it is connected to the unbroken line (since the recirculation loop equalizer valve is locked closed). With this LPCI injection arrangement, only one LPCI loop can be dis-abled by any single failure and the largest (suction line) break can now derive credit for earlier reflooding cue to the availability of at least one half of 1553 114

. the LPCI system. The resulting faster core flooding and attendant reduced period of hot node uncovery reduces the PCT calculated for the suction line break to the extent that it potentially could become non-limiting relative to a recirculation discharge line break. At the same time with the subject LPCI modification, the discharge break consequences remain unchanged. All LPCI flow is still lost out the break for the LPCI system connected to the broken loop (since it cannot be isolated from the break by the recirculation discharge line isolation valve),

while a postulateo LPCI injection valve failure prevents LPCI flow from reaching the core via the intact recirculation loop. That is, as was the case with LPCI loop selection logic, no LPCI flow is available to flood the core. Therefore, although the discharge break is in a smaller diameter line than the suction line (and would normally be expected to yielo a lower PCT), the lack of LPCI flow delays reflooding (relative to the suction break where LPCI flow from at least one system is now available) to the extent that this creak location can become l imi ti ng.

Accordingly, for BF-3 the.'et benefit of the proposed LPCI modifi-cations is that the formerly limiting (in terms of PCT and MAPLHGR requirements)

DBA suction line break becomes less severe and thereby improves overall ECCS performance over the spectrum of breaks and worst single failures.

3.0 Eval ua tion 3.1 Reload 3.1.1 Nuclear Characteristics For Cycle 3,144 fresh pressurizea type P8DRB265L fuel bundles will be loaded into the core. The remainder of the fuel bundles in the core will be a combina-tion 8x8 and 8x8R fuel bundles exposed during the previous two cycles.

The fresh fuel will be loaded and the previously peripheral fuel will be shuffled inwara so as to constitute an octant-symmetric core pattern, which is acceptable.

Based on the data provided in Sections 4 and 5 of Reference 3, bcth the control rod system and the st4ndby liquid control system will have an acceptable shutdown capability during Cycle 3.

3.1.2 Thermal-Hydraulics 3.1.2.1 Fuel Cladding Integrity Safety Limit MCPR As stated in Reference 3, for BWR cores which reload with GE's Pdx8R fuel, the allowable minimum critical power ratio (MCPR) resulting from either core-wide or localized abnormal operational transients is equal to 1.07.

When meeting this MCPR safety limit during a transient, at least 99.97, of the fuel rods in the core are expected to avoid boiling transition.

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. The 1.07 safety limit minimum critical power ratio (SLMCPR) to be used for Cycle 3 is unchanged from the SLMCPR previously approved for Cycle 2.

The basis for this safety limit is acuressed in Reference 15, wnile our generic approvals are given in References 16 and 17.

3.1.2.2 Operating Limit MCPR Various transient events can reduce the MCPR from its normal operating level. To assure that the fuel cladding integrity safety limit MCPR 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 Getermine whicn event results in the largest reduction in the minimum critical power ratio. These events have oeen analyzed for both the exposed 6x8 and 8x8R fuel and the fresh PoxbR fuel. Addition of the largest reductions in critical power ratio to the safety limit NCPR establishes the operating limits for eacn fuel ty pe.

The transient events analyzed were load rejection without bypass, feedwater controller failure, loss of 100*F feedwater heating and control rod withdrawal error.

3.1.2.2.1 aonormal Operational Transient Analysis Nethods The generic methods used for these calculations, including cycle-irwpendent initial conditions and transient input parameters, are described in Reference 15. Our acceptance of the cycle-incependent values appears in Reference 16. Additionally, our evaluation of the transient analysis methods, together with a description and summary of the outstanding issues associated with these methods, appears in R ef erence l 6.

Supplementary cycle-dependent initial conuitions and transient input parameters used in the transient analyses appepr jn the tables in Sections 6 and 7 of Reference 3.

Our evaluation (17, has also addressed the methods used to develop these supplementary input values.

3.1.2.2.2 Thermal-Hydraulic,Hethods At the time we completed our evaluation of the generic methods, the acceptability of the GEXL critical power correlation (20), for use in connection with the retrofit fuel design, had not been adequately documented by GE. The staff found, however, that the then available 8x8R critical power test cata was sufficient to support the accept-ability of GE's dxeR fuel design for BWR core reloads for one operating cycle. Accordingly, we stated (16) that future BWR core reload applications involving retrofit 8x8 fuel for a second operating cycle would have to include additional information which adequately justified the correlation for application to 6xdR fuel operating beyond one cycle. Since the Reload 2 licensing suomittal(3) dia not adoress this issue, we requested (21) that the licensee provide the request by referencing information (g2) licensee responded to our required additional information. Th furnished to the staff by GE which references a report (23) prepared by GE on this same subject.

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A Reference 23 provides the results of full scale critical power tests performed on 8xbR fuel bundles. The tests, which included both tran-sient and steady-state simulations, followed the same approved pro-cedures(20) used for the standard 8x8 (single water rod) ana 7x7 (all fueled rods) tuel cesigns. The analysis of a total of S77 stcacy-state data points was performed using methods also previously approvea Dy the staf f.

The data, involving nine test assemDlies which spanned a range of local power peaking and flow conditions, showed according to GE, that the GEXL correlation was applicable to the exBR fuel if adjust-ment were made to the adaitive constants used in the formulation of the rod-by-rod R-f actors. The local power peaking dependent R-f actors are based on the new additive constants shown in Figure 3-1 of Reference 23 which were also used for the BF-3, Reload 1, exbR critical bundle power predictions. Using tnese new aaditive constants, bE performed a data analysis to assess the accuracy and precision of the GEXL correlation.

The results of this analysis showed that the correlation fit provides for a mean predicted-to-measurea critical power ratio of 0.9879 with a standard deviation of 0.0234.

When viewed over the range of its applicability (which is the same as the standard 8x8 fuel), the GEXL correlation is therefore somewhat conservatively biased while the statistical variation between the pre-dicted and measured critical power is somewhat less than that associated with the standard 8x8 assembly (20), i.e., 2.34% vs 2.8%.

Thus, when viewed over its range of applicability, the 8x8R GEXL correlation (with new additive constants) has somewhat better precision in pre-dicting dx8R critical bundle powers than the 7x7 and 6x8 GEXL formulations are for predicting 7x7 and 8x8 critical bundle powers respectively. Furthermore, f rom these results it may also be con-cluded that the 3.6; standard aeviation and best estimate assumption of the GEXL correlation (which were actually used in the GETAB statistical analysis to aerive the 1.07 safety limit hCPR) cound the statistical characteristics associatea with the subject 6x8R GEXL correlation.

The additional information furnished by GE is also intended to be applicable to all BWR cores which contain 8x6R fuel. Accordingly, this information is also currently being generically reviewed by the staff. Although our evaluation is not yet Complete, baseo on our review to date, we believe that for the range of testing, the dxbR GEXL correlation has an acceptability and applicability which is equivalent to the 7x7 and bx8 GEXL correlations previously approved by tne staff. From our review of the subject data to date, we have il'.o observed that f or those cri tical power test condi tions spet i t i-

y representative of second cycle f uel operating at normal operating u

al-hydraulic state points, the correlation is somewhat nonconserva-ti., in its predictions. This observation focuses in on a correlation behavioral concern not explicitly addressed in tne overall uETAU metnous approved ( 23) for the 7x7 ana 8x8 fuel types.

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! Again, this subject is being generically reviewed by the staff.

However, until this review is complete, we believe that for Cycle 3 of BF-3, there is sufficient conservatism implicit in the generic de-termination of the 1.07 safety limit MCPR to offset a possible non-conservatism associated with this concern. That is, specifically, the generic GETAB statistical analysis assumed a 3.6% correlation uncertainty while GE's analysis of the 8xbR test data results in a 2.34s standard deviation. Additionally, the generic evaluation considered an all 8x8R equilibrium core, whereas the Cycle 3 BF-3 core involves 8x6, 8x6R and P8x8R fuel in a non-equilibrium condition. In view of these conservatisms (which are representative of a typical non-equilibrium 8xBR reload core) we believe that the overall thermal-hydraulic (GETAB) methods are adequate for establishing conservative MCPR operating limits for Cycle 3 of BF-3.

However, as 8x8R equilibrium conditions are approached, this conservatism will diminish.

In order that this conservatism not be substantially eroded, this issue should be addressed for the next reload cycle of BF-3, 3.1.2.2.3 Plant System Transient Simulation Methods in the analysis of the load rejection with bypass failure and the feedwater controller f ailure transients, the licensee has taken credit for the beneficial effects of the prompt recirculation pump trip (RPT) as was the case in the previous operating cycle.

The RPT feature has the effect of reducing the transient ACPR curing reactor core pressurization events, by tripping breakers in the electrical circuit Detween the motor-generator sets and the recirculation pumps on closure of turoine stop or control valves.

The prompt RPT immediately reouces core flow and thereby increases core voios. The rapid voiding provides negative reactivity which supplements scram negative reactivity.

In this manner, the RPT reduces the thermal power rise during pressurization events. This RPT teature is a thermal margin improvement option which was not generically (l71 approved in our evaluation of the reference reload topical report.

The CPR benefit associated witn the prompt RPT was calculated with the REDY code.(24) The REDY code employs a two node steamline thermal-oint kinetics neutronics model. Several hydraulic model and gM) at the Peach Bottom Unit 2 boiling water pressurization tests reactor were intended to show the validity of these REDY models.

The experimental results showed, that the REDY steamline mooel did not accurately predict the pressurization rate which causes the re-duction in CPR. Futhermore, the REDY point kinetics model could not simulate the transient axial reactivity in the core. GE immediately pro-vided calculational comparisons of REDY to test results, and attempted to demonstrate that although REDY did not accurately model some transient effects, it did provide a conservative basis for current licensing calculations.

We agreed with GE's general conclusion that REDY provides a conserva-tive calculation for the current licensing basis transients on operating reactors. However, we also recognized that REDY's inability to accurately predict pressurization rate and axial reactivity response, limits the 1553 ii8

. simulation of RPT effects. The Peach Bottom tests demonstrated in the steam lines.(26,27)y of REDY to simulate) a pressure wave (the existence of inabilit In addition, it was noted that the power rise associated with the pressurization was significantly greater in the upper portion of the core than in the lower portion.

Quantative comparison of the tests with REDY calculations inoicated that the REDY model underpredicted the pressurization rate butThus, overpreaicted the core's response to pressurization effects.

there are two aiscrepancies between REDY simulated effects and real transient's effects. One is non-conservative and the other is con-servative. It is not possible to state from these comparisons alone which effect woulu predominate for a given transient.

Af ter the analysi:. of the test results, con'parisons were made between REDY simulations.ind simulations using detai eg)steamline modeling 2

These comparisons, dnd a time-vary 10'J axial power distribution.

although limited, indicate a trend in which REDY-based calculations conservatively preoicted a CPR for more severe transients but under-predictdCPR (for a given set of input parameters) for milder transi ents. (28 ) These calculations also showed that the ACPR benfits derivea from the kPT feature may be overprecicted by REUY when compared with tne predictions of the detailed steamline and core mooel.

In view of this information, we decided to take no action for three reasons:

(1) operating limit MCPRs are always based upon the most severe transient for each fuel type, (2) these limiting transients were sufficiently severe to be in the range where REDY-based cal-culations are conservative, and (3) GE was developing a more sophisticated transient simulator to accurately predict the questioned phenomena.

However, with the addition of the RPT feature, the limiting pressure and power increase transient analyses generally predict adCPR in the range where REDY is less conservative. We find that full credit for the RPT effect cannot be justifdied solely on a REDY analysis.

The first Twa alternatives have been considered to resolve this issue.

alternative is to provide additional justification for the proposed speci fi ca ti on.

The GE ODYN code has more nodes to model steamline dynamics than REDY and also has a one-dimensional axial core neutronics model. ODYN's development has been cased on first principles and ODYN is Currently under a staff verified by the Peach Bottom tests.

When approvea. ODYN will be used for calculating the ACPR review.

tor pressurization events such as the loao rejection with Dypass.

Until approval, we believe that ODYN could be used to simulate tnerecy, provice assurance of its c.CPR uenefit.

the RPT effects and, During this time, we will accept the greater ACPR of tne ODYN and Once ODYN receives generic approval, we will accept RLUY calculations.

an ODYN calculation. However, the licensee has beer Sformed that other justification which the

ensee submits we will evaluate any and all applicable calculations and data which become available to us 1553 119

i D

"D D ~T

' through other channels.

The other alternative is to sonnervatively bound the ACPR from the REDY calculation. With the RPT modeled, the available comparison of ODYN and REDY predictions shows a ACPR di f ference of about 0.02.

This calculation is for a specific BWR which is different in plant size and core loading tnan the Browns Ferry Units.

From this information we and the licensee have agreed that a conservative bound to the REDY calculation with RPT would oe assured with a U.03 f CPR increase for rapid presrurization transients.

Abnormal Operational Transient Analysis Results 3.1. 2. 2. 4 The licensee reports in the reload supplement (3) that the most limiting event for the 6x8R and P6xbR fuel types is the load re-This transient results in a CPR reduction of jection without bypass.

0.15 and 0.16 for the 8xeR and Pdx6R fuel assemolies, respectively, as predicted by the REDY coce. Since the load rejection without by-pass transient is a pressurization type event the.03 increase is applicable to these results. For the standard 8x8 fuel type the control rod withdrawal event is reported to be most limiting, with a CPR reduction at 0.21.

The next most severe event for the 8x0 fuel is the load rejection without bypass with a transient 4CPR of 0.15 Thus the control rod withdrawal remains as predicted by REDY.

limiting relative to the load rejection transient even when a 0.03 In response (2) to CPR adjustmeg)is applied to the latter event.on this subject the licensee has pro our concern (

increase the fuel dependent operating limits by.U3, as appropriate, on an exposure dependent basis. Since the severity of pressurization events increase toward end of cycle, the licensee has proposed (2) to add a.03 penalty to the dx8R and P8x6R REDY predictions (3) for exposures between EOC3-2000 Mwd /T and EOC3 for estaDlishing the required operating limits. No penalty has been proposeo for exposures between BOC3 and EOC3-2000 Mwa/T for these type fuels. From our review we nave concludeo that the licensee has not proviced an adequate basis for the proposed operating limits from BOC3 to EUC3-zuuo Hwc/T. That is the intermediate exposure operating limits were not developed using the methods described in Reference 15, nor were adequate alternative evaluation bases provided. This position has been discussed with the licensee and he agreed to accept a single fuel dependent operating limit based on the end-of-cycle REDY analysis with a.03 CPR penalty acced. Accordingly, based on our review of the licensee's suomitted calculated results and the

.03 CPR adjustluent applicable to REDY calculations for pressurization transients whicn mouel the beneficial effects of tne RPT f ea ture, the licensee will be required to meet the following MCPR operating 1imits:

Fuel Type MCPR Operatina Limit 6xo 1.20 dxcR 1.26 P8x8R 1.26 1553 120

. With BF-3 operated in accordance with the above MCPR operating limits, we agree tnat the 1.07 SLHCPR will not be violated even in the event of the most severe abnormal operational transients.

3.1.2.3 Fuel Cladding Integrity Safety Limit LHGR The control rod withdrawal error and fuel loading error events were reanalyzed by the licensee to also determine the maximum transient linear heat generation rates (LHGRs). The results for BF-3 Cycle 3 show that the fuel type-dependent and exposure-dependent safety limit LHGRs, shown in Table 2-3 of Reference 15, will not be violated should these events occur.

Thus, fuel failure due to excessive cladding strain will be precluded. We find these results, which adequately account for the effects of fuel densification power spiking, to be acceptable.

3.1.3 Accident Analysis 3.1.3.1 ECCS Appendix K Analysis On Uecemoer 27, 1974, the Atomic Energy Commission issued an Order for Modification of License, implementing the requirements of 10 CFR 60.46,

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." One of the requirements of the Orcer was that prior to any license amendment authorizing any core reloading... "the licensee shall submit a re-evaluation of ECCS performance calculated in accordance with an acceptable evaluation model which conforms to the provisions of 10 CFR Part 50.46."

The Order also required that the evaluation shall be accompanied by such proposed changes in Technical Specifications or license amendments as may be necessary to implement the evaluation assumptions and results.

For Cycle 2, the licensee re-evaluated the adequacy of BF-3 ECCS performance in connection with the retrofit 8x8 reload fuel design. The methods used in this analysis were previously approved by the staf f.

For Reload I, we reviewed the ECCS analysis results submitted by the licensee for the Cycle 2 reload fuel and conclucea that BF-3 would be in conformance with all the requirements of 10 CFR 50.46' and Appenaix K to 10 CFR SO when operated in accordance with tne 8x6R HAPLHGR versus Average Planar Exposure values which appeared in the proposeo plant Technical Specifications. Except for prepressur-ization, the Reload 2 fuel is the same design as the Reloaa 1 fuel.

In Reference 17, we stated that LOCA analyses previously performed and accepted for unpressurized 8x8 fuel are conservatively bounding for prepressurized fuel of that type (enrichment pattern). Accordingly we find it acceptable for the licensee to utilize the 8x8R MAPLHGR vs Average Planar Exposure technical specification limits for the reload P8x8R fuel in connection with showing compliance with the requirements of 10CFR50.46.

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3.1.3.2 Control Rod Drop Accident For Cycle 3, the key plant-specific and cycle-specific nuclear char-teristics for the worst case control rod drop accident (CRDA) occurring during hot startup conditions are conservatively bounded by the values used in bounding CRDA analysis given in Reference 16.

The bounding analysis, which includes the adverse effects of fuel densification power spiking, shows that the peak fuel enthalpy will not exceed the 280 cal /gm design limit. Therefore, for Cycle J of tuel enthalpy associated with a CRDA from the hot BF-J, the peak startup conaition will also be within the 260 cal /gm design limit.

For the worst case control rod drop accident occurring curing cola startup conditions, however, not all of tne key plant-specific and are within the values used in cycle-specific nuclear cnaracteristicsThat is, although the actual Cycle 3 Ocppler the generic CRDA analysis.

coefficient and scram reactivity shape function conservatively f all within the values assumed in the bounding analysis, the accident re-Therefore, the licensee has performec activity shape function does not.

a plant-specific control rod drop accident analysis applicable to BF-3 for Cycle 3.

The results of this analysis, using the approved methods aescribed in Reference 16, show that the positive reactivity insertion rate of the dropped rod is sufficiently compensatea by Doppler feed-back and scram reactivity effects to limit the peak energy ceposition in the fuel to 278 cal /gm.

Thus, we conclude that the peak enthalpy associated with a control rod drop accident occurring from any in-sequence control rod move-ment will be below the 280 cal /gm design limit.

3.1.3.3 Fuel Loading Error The licensee has considered the effect of postulated fuel loading errors on bundle CPR. An analysis cf the most severe tuel loading errors were performed using GE's revised analysis methods which The have previously been reviewed and approved by the staff.

results show that the worst possible fuel bundle mistoadings will not cause a violation of the 1.07 safety limit MCPR even when assuming the proposed OLMCPRs. These results include the application of a 0.02 penalty factor appliea to the CPR results of the misoriented fuel bundle analysis, as required by our approval of the revised methods.

Tnus, the requirea operating limit MCPRs will effectively preclude DNB related fuel f ailures caused by either fuel cladding overheating or cladding oxidation, which might otherwise occur because of a fuel These results are acceptable to the staff.

loading error.

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3.l.4 Overpressure Analysis For Cycle 3, the licensee has reanalyzed the limiting pressurization event to demonstrate that the ASME Boiler and Pressure Vessel Code requirements are met for BF-3.

The methods used for this analysis, when modified to acc(ouqt for one failed safety valve, have also been previously approved IS) by the staff. The acceptance criteria for this event is that the calculated peak transient pressure not exceed 110; of design pressure, i.e., 1375 psig. The reanalysis shows that the peak pressure at the bottom of the reactor vessel does not exceea 13U0 psig for worst case end-of-cycle conditions, even when assuraing the effects of one failed safety valve. This is acceptaDie to the staff.

3.1.5 Thermal-Hydraulic Stability A thermal-hydraluic stability analysis was performed for this reload using the methods described in Reference 15. The results show that the fuel type dependent channel hydrodynamic stability decay ratios and reactor core stability decay ratio at the least staole operating state (corresponding to the intersection of the natural circulation power curve and the 106t rod line) are 0.273 (8x6R/P8x8R), 0.3o3 (Bx6) and 0.79 respectively. These predicted decay ratios are all well below the 1.0 Ultimate Performance Limit decay ratio proposed by GE.

The staff has expressed generic concerns regarding reactor core thermal-hydraulic stability at the least stable reactor condition.

This condition could oe reached during an operational transient from high power if the plant were to sustain a trip of both recirculation pumps without a reactor trip. The concerns are motivated Dy increasing decay ratios as equilibrium fuel cycles are approached and as reload fuel designs change. The staff concerns relate to both the conse-quences of operating with a decay ratio of 1.0 and the capability of the analytical methods to accurately predict decay ratios. The General Electric Company is addressing these staff concerns through meetings, topical reports and a stability test program.

It is expected that the test results and data analysis, as presented in a final test report, will aid considerably in resolving the staff concerns.

Prior to Cycle 3 operation, the staff as an interim measure, acced a requirement to the BF-3 Technical Specifications which restricted planned plant operation in the natural circulation mode. Continuation of this restriction will also provide a significant increase in the reactor core stability operating margins during Cycle 3.

On the basis of the foregoing, the staff considers the thermal-hydraulic stability of BF-3 during Cycle 3 to be acceptable.

3.1.6 Physics Startup Testing The licensee will pefona a series of physics startup tests and pro-vide assurance that tne conditions assumed for the transient and accident analysis calculations will be met during Cycle 3.

The test will verify that the core has been loaded as intended, that the incore monitoring system is functioning as expected and that the process computer has been reprogrammed to properly reflect 1553 123

- changes associated with the reload. The test program is consistent with that previously found acceptable for BF-3.

We find this test program acceptaole for Cycle 3.

3.1.7 Technical Specifications The proposed Technical Specifications for Cycle 3 operation of 0F-3 include revisea operating limit minimum critical power ratios (OLhCPRs) for each fuel type in the core. As discusseo in Section 3.1.2.2 herein, the fuel-dependent operating limit MCPRs proposed by the licensee have been adjusted, with his agreement, to account for possible excess end-of-cycle recirculation pump trip benefits cal-Thus culated by the REDY Code for pressurization type transients.

the OLMCPRs agreed to by the licensee and the staff for the entire third cycle are 1.28,1.25 and 1.26 for the 8x8, 8xbR and P8x8R fuel types respectively. These MCPR operating limits are acceptable.

Additionally the licensee has proposed MAPLHGR vs Average Planar Exposure limits for the prepressurized reload 8x8R, which are the same as the unpressurized 8x8R fuel. As discussed in Section 3.1.3.1.

this is acceptable.

3.2 LPCI Modification 3.2.1 Codes and Methods The reanalysis of the Loss of Coolant Accident for Browns Ferry Unit No. 3 with LPCI modifications was performed using a General Electric evaluation modelwhichisgenerallydescribedinReference(g)modelchangestg2m)

The model us9 fqq this analysis also includes previously approved made to the REFLOOD and CHASTE computer codes. Additionally, other pre-viously approved (31) model changes (34,35,36) which take into account the beneficial effects of alternate reflood flow paths (via holes drilled into the fuel assembly lower tie plates) have been included in the reanalysis.

In sunmary therefore, the LOCA analysis of BF-3 with LPCI modifications was performed using approved calculational mo;als and methods.

3.2.2 Analysis Results 3.2.2.1 Lead Plant Reference In support of the BE-3 LOCA reanalysis, the licensee has referenced a pre-viously approved (37} Loss-of-Coolant Accident anlaysis(38) performed for the James A. FitzPatrick Nuclear Power Plant (JAF). James A. FitzPatrick is the

" lead plant" BWR/4 with LPCI loop-selection-logic removed. The lead plant reference provides detailed and expanded analysis results and documentation which justifies the extent to which break size, break location and single failure combinations must be considered when evaluating the LOCA consequences of specific BWR/4s with LPCI modifications.

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. The results from Reference 38 show that the most limiting breaks occur in the recirculation piping when failure of the LPCI injection valve is assumed.

In particular the results for JAF show that the most limiting break location is the recirculation discharge line rather than the larger diameter recirculation suction line and is due to the effects of the LPCI modification associated with JAF, BF-3 and other BWR/4s. The basis for the discharge break being more limiting than the suction break is discussed in Section 2.2 herein. Furthermore, the break spectrum (i.e. peak cladding temperature vs break size)for the lead plant shows that a break in the discharge piping have a break area equal to approxi-mately 80% of the area associated with the largest discharge line break is l imi ti ng.

The reason the limiting break size is less than 100% of the maximum possible limiting location break area is provided in our safety evaluationt3/:

for the lead plant.

It should be noted, however, that for plants with LPCI modification, such as JAF and BF-3, the peak cladding temperature resulting from a recirculation discharge line break and the PCT resulting from a recirculation suction line break are very nearly the same. That is, small differences in reactor system aesign (e.g., flow areas inside the vessel, active fuel and bypass regions; exact pipe sizes; exact design of the LPCI system) determine which break location and break size is actually limiting for any particular plant. Based on our review of the lead plant reference we conclude that minor differences between the lead plant and BF-3 resulting in a change in worst break location or break size between these plants would not significantly effect our conclusions reqarding the break spectrum, the worst break location (on the recirculation line piping as opposed to other pipes such as the feedwater or core spray lines) or worst single failure. Accordingly, we conclude that the James A.

Fitzpatrick Loss of Coolant Accident analysis is an acceptable lead plant ref erence for BF-3.

3.2.2.2 Plant Specific Results Supplementing the lead plant analysis, the licensee has submitted additional ECCS performance calculations (6) which specifically model the BF-3 plant with LPCI loop-selection-logic removed. These plant-specific analyses provide detailed results for the spectrum of postulated breaks occurring in the BF-3 recirculation suction and discharge piping with assumed LPCI injection valve failure. Similar to the lead plant analysis, the LOCA analysis performed for BF-3 shows that the most limiting break location is the recirculation dis-charge line.

In the case of BF-3, the limiting break area (i.e. the design basis accident) is approximately 66% as large as the largest discharge line 1553 125

. break assuming f ailure of the injection valve in the int 6ct loog. )Formerly, the BF-3 (with LPCI loop-selection-logic) LOCA analysis resultsl29 showed that the limiting Dreak size, break location ano single failure conoition was the complete severence of the suction line piping and LPCI injection valve failure. The ODserved shif t of the limiting Dreak location from the suction line to the discharge line is not unexpected and is prinicply due to the LPCI modification oescribed earlier. Complete severence of the recirculation suction line is now the second most limiting break size and location for 8F-3.

The results for BF-3 have also been compared with the most recently accepted LOCA conformance calculations (39, 40) performed for Browns Ferry Units 1 and 2.

Both units were analyzed with LPCI loop-selection-logic removea.

The comparison shows that tne limiting break size and location is different for UF-3 than the limiting break size and location for Units 1 and 2.

For Units 1 and 2 complete severence of the larger diameter suction line piping is limiting while a break in the discharge line piping having a break area equal to 60% of the area associated with the complete severence of the dis-charge pipe was shown(39, 40) to be the second most lim'iting size and location.

The fact that the worst break size and location is different among these virtually identical LPCI-modified BWR/4s, can ce traced to the cifferent fuel types (incluaing number of fuel issemblies with drilled lower tie plates) loaded in the respective cores. Tb cot es of Browns Ferry Units 1 and 2 contain both 7x7 and 8x8 fuel types land not all fuel assemblies drilled) while the BF-3 core contains only 3x8 fuel types (with all fuel assemblies drilled). On a system level, cores with 7x7 fuel tend to reflood up to the high pc.4er axial plane somewhat later cue to the limiting effects of counter-current flow on the core spray contribution to vessel reflooo rate. This can be seen for example by comparing the 6F-1 and BF-3 dryout, uncovery and reflood times for the same discharge breaks.

Al though the dry out and uncovery times are about the same for the same breaks the reflood time is significantly later for BF-1 (some 7x7 fuel) than for 6F-3 (all 8xd fuel). The other important effect of fuel type relates to the differences in the amount of stored energy which can De removed by the time of boiling transition (loss of good heat transfer from fuel rod to coolant). For 7x7 fuel it takes approximately 25 seconds to re-move the stored heat via nucleate boiling while it takes only about half this time for 8x8 fuels. Accordingly, the PCT of cores with 7x7 fuel are more 1553 126

I I

! sensitive to differences in boiling transition times associated with different break sizes and break locations. More heat will still be stored in the 7x7 fuel to heat up its clauding than in the bxB fuel to heat up its claading once nucleate boiling heat transfer is lost. That is although the dryout times of the fuel in the BF-1 (7x7) and BF-3 (8x8) cores are about the same for a given break in the spectrum of pipe creaks, the cladding temperature at the time of hot node uncovery is substantially higher for the Unit 1 fuel than for the Unit 3 fuel due to the greater stored energy still containea in the 7x7 fuel at fuel dryout.

In summary, therefore, the combinea thermal and hydraulic effcts of fuel type on vessel reflood and cladding heat-up phenomenon results in a shif t of the worst break loc. tion and size from the largest suction line creak (BF-1 and BF-2) to an intermeaiate size break in the discharge line (BF-3).

The PCT for the limiting break size and location was conservatively calculated with the addea conservatism of applying the 80% discharge break LAMd-SCAT Code (earlier boiling transition time) results to the 66% discharge Dreak SAFE /REFLOOD Coce (uncovery time) results. Thus any slight non-conservatism (at most 2*F to 5*F) due to the possibility that the PCT occurs slightly above or below the limiting break size is more than compensated for by this unrequired (extra) ccnservatism in the CHASTE (heatup) analysis. The CHASTE (fuel cladding heatup) reanalysis was performed for each of the initial and reload fuel types assuming the same respective tables of Maximum Average Planar Linear Heat Generation Rates (MAPLHGR) vs. Average Planar Exposure as those used in connection with the previously accepted ECCS conformance analysis (29) performed for BF-3 with LPCI loop-selection-logic. Accorcingly, since ECCS performance is improved relative to the formerly limiting suction line break, the overall peak ladding temperature for the worst break location, break size, single tpilyre, fuel type and exposure has been lowereo.

Formerly the licensee reporteal29)a PCT of 1963*F for BF-3 with LPCI goop sclection logic.

With the LPCI modification the licensee now reports ( ) a PCT of 1790*F. Addi-tionally, operation of BF-3 at these MAF.rtCR values results in a local claading oxidation of less than it and a core wide metal-water reaction of.05; for the limiting break size with LPCI injection valve failure (i.e. the DBA). These calculated values also meet the requirements specified in 10CFRbo.40.

With regard to small break consequences, the licensee states (fd that the generic results reported in Reference 41 are applicable to BF-3.

The bounding analysis referenced provides the PCT for the worst size small Dreak occurring in the recirculation discharge piping of a BWR/4 with LPCI modifi-cations. The analysis assumes a direct current power source failure (worst 1553 127

O single failure for a small break). For this assumed failure, Reference 41 indicates that one LPCI pump, one of the two core spray systems (i.e. two 50% capacity pumps) and the automatic depressurization system (AUS) are available to mitigate the accident. Tae generic analysis shows that PCT will be less than 2200* even when taking credit for only four of the six ADS valves.

The effects of a DC power source failure on the consequences of small and large breaks as reported in Reference 41 are also being generically reviewed by the staf f.

Although we have not yet completed our review of Reference 41, based on the systems stated to be available with a DC power source f ailure, we Delieve that there will not be changes to the generic stucy which coulo make the results of a plant-specific small break LOCA become more limiting than the worst large break LOCA.

3.P.3 Overall Evaluation of LPCI Modification We have reviewed the analysis of emergency core cooling system performance submitted by TVA for BF-3 with the proposed LPCI-modifications and cenclude that all of the requirements of 10 CFR 50.46 and Appendix K to 10CFRbu.40 will be met when the reactor is operated in accordance with the MAPLNbR versus Average Planar Exposure values given in Tables 3.1.5-1 through 3.5.1-3 of Reference 3.

3.3 Modification of RWCU System Piping As discussed in Section 1.3 of this safety evaluation, the reactor water cleanup (RWCU) System piping was modified so that the return flow is distribu-ted equally among the feedwater lines. This modification, which has been reconmended by URC and GE, will allow the higher temperature RWCU return water to be mixed at low flow rates with the large volume of feedwater, thereby lessening the thermal cycling on the feedwater nozzle and the consequent thermal f atigue. The modification required the addition of a check valve.

The change to the Technical Specifications is to add this check valve to Table 3. /.G. " Check Valves on Drywell Influent Lines" as one of the pene-tration and isolation valves which must be inc luded in the containment leak t es t Londuct ed each operal inq cycle.

In partial response to the USNitC Of f ire of in. pes Lion and Inf on ement's Hulletin /9-OH, during this outage IVA added new addi tional hydrogen and oxygen sensing lines into the primary containiiient.

These lines are isolatable by the usual inboard and outboard isolation valves and outboard block valves. Since these valves must also be included in the periodic containment leak test, they were added to Table 3.7.D " Primary Con-tainment Isolation Valves". We conclude that these plant modifications improve plant safety and that the proposed changes to the Technical Specifications are appropriate and acceptable.

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3.4 Replacement of Two Safety-Relief Valves Raising the lift settings of the subject dual action SRVs affects those plant transients which result in an increase in reactor system pressure sufficient to cause safety / relief valve actuation of the highest pressure setpoint Accordingly, the licensee has reanlayzed the most severe pressuriza-groups.

The limiting events for Browns Ferry Unit 3 are generator tion transients.

load rejection with bypass system failure (LR w/o BP) and main steam isolatiort valve (MSIV) closure with indirect high flux scram (vessel overpressure protection analysis).

3.4.1 Abnor:ral Operation Transients For BF-3 the largest change in bundle critical power ratio (CPR) for the retrofit 8x8R fuel types is caused by the load rejection without bypass pressurization event. This event, which is initiated by fast closure of the turbine control valves, causes a rapid collapse of moderator voids in the core lhe collapse of the voids causes a significant addition of positive reactivity to the core, which results in a pronounced neutron flux spike, and a subsequent rise in core heat 11ux.

Before core heat flux can rise sub-stantially, the event is terminated by a reactor scra'n and prompt recircula-tion pump trip caused by a fast closure trip signals developed at the turbine control valves.

The licensee reanalyzed this event using methods which are the same as those used for the most recent BF-3 reload safety analysis.(3) For the revised thermal margin analyses, in addition to the assumed 25 psi increase in valve lif t pressure, the safety / relief valves capacities of the two Crosby valves were modeled to reflect the somewhat lesser steam relief rate of these valves compared to the two Target Rock valves they will replace.

The reanalyses (

shows that the proposed change in SRV setpoint would not result in a significant increase in the critical p er ratio for the lwr /oBP event when compared wi th the most recent analyses \\p1 Accordingly.

the staff f inds it acceptable to retain the present operating limit minimum CPRs and that the proposed SRV selpoint change is acceptable with regard to fuel thermai maryin considerations.

The licensee also reanalyzed the load rejection without bypass event from the viewpoint of peak transient reactor system pressure. For calculating peak pressure the plant transient analysis used the same models and methods as for the fuel thermal margin analyses, including credit for the prompt recirculation pump trip feature. The results show that a peak transient pressure increases by approximately 12 psi with the proposed change setpoint However, the reanalysis shows that a margin greater than 25 to the change.

This is lif t pressure of code safety valves (1250 psig) is still available.

acceptable.

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. 3.4.2 Overpressurization Analys_is The licensee has also provided the results of a bounding overpressurization analysis, to demonstrate t hat an adequate margin exists to the ASME Code allowable pressure, with the proposed revised SRV settings. The ASME Code allows peak transient pressures up to 100% of vessel design pressure, i.e.,

1375 psig. The most limiting event was taken to be the closure of all main steam isolation valves with a reactor trip on high neutron flux which was the same event analyzed for the most recent reload. The analysis conservatively assumed an initial reactor power of 104.5% and 100% core flow, an end-of-cycle scram reactivity insertion rate curve and all safety / relief valves operative. As for the load rejection without bypass, the Crosby valve char-acteristics modeled reflected the lower relief capacities and the higher opening pressure setpoint of these valves. The reanlaysis included a (ATWS) recirculation pump trip on high reactor pressure since the attendant flow reduction has the effect of increasing peak transient pressure. The results show that the substitution of the two Crosby valves for the Target-Rock valves increases the peak pressure at the bottom of the reactor vessel by approximately 13 psig leaving a margin of 82 psi to g l375 psig Code allow-able safety limit. Furthermore, a generic analysis, showing the sen-sitivity of peak transient pressure to total relief capacity, when applied to Browns Ferry 3, shows that the failure-to-open of one SRV would cause pressure to increase by less than 20 psi. Therefore, the maximum transient reactor vessel pressure for MSIV closure at end-of-cycle assuming an indirect high neutron flux scram and one failed safety valve will still show ample margin to the pressure safety limit. These results are acceptable to the staff.

4.0 Environmental Considerations We have determined that this 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 this amendment involves 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 pre-pared in connection with the issuance of this amendment.

5.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequenc es of accidents previously considered and does not involve a significans 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 con-ducted 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:

November 30, 1979

I l

- ?3 -

6.0 References 1.

Tennessee Valley Authority letter (L. Mills) to USNRC (H. Denton) dated August 6, 1979.

2.

Tennessee Valley Authority letter (L. Mills) to USNrtC (H. Denton) dated October 25, 1979.

3.

" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Power Station Unit 3, Reload 2",NED0-24199 dated June 1979.

4.

" Proposed Changes to Browns Ferry Unit 3 Technical Specifications" sub-mitted as Enclosure 1 to TVA letter (L. Mills) to USNRC (H. Denton) dated August 6,1979.

h.

Tennessee Valley Authority letter (L. Mills) to llSNRC (ll. Denton) dated

' ep tember Pri, 19 /9.

6.

" Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3",

NED0-24194A dated July 1979.

7.

" Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, NUREG-0313, dated July 1977 and Revision 1 to NUREG-0313 dated October 1979.

8.

" Investigation and Evaluation of Stress-Corrosion Cracking in Piping of LWR Plants", NUREG-0531 dated February 1979.

9.

General Electric Company Service Information Letter No. 208, Revision 1,

" Minimizing Feedwater Nozzle Thermal Duty", dated October 1978.

10.

Tennessee Valley Authority letter (L. M. Mills) to USNRC (H. R. Denton) dated September 26, 1979.

11.

II',NRC Of fice of Inspect ion and Enforcement, IE Circular No. 79-18 " Proper Im.t.illation of larip t Roc k Safety-Relief Valves.

12.

USNRC Office of Inspection and Enforcement, IE Bulletin 74-4 and 74-4a,

" Malfunction of Target Rock Safety Relief Valves" issued March 2,1974.

13. Tennessee Valley Authority letter (L. M. Mills) to USNRC (H. R. Denton) dated October 10, 1979.

14.

Revised Appendix A, " Analysis For Alternate Safety-Relief Valves" dated September 25, 1979 to NED0-24199, " Browns Ferry Unit 3 Reload 2 Supple-mental Licensing Submittal" dated June 1979.

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l 15.

" Generic Reload Fuel Application," NEDE-240ll-P-A, dated August 1976 incluuing Appendix 0 dated August 1976.

16.

USNRC letter (D. Eisenhut) to General Electric (R. Gridley) datedflay 12, 1976 transmitting " Safety Evaluation for the General Electric Topical Report Generic Reload Fuel Application (NEDE-2440ll-P) dated April 1976.

17.

UShRC letter (T. Ippolito) to General Electric (R. Grialey) datea April 1979 transmitting " Safety Evaluation Supplement f or an Amencuent cated August 1978 to the beneral Electric Topical Report Generic Reload Fuel Application."

18.

" Status Report on the Licensing Topical Report" ' General Electric Boiling Water Reactor Generic Reload Application for Oxd Fuel' t&EDO-20360 Revision 1 and Supplement 1 by Division of Technical Review, ONRR, USNRC, April 1975.

19 "beneral Electric Boiling Water Reactor beneric Reload Appli-cation for ex6 Fuel." Revision 1 and Supplement i April 19/4, NLUO-20Jbu.

20.

" General Electric BWR Thermal Analysis Basis (GETAB) Date, Correlation and Design Application," NEDO-10958, November 1973.

21.

USNRC letter (T. Ippolito) to TVA (H. Parris) dated Septemoer 19, 1979.

22.

General Electric letter (R. Engle) to USNRC (0. Eisenhut and R. Tedesco), dated March 30, 1979.

23.

General Electric letter (R. Gridley) to USNRC (D. Eisenhut dnd D. Ross), datec Uctober 5,1970, transmitting " General Electric Information NEDE-24131, Basis for ex6 Retrofit Fuel Thermal Analysis Application."

24.

" Analytical Methods of Plant Transient Evaluations for tra General Electric uoiling Water Reactor," uEDO-luoV2, Feoruary 1973.

25.

"Iransient and Staoili ty Tests at Peach Bottom atoiiiic Power Station Unit 2 at End of Cycle 2," Carmichael, L. A. and Niemi, R. 0., EPRI-uP-564, June 1976.

26.

General Electric letter (R. Engel) to USNRC dated July 11, 1977.

27.

General Electric letter (E. Fuller) to USNRC datea October 25, 1977.

28.

" Input of One Dimensional Transient Model on Plant Operating Limits," enclosure of letter, E. D. Fuller (GE) to USkRC dated June 26, 197o.

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- 29.

" Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3," NEDO-24127, June 1978.

30.

" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with IUCFR50 Appendix K" NED0-20So6 submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to the SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyory (GE) to Victor Stello, Jr., (NRC) dated December 20, 1974.

Safety Evaluation for General Electric ECCS Evaluation Model Modifi-31.

G. Sherwood cation transmittea by letter from K. R. Goller (NRC) to 6.

(GE), dated April 12, 1977.

32.

General Electric letter (A. Levine) to USNRC (D. Ross) dated 27, 1977 transmitting General Electric (GE) Loss-of-Coolant J anuary Accident (LOCA) Analysis Mooel Revisions-Core Heatup Code CHASTEOS.

33.

General Electric letter (A. Levine) to USNRC (D. Vassallo), dated March 14, 1977 transmitting Request for Approval for Use of Loss-of-Coolant Accident (LOCA) Evaluations Mocel Code REFL00005, 34.

Supplemental Information for Plant Modification to Eliminate Sig-nificant In-Core Vibrations, Supplement 1, NEDE-21156-1, September 1976.

35.

Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations, Supplement 2, NEDE-21156-2, January 1977.

36.

General Electric letter (R. Engel) to USNRC (V. Stello), Answers to NRC Ouestions on NEDE-21156-2 J anuary 24, 1977.

3/.

USNRC letter (R. Reid) to PASNY (G. Berry) dated September lo,1977 transmitting Safety Evaluation by the Of fice of Nuclear Reactor Regulation Supporting Amenament No. 30 to Facility Operating License No. LPR-59.

38.

PASNY letter (G. Berry) to USNRC (R. Reid) datea July 29, 1977 trans-mitting Loss-of-Coolant Accident Analysis Report for James A. Fitzpatrick Nuclear Power Plant (Lead Plant), NED0-21662-2, July 1977.

" Loss of Coolant Accident Analysis for Browns Ferry Nuclear Plant 39.

Uni t 2" NED0-24068-1, February 1978.

40.

" Loss of Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 1" NED0-24056, September 1977.

41.

General Electric Letter (R. Engel) to USNRC (P. Check), "UC Power Source Failure for BWR/3 and 4," dateo November 1, 19/d.

42.

GE letter ((1. Stuart) to USNRC (V Stello) " Code Overpressure l'rotection Analysis - Sensitivity of Peak Vessel Pressure to Valve Operability," dated December 23, 1975.

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