ML20028G368

Safety Review of Browns Ferry Nuclear Plant Unit 1 at Core Flow Conditions Above Rated Flow During Cycle 5.

ML20028G368
Person / Time
Site:	Browns Ferry
Issue date:	05/31/1982
From:	Fischer D, Gridley R GENERAL ELECTRIC CO.
To:
Shared Package
ML20028G366	List: ML20028G364 ML20028G367 ML20028G368
References
82NED056, 82NED56, DRF-L12-00515, DRF-L12-515, NEDO-22135, NUDOCS 8302090006
Download: ML20028G368 (31)

Text

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SAFETY REVIEW OF BROWNS FERRY  ;

NUCLEAR PLANT UNIT NO.1 AT CORE FLOW CONDITIONS ABOVE.

RATED FLOW DURING CYCLE 5

NED0-22135 DRF L12-00515 82NED056 Class I May 1982 SAFETY REVIEW OF BROWNS FERRY NUCLEAR PLANT UNIT NO. 1 AT CORE FLOW CONDITIONS ABOVE RATED FLOW DURING CYCLE 5 s

Approved: Approved: k' 24 "k1 D. L. Fischer, Manager R. . Gridley Manager Core Nuclear Design Fuel and Services Licensing Safe'ty and Licensing Operation l

l NUCLEAR POWER SYS1;MS DIVISION

• GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENERAL $ ELECTRIC

NEDO-22135 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending TVA's operating license of the Browns Ferry Nuclear Plant Unit 1. The information contained-in this report is believed by General Elec-tric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract between the Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear. Plant Units 1 and 2, dated June 17, 1966, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this doc'tment makes any representation or warranty (express or implied) as to the complet'eness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

NEDO-22135 CONTENTS fage ABSTRACT vii

1. INTRODUCTION AND

SUMMARY

l-1

2. SAFETY ANALYSIS 2-1 2.1 f.Laormal Operational Transients 2-1 2.2 LOCA Analysis 2-1 2.3 Rod Drop Accident 2-2 2.4 Rod Withdrawal Error 2-2

3. REACTOR INTERNALS PRESSURE DROP 3-1 3.1 Reactor Internals 3-1 3.2 Fuel Channels 3-1 3.3 Fuel Bundles 3-1

4. FLOW-INDUCED VIBRATION 4-1

5. . THERMAL-HYDRAULIC STABILITY ANALYSIS 5-1

6. OPERATING LIMITATIONS 6-1 6.1 Operating Map 6-1 6.2 K Factor 6-1 6.3 HkghFlowRodBlock 6-1

7. REFERENCES 7-1 l

iii/iv

_ _ _ . - . . _ . _ _ ~

NEM -22135 TABLES Table Page 2-1 Core-Wide Transient Analysis Results 2-3 2-2 Overpressurization Analysis 2-4 2-3 Summary of Local Rod Withdrawal Error Transient with l Limiting Instrument Failure 2-5 l

i 6-1 MCPR Operating Limits at Increased Core Flow for Browns Ferry Unit 1, Cycle 5 6-2 4

j ILLUSTRATIONS Figure Page

)

1-1 Operating Map 1-2 1

2-1 Generator Load Rejection without Bypass (104.5% Power, 105% Flow) 2-6 2-2 Feedwater Controller Failure (104.5% Power, 105% Flow) 2-7 2-3 MSIV Closure, Flux Scram (104.5% Power, 105% Flow) 2-8 2-4 Limiting Rod Pattern (105% Core Flow) 2-9 4

a j v/vi

NEDO-22135 ABSTRACT A safety evaluation has been performed to show that Browns Ferry Nuclear Plant, Unit 1 can increase core flow to opente within the region of the ope m ting map bounded by the line.between 100% power, 200% core flou (200,100) and 100% power,10S% core flow (100,105).

The plant can then continue to operata in the region of the oper-ating map bounded by the constant recirculation pump speed line between 100% power,10S% flow, (200,105), and 50% power,113% flou (50,113).

The minimum critical power m tio (MCPR) ope m ting limits increase from the values established in the Reload 4, Cycle S licensing submittat and are identified in Table 6-1.

vil/viii

i NEDO-22135 l

1. INTRODUCTION AND

SUMMARY

,i l This report presents the results of a safety evaluation for operation of the Browns Ferry Nuclear Plant Unit I with increased core flow during cycle 5.

f i The core flow can be increased co operate the plant within the region of the

] operating map bounded by the line between 100% power, 100% core flow (100,100) and 100% power, 105% core flow (100,105). The plant can then continue to operate in the region of the operating map bounded by the conr. tant recircula- i tion pump speed line between 100% power, 105% core flow (100,105) and 50%

power,113% core flow (50,113). The region of operation with increased core flow is bounded by BCDE on the operating map in Figure 1-1.

j In order to justify operation with increased core flow (ICF), the limiting abnormal operational transients reported in Reference 1 for rated flow opera-4 tion were reanalyzed for increased core flow operation. The loss-of-coolant-accident (LOCA), rod drop accident, and rod withdrawal error event were also reevaluated for increased core flow operation.

l The effect of the increased pressure differences (due to the increased core flow) on the reactor internal components, fuel channels, and fuel-bundles were also evaluated to show that the design limits will not be exceeded. The effect of the increased flow rate on the induced vibration responsa of the reactor internals was also evaluated to ensure the response was within accept-able limits. The thermal-hydraulic stability was evaluated for increased core flow operation.

i i

The limits for plant operation are given in Table 6-1.

i i

1-1 i

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NEDO-22135 110 POINT POWER /F LOW 8 100/100 8 C 10rV106 0 100 -

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NOTE: LETTERS AND NUMBERS f ARE SPECIFIC POINTS / CONSTANT REF ERENCED IN THE TEXT /

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I RATED FLOW CONTROL 80 - LINE (CONSTANT CONTROL ROD POSITION) /

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Figure 1-1. Operating Map I

l-2

.. - . - = . - - ._ - . - - - - - ._ - _- - . .- . ___ - ..

l l

NEDO-22135 l 1

i

! 2. SAFETY ANALYSIS t

l 2.1 ABNORMAL OPERATIONAL TRANSIENTS l I

The limiting abnormal operational transients analyzed in the Reload-4, Cycle 5 i reload licensing submittal were reanalyzed for increased core flow.

i.

j Nuclear data was used to analyze the load rejection without bypass and feed- i water controller failure events at the (104.5,105) condition. The results f

of the transient analysis for these events are presented in Table 2-1 along with the transient results contained in the Reload-4 licensing submittal l

(Reference 1). The transient responses are presented in Figures 2-1 and 2-2.

Increasing the core flow from 105% to 113% of rated along the constant pump

]

speed line as power decreases (line CD in Figure 1-1) will result in lower

) void fractions, and hence decreased void reactivity effect. The degradation f in the scram reactivity will be more than compensated for by the decreased j void reactivity and the reduced steam flow. The net result is that-.the limit-t

! ing abnormal operational transients become less severe wher. the plant coasts i down in power along the constant pump speed line from (100,105) to (50,113)

(Reference 2). Therefore, it is concluded that the operating limits given in Table 6-1 are also bounding for coastdown operation along the constant pump l speed line from (100,105) to (50,113).

i l

The limiting transient for overpressurization analysis, main steam isolation valve (MSIV) closure with flux scram, was reanalyzed at 104.5% power and 105%

} flow. The results are within prescribed limits and are presented in Table 2-2 and in Figure 2-3.

i

, 2.2 LOCA ANALYSIS The LOCA calculations performed for increased core flow operation were done ,

for a similar plant and have been found to be applicable to Browns Ferry, ,

j Unit 1. The effect of increased core flow on LOCA analyses is not significant because the parameters which most strongly affect the calculated peak cladding  ;

2-1

, - . ~ , - , , , - , _ - . , - - - ~ , , - - ,.

l NEDO-22135 temperature (PCT), i.e. , high power node boiling transition time, and core reflooding time, have been shown to be relatively insensitive to increased core flow.

A LOCA analysis at 105% rated steam flow and 105% rated core flow was performed over the larbe break region for Browns Ferry 1 to confirm that previous analyses performed at 105% rated steam flow and 100% rated core flow are con-servative. The SAFE and REFLOOD codes were used to quantify the change in uncovery time and reflooding time. The change that occurs is a shift in the worst break size for 8x8 fuel from 66% to %70% of the full-sized discharge break. The hot node total uncovery time is less for 105% core flow compared with 100% core flow, and the time of hot node uncovery is delayed; therefore, peak cladding temperatures for the worst large break for the increased core flow case are N30*F lower than the 100% core flow case.

Therefore, it is concluded that the LOCA analysis and maximum average planar linear heat generation rates (MAPLHGRs) determine ~d for the Browns Ferry Unit I core (Reference 3) are unchanged for use in the increased core flow region of i

the operating map.

2.3 ROD DROP ACCIDENT This event is a startup accident which is most severe in a nonvoided, xenon free core. The results reported in the referenced licensing submittal are for cold and hot standby conditions at beginning of cycle and those results would not be affected by increased core flow.

The effect of increased core flow is to change the conditions for the end of cycle (EOC) analysis. These changes have been evaluated and, as in the case of the EOC analyses at rated flow, the results are shown to be bounded by the results reported in the licensing submittal, Reference 1.

2.4 ROD WITHDRAWAL ERROR The results of the rod withdrawal error analysis are presented in Table 2-3.

The limiting RWE rod pattern is shown in Figure 2-4.

2-2

I Table 2-1 i CORE-WIDE TRANSIENT ANALYSIS RESULTS i

i .

l Unadjusted ACPR l Power Flow $ Q/A

! . Transient Exposure (%)* (%)* (%)* (%)* 8x8 8x8R P8x8R Plant Responsa f

l LR w/o BR EOC 104.5 100 613 122 0.19 0.19 0.21 (Reference 1)

(Licensing l Submittal)

LR w/o BP EOC 104.5 105 659 124 0.21 0.21 0.22 Figure 2-1 FWCF EOC 104.5 100 389 121 0.16 0.16 0.17 (Reference 1)

(Licensing 2 -

y Submittal) h a e FWCF E0C 104.5 105 393 122 0.16 0.16 0.18 Figure 2-2

• Percent of reference t

}

4

NED0-22135 Table 2-2 OVERPRESSURIZATION ANALYSIS

! Power Core Flow s1 v Plant

! Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1238 1272 Reference 1

j. (Flux Scram)

MSIV Closure 104.5 105 1243 1279 Figure 2-3 (Flux Scram) i l

l l

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l 2-4

NEDO-22135 Table 2-3

SUMMARY

OF LOCAL ROD WITHDRAWAL ERROR TRANSIENT WITH LIMITING INSTRUMENT FAILURE ACPR LHGR*

Rod Block Rod Position Reading (ft withdrawn) 8x8 8x8R/P8x8R 8x8 8x8R/P8x8R 104 3.0 0.14 0.08 ,10.3 13.7 105 3.5 0.17 0.10 11.5 13.7 106** 4.0 0.19 0.13 12.4 14.7 107 4.5 0.22 0.15 12.8 15.4 108 4.5 0.27 0.15 12.8 15.4 109 5.0 0.30 0.17 13.1 15.9 109.4*** 5.0 0.31 0.17 13.1 15.9 110 5.5 0.33 0.20 13.3 16.4 i

• Includes a 2.2% peaking penalty for fuel densification

• • Current rod block monitor setpofnt

• • • Rod block signal corresponding to a 106% setpoint at 105% flow and standard rod block monitor instrumentation

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NED0-22135

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NEDO-22135 Notes: 1. Number indicates number of notches withdrawn out of 48.

Blank is a withdrawn rod.

2. Error rod is (26,43) .

f 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 0 0 0 0 55 28 32 24 32 28 51 0 4 8 8 4 0 47 24 36 44 36 24 43 0 0 8 0 0 8 0 0 39 28 32 44 36 44 32 28 35 0 0 4 4 4 4 0 0 31 24 28 24 32 24 28 24 27 0 0 4 4 4 4 0 0 23 28 32 44 36 44 32 28 19 0 0 8 0 0 8 0 0 15 24 36 44 36 24 11 0 4 8 8 4 0 I 7 28 32 24 32 28 3 0 0 0 0

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Figure 2-4. Limiting Rod Pattern (105% Core Flow) 2-9/2-10

NED0-22135

3. REACTOR INTERNALS PRESSURE DROP For Browns Ferry Unit 2, reactor internals pressure differences have been calculated for the increased core flow condition and evaluated against allow-able limits (Reference 4). The evaluation included consideration of upset, amergency, and faulted conditions, in addition to conditions during normal operation. Because of the similar structural design of Browns Ferry Units 1 and 2, the results for Unit 2 have been evaluated and found to be bounding

. for Unit 1.

1 I

3.1 REACTOR INTERNALS The reactor internals which would be most affected at the increased core flow conditions are the core plate, guide tube, shroud support, shroud, and top guide. These components were evaluated under normal, upset, emergency, and faulted conditions. The pressure differentials for these components during increased core flow operation were found to produce stresses that are within the allowable limits.

3.2 FUEL CHANNELS The fuel channels were also evaluated under normal, upset, emergency and faulted conditions for increased core flow. The channel wall pressure differ-entials were found to be within the design allowable values (Reference 5).

3.3 FUEL BUNDLES The margin to fuel bundle lift was reevaluated for increased core flow opera-tion. The analysis considered the added bundle lif t component due to increased h core flow in addition to the effect of the design basis LOCA, the control rod friction force due to scram, and the design basis earthquake. The fuel bundle i lift margin is 132 lb during the worst-case faulted event from operating con-ditions of 104.5% power, 105% flow. Thus, the effect of increased core flow is clearly acceptable in terms of avoiding fuel bundle lift.

3-1/3-2

NED0-22135

4. FLOW-1NDUCED VIBRATION To ensure that the induced vibration response of the reactor internals is within the design expectations, a single reactor of each product line and size undergoes an extensive vibration test during startup. All other reactors of the same product line and size undergo a less rigorous confirmatory test to assure similarity to the base test.

The extensive vibration testing was performed on Browns Ferry Unit 1. Based on a review of the test data, the reactor internals response to flow-induced vibrations is expected to be within acceptable limits for plant operation in the increased core flow (ICF) region indicated in Figure 1-1.

)

i 4-1/4-2

I . .

NEDO-22135

5. THERMAL-HYDRAULIC STABILITY ANALYSIS The results of this analysis for a similar plant at increased core flow condi-tions give decay ratios for both the reactor core stability and the channel hydrodynamic stability that are less than those calculated for the normal operating conditions.

The reactor core stability and the channel hydrodynamic stability decay ratios reported in the Reload-4 licensing submittal are therefore bounding for in-  !

creased core flow operation during cycle 5.

1 i

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5-1/5-2

NEDO-22135

6. OPERATING LIMITATIONS The MCPR operating limits established for operation in the increased core flow region of the operating map (Figure 1-1) are summarized in Table 6-1 along with the limits previously established in the existing Reinad 4 licensing basis (Reference 1) .

All other operating limits previously established for the reload licensing basis have been shown to be bounding.

Following are restrictions or limitations unique to increased core flow operation.

6.1 OPERATING MAP The increased core flow reactor internal pressure differences and fuel bundle lift calculations were analyzed, and are applicable only for reactor operation within the region bounded by BCDE on the power flow map (Figure 1-1).

6.2 K FACTOR g

For core flows greater than or equal to rated core flow, the Kg factor is equal to 1.0.

6.3 HIGH FLOW ROD BLOCK 3

Current technical specifications call for a rod block at 110% drive flow.

For operational convenience, the rod block may be changed to 115% drive flow.

6-1

1 NED0-22135 Table 6-1 MCPR OPERATING LIMITS AT INCREASED CORE FLOW FOR BROWNS FERRY UNIT 1, CYCLE 5 Pressurization Events Option A Option B Exposure Range: BOC TO EOC 8x8 8x8R P8x8R 8x8 8x8R P8x8R Generator Load Rejection 1.34 1.34 1.35 1.24 1.24 1.25 without Bypass (1. 32) * (1. 32) (1.34) (1. 23) (1.23) (1.24)

Feedwater Controller Failure 1.28 1.28 1.30 1.25 1.25 1.27 (1. 28) (1. 28) (1.29) (1. 25) (1.25) (1.26)

Nonpressurization Events 8x8 8x8R P8x8R Loss of Feedwater Heating (1.22) (1.22) (1.22)

Fuel Loading Error (1.22) (1.22) (1.22)

Rod Withdrawal Error 1.38** 1.24 1.24 (1.26) (1.21) (1.21)

• Values in parentheses are the previously established limit at rated core flow from Reference 1.

• • This value is for normal flow-biased rod block instrumentation corresponding to a setpoint of 106% rated power at 105% of rated flow. Instrumentation signal clipped for a setpoint of 106% would result in a MCPR operating limit of 1.26 for the rod withdrawal error.

(

6-2

NEDO-22135

7. REFERENCES

1. " Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload No. 4 (Cycle 5)," Ceneral Electric Company, March 1981 (Y1003J01A19).

2. R. E. Engel (GE), letter to T. A. Ippolito (NRC), "End of Cycle Coastdown

Analyzed with ODYN/TASC," September 1, 1981.

3. " Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear !1 ant, Unit 1,"

General Electric Company, September 1977 (NED0-24056).

i

4. " Safety Review of Browns Ferry Nuclear Plant Unit No. 2 at Core Flow Conditions Above Rated Flow at the End of Cycle 4," General Electric Company, March 1982 (NEDO-22096).

5. "BWR Fuel Channel Mechanical Design and Deflection," General Electric Company, September 1976 (NEDO-21354) .

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