ML20083E369

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Forwards Revised FSAR Pages as Addl Info Re Changes Being Made to Steam Generators to Minimize Tube Vibration to Resolve SER Outstanding Item 10.Mods Acceptable
ML20083E369
Person / Time
Site: Byron, Braidwood, 05000000
Issue date: 12/16/1983
From: Tramm T
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
7379N, NUDOCS 8312290174
Download: ML20083E369 (27)
v • d • e


Text

-

Commonwealth Edison One First National Plaza. Chicago. Ilknos 7 Address Reply to Post Office Box 767 Chicago. tilinois 60690 December 16, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 Steam Generator Tube Vibration NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457 References (a):

February 9, 1983 letter from T. R.

Tramm to H. R. Denton.

(b):

April 12, 1983 letter from T. R.

Tramm to H.

R. Denton.

(c):

July 18, 1983 NRC Summary of Meeting on July 7, 1983 with the Technical Review Committee.

(d):

July 18, 1983 letter from L. D.

Butterfield, Jr. to D.

G.

Eisenhut.

(e):

August 1, 1983, letter from L. D.

Butterfield, Jr. to D.

G.

Eisenhut.

Dear Mr. Denton:

This letter provides additional information regarding the changes being made to the Byron and Braidwood steam generators to minimize tube vibration.

NRC review of this information should enable closure of Outstanding Item 10 of the Byron SER.

As described in previous generic correspondence meetings and hearings, selected steam generator tubes are being expanded at two tube support plates and 10% of the main feedwater flow is being diverted to the auxiliary feedwater nozzle.

Extensive reviews and analyses of these modifications have been performed by Westinghouse and by the Counterflow Steam Generator Owners Review Group.

These efforts have verified that the modifications will be ef fective in reducing tube vibration to accept-able levels and that the modifications will introduce no unacceptable safety consequences during normal or transient operating conditions.

8312290174 031216 PDR ADOCK 05000454

\\

r E

PDR 0

H. R. Denton December 16, 1983 l

l This letter provides the plant-specific information, details of this modification and the results of plant-specific safety analyses on the split-feed arrangement.

Tables 2.5-1 and 2.5-2 of references (d) and (e) list the FSAR sections and design transients which have been reviewed.

Special attention has been given to the prevention of esterhammer and to the impact of these changes upon plant operating procedures.

We have concluded that the Byron and Braidwood plants can be operated safely with the modified steam generators and feedwater piping.

Attachment A to this letter contains revised FSAR pages.

These contain the necessary revisions to the facility description and accident analyses.

These pages will be incorporated into the FSAR at the earliest opportunity.

As indicated in references (d) and (e), inservice inspection plans and tube plugging criteria are also being addressed on a plant-specfic basis.

For Byron and Braldwood, the Technical Specifications already proposed adequately cover these issues.

The extensive inservice inspection program already agreed upon provides for the early detection of unanticipated problems with steam generator tubing.

The 40% plugging criteria appears adequate to prevent mid-cycle failure of any tubes which are found to experience minor degradation.

To provide additional assur-ance uf the adequacy of the tube modifications, vibration measurements are to be made on selected tubes during operation of one of the first domestic units with expanded tubes.

If these measurements indicate the need for additional inservice inspection, ISI changes can be easily incorporated into individual plant Technical Specifications such as Byron's.

Please address further questions regarding this matter to this One signed original and fifteen copies of this letter and the enclosures are provided for NRC review.

Very truly yours, 0/W T. R.

Tramm Nuclear Licensing Administrator 1m Enclosure 7379N

ATTACHMENT A FSAR-Revisions to Implement 90/10 Feedwater Flow Split 1.

Table 3.6-12 Corrections 2.

Table 3.9-16 Revisions 3.

Table 4.1-1 Revisions and Corrections 4.

Table 4.4-1 Revisions and Corrections 5.

Figure 4.4-9 Revisions 6.

Page 5.1-4 Revisions 7.

Table 5.1-1 Revisions 8.

Figure 5.1-2 Revisions 9.

Section 5.4.2.5.3 Replacement Section 10.

Section 5.h.2.5.4 New Section 11.

Table 6.2-58 Revisions 12.

Section 10.4.7.3 Replacement Section

'13.

Section 15.0.3.2 Revisions 14.

Figure 3.6-2 Revisions 15.

Figure 10.4-1 Revisions 7379N

l TABLE 3.6-12 CALCULATED STRESSES FOR POSTULATED BREAK POINTS (For ASME Sec. III Class 2&3 and ANSI B31.1 Piping Systems)

PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS 0.8 (l.2S +S )

b A

LINE NUMBER (S)

ID (psi)

(psi)

FE?DWATER Loop 1 1FWO3DA-16" B20A 17660 32400.

1FWO3DA-16" B20B 16832 32400.

1FWO3DA-16" B40A 13586 32400 s

?

IFWO3DA-16" B65A 19150 32400

?

m 1Fw03DA-16" B65B 20934 32400 m

1FWO3DA-16" B80 15907 32400

c FEEDWATER Loop 2 1FWO3DB-16" BSA 10706.

32400 1FWO3DB-16" B30A 17847.

32400 1FWO3DB-16" B30B 17293.

32400 1FWO3DB-16" BSSA 12908.

32400 1FWO3DB-16" B85A 16971.

32400 1FWO3DB-16" B85B 18565.

32400 1FWO3DB-16" B100 14974.

32400

7...

-5559

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U/B-FSAR TADL1: 3.9-16 (Cont'd)

ACTUATED SIZE BODY QUALITY ISI TAG NUMBER BY (in.)

TYPE GROUP CATEGORY P&ID 3

Check A

C M-64-5 ICV 8378A,8 3

Check A

C M-64-5 1CV9379A,B 3

Check B

C M-64-5 1CV8381 2

Check Il C

M-64-4 1CVB442 4

Chuck B

C M-64-3 1CV8481A,B 8

Check n

C M-64-4 1CV8546 1CV8804A Motor 8

Gato it il M-f> 4 -4 4

Plug u

A M-f. 3 - 1 1FC009 4

l'l uy 16 A

M -(. 3 - 1 IPC010 IFP010 Air 4

Unto u

H M-S2-1 IFPoll Air 4

Gatu in H

M 's2-1 1rWOO9A-D llyd r au l t r2 16 Gata H

14 M - 3 f. - l l

.7$

O l uisu H

H M - S t. - l 1rW0lLA-D

. _ _ _ _ _ 1FWO35A-D Air 3

Unto 18 l$

M-sh-l 3

Chuck u

e M-36-1 IFWO36A-D 1rWO39A-D Air t.

Umt a n

n M. t r.

i l

IFWU4JA-D Al t-1 O l ut.o 11 h

M t r.

I 6

Check B

C M-J6-1 l

1FWO75A-D LIA0M Air 1

U l ot en H

A M - Ye - >

11A0bb Ali i

uloba H

A M-W /

.5 Globu u

'A M-$$-2 IIA 088

.75 Check 11 A

M-%-2 IIA 091 1MS001A-D Comp. Air 32.75 Cato u

A M-35-1 1MS013A-D 6

Relief B

C M-35-1 6

Relief B

C M-35-1 1MS014A-D IMS015A-D 6

Relief B

C M-35-1 1FS016A-D 6

Relief B

C M-35-1 6

Relief D

C M-35-1 IMS017A-D 1MS018A-D Comp. Air 0

Relief D

C M-35-1 IMS021A-D 3

Globe B

A M-35-1 1MS101A-D Air 4

Gate B

A M-35-1 0@G059 Motor 3

Butterfly B

C M-47-2 0@G060 Motor 3

Butterfly D

C M-47-2 0@G061 Motor 3

Butterfly D

C M-47-2 0@G062 Motor 3

Butterfly B

C M-47-2 0@G063 Motor 3

Butterfly B

C M-47-2 0@G064 Motor 3

Butterfly B

C M-47-2 1 G057A Motor 3

Butterfly B

A M-47-2 l@G079 Motor 3

Butterfly B

A M-47-2 l@G080 Motor 3

Butterfly B

A M-47-2 l@G081 Motor 3

Dutterfly D

A M-47-2 1 G082 Motor 3

Butterfly D

A M-47-2 l@G003 Motor 3

Butterfly B

A M-47-2 l@G084 Motor 3

Dutterfly u

A M-47-2 l@G085 Motor 3

Dutterfly D

A M-47-2 1PR001A,B Air 1

Globe B

A M-78-10 1PR031 Air 1

Globo B

A M-78-10 3.9-106

TABLE 4.1-1 s

l REACTOR DESIGN COMPARISON TABLE l

l BYRON AND BRAIDWOOD BYROE AIC BRAIDWOOD UNITS 1 and 2 UNITS I and 2 WCAP-950C THERMAL AND RYDRAULIC DESIGN PARAMETERS (fEtt PARASITIC FUEL)

(OPTIVIZED FUEL)

REFERDeCE CIESIGN 1.

Reactor Core Beat output, (1004), set, 3411 3411 3411 6

2.

Reactor Core Beat Output, 10 Stu/br 11641.7 11641.7 11641.7 3.

Beat Generated in Fuel, t 97.4 97.4 97.4 III 4.

Core Pressure, Nominal, psia 225C 2280 2290 III 2220 2250 2250 5.

System Pressure, Minimus Steady State, psla 6.

Minimum DNBR at Nominal Conditions Typical Flow Channel 2.09 2.4~

2.40 Thimble (Cold Wall) Flow Channel 1.74 2.32 2.26 f

7 Minimum DNBR for Design Transients

{-

Typical Flow Channel

>1.30

>1.49

>1.49 1

3*

>1.47

> 1. 47 1

Thimble Flow Channel II tr5.3-1 (2) 8.

DNB Correlation

  • R* (W-3 with Modified SEB-!

Spacer Factor)

C00:. ANT PLOW 9.

Total Thermal Flow Rate, 10 lb,'hr 138.6 144.5 142.3 l

5

10. Ef fective Flow Rate for Beat Tra:sf er, 10 lb,/hr 132.4 135.'

134.'

l 6

11. Ef f ective Flow Area for Heat Transfer, f t 51.1 54.1 54.1
12. Average Velocity Along Fuel Rode, f t/see 16.4 14.2

!!.i 0

13. Average Mass Velocity, 10 lb,/hr-ft 2.59 2.56 2.41 (1) Values used for thermal hydraulic core analysis (2) The W-3 correlation is used for analysis of some accidents inv::1ving dep essurization cf the steam system.

(See Table 15.0-2, sheet 1)

I, l

1 4

TAaLE 4.1-1 (Cont'd) s BTDON AND BRAIDWOOD BYROW AND BRAIDWOOD UNITS 1 and 2 UNITS 1 and 2 UCAP-9500 l

THf2 MAL AND HYDRAULIC DESIGN PARAMETERS (IDW PA.RASITIC FUEL)

OPTIMIZED FUEL REFFv3NCE DESitM COO!. ANT _ TEMPERATURE, U{

14. Ncuinal Inlet 556.9 559.2 561.6
15. Average Rise in Vessel 61.1 58.4 58.5
16. Average Rise in Core 63.6 60.7 61.8
17. Average in Core 590.4 591.1 594.2
18. Average in Vessel 5??.4 588.4 592.3 HEAT TRANSFER 2
19. Active Beat Tr ansf er, Surf ace Area, it 59,700 57,530 57,500
20. Average Beat Flux, Btu /hr-f t 185,800 197,200 197,200 "b
21. Manism Beat Fluz f or Normal Operation,

(

b 2

440.330 457,500 457,500 Y

Bts/hr-f t h

22. Average Linear Power, kW/f t 5.44 5.44 5.44
23. Peat Linear Power for Normal Operation, kW ft(*)

12.6 12.6 12.6

24. Peak Linear Power Resulting f rotn Overpower Transients / Operator Errors (assasing a I**3 masince overpower of 1186), kW/ft 18.0 18.0 18.C 25 Peak Linear Power f or Prevention of Centerline Melt, kW/ f t (* * * )

>18.0

$1e.0

>18.C This limit is associated with the value of Fo = 2.32

    • See Sabsection 4.3.2.2.6
      • See Sabsection 4.4.2.11.4.

TABLE 4.1-1 (Cont'd) s BYRON AND BRAIDWOOD BYRON AND BRAIDWOOD UNITS 1 and 2 UNITS 1 and 2 WCAP-9500 THERMAL AND HYDRAULIC DESIGN PARAMETER _S (LOW PARAS!?!C FUEL)

(OPTI411ED FUEL)

REFERENCE DESIGN

26. Power Density, kW per Liter of Core *I 104.5 104.5 104.5 I
27. Specific Power, kW per kg Uranium 38.4 41.9 41.9*

~

FUEL CENTRAL TEMPERATURE

28. Peak at Peat Linear Power for Prevention of Centerline Melt, OF 4700 4700 4700
29. Pressar e Drop **I I

I Across Core, psi 26.9+2.7 ***I 26.312.6 25.7+2.6 l

Across Vessel, Including Nozzle psi 47.4+4.7

      • I 46.4+4.6 45.7+4.6 l {

I i

CORE MECHANICAL DESIGN PARAMETERS

n a

k

30. Design RCC Canless RCC Casless RCC Canless 17 x 17 17 s 17 17 x 17

(

31. Namber of Fuel Assemblies 193 193 193
32. 00 Rods per Assembly 264 264 264 2
33. Rod Pitch, in.

3.496 0.496 0.496

34. Overall Dimensions, in.

8.426 x 8.426 8.426 s 8.426 f.426 x 8.426

35. Fuel Weight (as UO ),

Ib 222,739 204,236 234,236 2

36. Clad Weight, Ib 50,913 43,376 43,376 Based on cold dimensions and 954 of theoretical density fuel Based on best estimate reactor flow rate as discussed in Section 5.1

-+

Pressure drops revised based on results f ram Ref erence 2.

+++

O t

_. _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ ~. _

a

.=

TABLE 4.1-1 (Cont'd) i l

BYRON AND BRAIDMOOD BYROE AND BRAr e UNITS 1 and 2 UNITS 1 and 2 NCAP-9500 THERffL AND BYDRAULIC DESIGN PARAMETERS (LON PARASITIC FUEL)

(O MINIS B FUEL)

REFERSBCE DESIGN

37. Number of Grids per Assembly 8 - Type R 2-Type R, 6-Type 3 2-Type R, 6-Type 3
38. Composition of Grids Inconel 718 2 End Grids -

2 End Grida -

Inconel 718 Inconel 718 6 Intermediate 6 Intermediate Grida - Zircaloy 4 Grids - 31rcaloy 4 39 Loading Technique 3 Region Monunif orn 3 Region Monanifore 3 Region Nonuniform CORE MEOHANICAL D?.SIG4 PARAMETERS FUEL RODS 5

4 0. Number 50,952 50,952 50,952 8,8 7

41. Oatside Diameter, in.

0.374 0.360 0.360 E

i

42. Dia: net.ral Gap, in.

0.0065 0.0362 0.0C62

=

.a 4 3. Cladding Micioness, in.

0.0225 0.0225 0.0225

~

44. Cladding Ma erial 2i r ealoy-4

' '

  • aloy-4 Zir aloy-4 FUEL PELLETS 1

r

45. Material U0 Sintered 0 51 stere $

U0 Sintered 1

2 2

2 l

46. Density (4 of Theoretical) 95 95 1+ 5
47. Diameter, in.

0.3225 C.3088 0.3088 l

~

  • l 0.530 C.507 0.507
48. Length, in.

l l

l l

l

m-s-

er 1

4 TABLE 4.4-1 s

m s

o g.

s..

1"'ERNAL AND BYDRAULIC CDNPARISON TABLE I.

BYRON AND BRAIDMDOD WCAP 9500 SYRON AED Bpa--arno UNITS 1 AND 2 REFERENCE UNITS 1 AND 2 l

p_Es_I_cy /ARAMETERS thW FARASITIC FUEL DESIGN (OPTIMIIED FUEU L

Reactor Core Beat Output (1004), Mut 3411 3411 3411 0

Reactor Core Beat Output, 10 Btufbr 11641.7 11641.7 11641.7 i

j Beat Generated in Fuel, t 97.4 97.4 97.4 g

II System Pressure, Nominal, psis 2250 2280 2280 1

System Pressure, Minimum Steady-State, psia (2) 2220 2250 2250 I'

II Minimus DNBR at Nominal Conditions Typical Flow Channel 2.09 2.4C 2.47 Thimble (Cold Wall) Flow Channel

(

1,74 2.26 2.32 III t

~

Minimum DNBR for Design Transients 1 49 11.49 4

g Typical Flow Channel 11.30 1

m Thimble Flow Channel 31.30

1.47 11.47 g

DNB Correlation

'R' (W-3 with WRB-1 WRB-1 Modifief Spacer Factor) 1 3

COO! ANT FIDW 6

~

Total Thermal Flow Rate,10 lb,/hr 138.6 143.3 144.9 Effective Flow Rate for Beat 132.4 1M 7 138.7 6

s Transfer, 10 lb,/br

'y Ef f ective F1 Area for Beat 51.1 54.1 54.1 Transfer, ft w

Average Velocity Along Fuel 16.4 15.8 If,2 lA-

?

Eods, (t/sec L

-- '7

' Average Mass velocity,10 lb,/hr-ft 2.59 2.49 2.56 6

2

{

i

~

'f'

(

., [I

)

)

y sw f,

, _.Js. '

r v

> ~

_ '+

4 4 :

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,a e

f

\\*

.y 3y y

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,_ / '

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, f j'

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TABLE 4.4-1 (Continued) a r

w

{

. 9 7

_11FTJtWAND HYDRAULIC _Q(b! SON TAR *I

)

~

4 1

'}je j

  • \\

a BYRON AMD' 3RAIDWOOD '

WCAF l;9500 BYRON AsD BRAIDu00D

'1

+

f

UNITS l'AND 2 REFERENCE UNITS 1 AND 2 h[

/

  • " ; 9 DESIGN PARAMETERS LOW PARAf,JTIC PUEL DrS?5

_ (OPTIMIIED FCEL) _

COOIANT TPMPERATURE

/

Nominal Inlet, F 556 9

' y 561.4

$$9.2

,,ky

},

Average Rise in Vessel, CF

.,," f, 5 e. S- _

58.4 61.1-j ' [f g-Average Rise in Core, OF

  • / './

63.6

/'

  • 1.3 60.7

[' f $* g q

I

  • i 591.1

{,,

,Q/

Average in Core, F

/

59 3.0

't

[/

/94.2

' ' 192./i, /g" / )

f id/'

Average in Vessel, F

't "587.4 588.4

/

y e

/

s

  • e

,.3 v

s n

~

m 1.

-,e

.s e

9 HEAT TRANSFER I

i

= -

2 I'

/,

Active Beat Transf er, Surf ace Area.- f t 59,700 57,500 57,500 189,800

  • )7,203 197,200 Average Beat Flux, Btu /hr-f t i

Maximus Beat Flus for Normal l

440,300 457,500 457.500 5 '-

2 Operation, Btu /hr-ft

((,

Average Linear Power, kW/ft 5.44 5.44 5.44

'r7 Peak Linear Power for Normal 12.6 12.6 12.6 W

Operation, kW/ft

,1 g

Peak Linear Power Resulting fras Overpower 18.0 18.C 18.0 Transients / Operator Errors (assuking a marinua M

overpower of 1184), kW/ft p

Peak Linear Power for Preversion of Centerline

>18.0

>18.C

>18.0 I

I Melt, kW/ft Power Density, aW per liter of core *3 104.5 1:4.5 104.5 I

Specific Power kW per kg Crantua *I 38.4 41.9 41.9

~

I l'

o J

e 9 6

/

5

.d

[..

I.

TABLE 4.4-1 (Continued) s THERIEhL AND NTDRAULIC (XMEPARISON TABLE BYm05 AND maar'wm MCAP 9500 RYaoss AltD RMIDuD00 DMITS 1 AIID 2 REFEREIBCE UNITS 1 AE3 2 l

DESIGN PARAsurrERS IAN PARASITIC FUEL _

_DESIGIf (OPTIMIEED FUE.1 l

FUEL CENTRAL TEMPERAMRE Peak at Peak Linear Power for Prevention 4700 4700 4700 of Centerline Melt, "F Pressure Drop "I I

Across Core, psi 26.912.7 25.7 1 2.6 26.3 + 2.6 l

l Across vessel, including nozzle psi 47.4 + 4.7

  • 45.7 1 4.6 46.4 + 4.6 D.

r o

E t3 i

I Tais limit is associated with the value of FO " 2 32 See subsection 4.3.2.2.6.

m See subsectiori e. 4.2.11.6.

Based on cold dimensions and 95% of theoretical density f uel

+

u Based on best estimate reactor flow rate as discussed in Se: tion 5.1

+++

Pressure Drops Mateu based on results from Reference 5.

(1)

These naabers are not directly comparable for each plant desig: dae to the incorporation of a dif ferent thermal design procedure and DK3 correlation in the prosect core.

1_

(2) value used for thermal hydraulic core analysis.

4 k

1 1

4 1

63 0

$20 4t7.6er 610 600

~

.u.

w w 590 hot 68S,8f6 a:

E F

E 580 5

T gyg n

570 T

e.6b COLD e..g g..

sj 557.00 F 550 I-1 I

l 5'40 0

20 h0 60 60 100 PERCENT P0nER e

%M SYRCN/ER * :OWOOO 277;O* S F'IN AL EM E'Y AN A VS;; 1?"AI Figure 4.4-9.

Rea::or Cooten: Sys:em Tem erature Percen: P o w er *.* a s i

9

B/B-FSAR operating planto have indicated that the actual flow has been well above the flow specified for the thermal design of the plant.

By applying the design procedure described in the following, it is possible to specify the expected operating flow with reasonabic accuracy.

Three reactor coolant flow rates are identified for the various plant design considerations.

The definitions of these ficws are presented in the following paragraphs.

Best Estimate Flow The best estimate flow is the most likely value for the actual plant operating condition.

This flow is based on the best estimate of the reactor ves,sel, steam generator and piping flow resistance, and on the best estimate of the reactor coolant pump head-flow capacity, with no uncertainties assigned to either the system flow resistance or the pump head.

System pressure drops, based on best estimate flow, are presented in Ta' le 5.1-1.

o Although the best estimate flow is the most likely value to be expected in operation, more conservative flow rates are applied in the thermal and mechanical designs.

Thermal Design Flow Thermal design flow is the basis for the reactor core thermal performance, the steam generator thermal performance, and the nominal plant parametera u.ced throughout the design.

To provide the required margin, the thermal design flow accounts for the uncertainties in reactor vesael, steam generator and piping flow resistances, reactor coolant pump head, and the methods used to measure flow rate.

The thermal design flow is approximately 5.9%

l less than the best estimate flow.

The thermal design flow is confirmed when the plant is placed in operation.

Tabulations of important design and performance characteristics of the reactor coolant systems, as provided in Table 5.1-1, are based on the thermal design flow.

Mechanical Design Flow Mechanical design flow is the conservatively high flow used in the mechanical design of the reactor vessel internals and fuel assemblies.

To ensure that a conservatively high flow is specified, the mechanical design flow is based on a reduced syctem resistance and on increased pump head capability.

The mechanical design flow is approximately 3.7% greater than the l

best estimate flow.

Maximum pump overspeed results in a peak reactor coolant flow of 120% of the mechanical design flow.

This overspeed condition, which is coincident with a turbine generator overspeed of 20%, is only applicable if, when a turbine trip would be actuated, the turbine governor fails and the turbine is tripped by the mechanical overspeed trip device.

5.1-4

B/D-FSAR TABLE 5.1-1 SYSTEM DESIGN AND OPERATING PARAMETERS Plant design life, years 40 Nominal operating pressure, psia 2250 12,074 Total system volume including 3

pressurizer and surge line, ft System liquid volume, including 11,695 pressurizer water at maximum 3

guaranteed power, ft Pressurizer spray rate, gpm 900 Pressurizer heater capacity, kW 1800 Pressurizer relief tank volume, ft 1800 SYSTEM THERMAL AND HYDRAULIC DATA (Based on Thermal Dcaign Flow) 4 PUMPS 3 PUMPS **

l RUNNING RUNNING NSSS power, MWt 3425 2569 Reactor power, MWt 3411 2560 Thermal design flows, gpm*

Active loop 94,400 98,000 l

0 Idle loop Reactor 377,600 294,000 6

Total reactor flow, 10 lb/hr 140.3 110,5 Temperatures, T Reactor vessel outlet 618.4 612.2 Reactor vessel inlet 558.4 552.3 9

5.1-8

B/B-FSAR TABLE 5.1-1 (Cont'd)

Steam generator outlet 558.1 552.1 Steam generator steam 543.3 538.0 Feedwater 440 408.0 Steam pressure, psia 990 947 Total steam flow, 10 lb/hr 15.13 10.84 13est estimate flows, gpm*

Active loop 100,300 105,300 l

0 Idle loop Reactor 401,200 315,900' l

Mechanical design flows, gpm*

Active loop 104,000 109,500 l

0 Idle loop Reactor 416,000 328,500 l

SYSTEM PRESSURE DROPS (Based on Four-Loop Best Estimate Flow)

Reactor vessel AP 44.7 Steam generator AP, psi 38.3 Hot let piping Ap, psi 2.3 Pump suction piping AP, psi 3.2 Cold leg piping AP, psi 2.3 Pump head, feet

'290 l

  • At pump discharge.

5.1-9

.,,}

NOTES TO FIGURE 5.1-2 MODE A STF50Y-STATE FL'LL POWER OPERATIO*1 PRESSURE TEMPERATURE FLOW LOCATION FLUID PSIG "F

GPM(I)

LB/HR(2)

-VOLUME 1

R.C.

2235.0 617.9 112,259 37.36

[2 2233.1 617.9 112,159 37.33

'3 2195.9 556.7 100,300 37.33 4

2192.4 556.7 100,494 37.40 25 2285.1 556.9 100,400 37.40 l6 2283.2 556.9 100,300 37.36 7(3) 2234.1 617.9 100 0.0333 8(4) 2285.1 556.9 100 0.0371

  • 9 2194.2

. 587.0 199 0.0704 10-18 SEE LOOP #1 SPECIFICATIONS If-27 SEE LOOP #1 SPECIFICATIONS 2q-36 SEE LOOP fl SPECIFICATIONS

' 17 2285.1 556.9 1.0 0.0004 38 2285.1 556.9 1.0 0.0004 39 2235.0 556.9 2.0 0.0008 720 40 STEAM 2235.0 652.7 1080 41 R.C.

2235.0 652.7 42 2235.0 652.7 2.5 0.0008 43 2235.0 652.7 2.5 0.0008 44 STEAM 2235.0 652.7 0

0 45 R.C.

  • 2235.0

<652.7 0

0 MINIMIZE 46 N

3.0 120 0

0 2

47 R.C.

2235.0

<652.7 0

0 MINIMIZE

\\

48 N

3.0 120 0

0 2

49 3.0 120 0

0 450 50 3.0 120 51 PRT 3.0 120 1350 WA1ER (1) At the conditiens specified.

6

, (2) X 10 (3) Location point refers to the three 1" connections on the hot leg.

(4) location point refers to the 2" connection on the cold leg.

5) M:.sI reccAT Calculthio.1 BYRON /BR AIDWOOD STATIONS O f b c S I S S E. N I C flW FINAL S AFETY AN ALYSIS REPORT is 100,200 ym. Utlu es oniMs Mle_ waldc4 FIGURE 5.1-2 Q' m;3; mally afTecAd REACTOR C00LM T SYSTE!!

PROCESS FLOW DIAGRA'1 i-f rc u lt4IdI'J "s.# #)

(SHEET 2 of 2)

Ltest best espia4e f \\cW.

B/B-FSAR 5.4.2.5.3 Mechanical and Flow-Induced Vibration Under Normal Operation In the design of Westinghouse steam generators, the potential for tube wall degradation attributable to mechanical or flow-induced excitation has been thoroughly evaluated.

The eval-uation included detailed analyses of the tube support systems for various mechanisms of tube vibration.

The primary cauce of tube vibration in heat exchangers is hydrodynamic excitation due.to secondary fluid. flow on the outside of the tubes..In the range of normal steam generator operating conditions, the effects of primary fluid flow inside the tubes and mechanically induced tube vibration are considered to be negligible.

To evaluate flow induced tube vibration in the preheater region of the tube bundle, Westinghouse undertook an extensive program employing data from operating plants, full and partial scale model tests, and analytical tube vibration models.

Operating plant data consisted of tube wear data from pulled tube eval-uations and eddy current tests and tube motion data from accelerometers installed inside selected tubes.

Model testing generated tube wear data, flow velocity distributions, tube motion parameters, and flow-induced tube vibration forcing functions.

The tube vibration analyses applied the forcing functions to produce tube motion data.

The results of this evaluation were consistent with the early operating experience of preheat steam generators.

On the basis of an extensive model test and analysis program, Westinghouse designed, verified, and implemented a modification to the steam generator to reduce tube vibratory response to preheater inlet flow excitation.

Additionally, the magnitude of the flow forcing function was reduced through implementation of a preheater flow bypass arrangement in the feedwater system.

The verification of the performance of the modifications in reducing tube excitation and response was done with input from a full-scale test under simulated conservative flow and tube support conditions.

Fatigue of the tubes in the preheater region which are subject to flow-induced excitation is not a concern since the maximum resultant stresses in the tube are below the endurance limit of the material.

For areas of the tube bundle other than the preheater, parallel flow analyses were performed to determine the vibratory deflections.

-These-analyses indicate that the flow velocities are-sufficiently low such that they result in negligible fatigue and vibratory amplitudes.

The support system, therefore, is deemed adequate with regard to parallel flow excitation.

5.4-15

B/B-FSAR To evaluate crossflow at the exit of the downcomer flow to th.e tube bundle and at tne top of the bundle in the U-bend area, Westinghouse performed an experimental research program of crossflow in tube arrays with the specific parameters of the. steam generator.

Air and water model tests were employed.

The results of this research indicate that these regions of the bundle are not subject to the vortex shedding mechanism of tube excitation.

Vortex shedding was found not to be a significant mechanism in these two regions for the following reasons:

a.

Flow turbulence in the downcomer and tube bundle inlet region. inhibit the formation of Von Karman vorticles.

.b.

Both axial and crossflow velocity components etist on the tubes.

The axial flow component disrupts the von Karman vortices.

This research program was also the basis for evaluation of the fluid-elastic mechanism due to cross flow at the tubesheet.

The evaluation showed the adequacy of the tube support arrangement.

Flow turbulence can result in some L.oe excitation'in these regions.

This excitation is of little concern, however, since:

a.

Maximum stresses in the tubes are at least an order of magnitude below the fatigue endurance limit of the tube material, and b.

Tube support arrangements preclude significant

' vibratory r.otion.

In summary, tube vibration has been thoroughly evaluated.

' Mechanical and primary flow excitation are considered negligible.

Secondary flow excitation has been evaluated.

From this evaluation, it is concluded that if tube vibration does occur, the magnitude will b'e limited.

Tube fatigue due to the vibration is judged to be negligible.

Any tube wear resulting from the tube vibration would be limited and would progress slowly.

This allows use of a periodic tube inservice inspection program for detection and followup of any tube wear.

This inservice inspection program, in conjunction with tube plugging criteria, provides for safe operation of the steam generators.

5.4-16

D/B-FSAR

~ ~ ~ 5 3. 2. 5. 4 A11'owable Tube wall Thinning Under Accident conditions

~

An evaluation is performed to determine the extent of tube wall thinning that can be tolerated under accident conditions.

Under such a postulated design-basis accident, vibration is of short enough duration that there is no endurance problem.

The results of a study made on "D series" (.75 inch nominal diameter.043 inch nominal thickness) tubes under accident loading are discussed in WCAP-7832 (Reference 3) and show that a minimum wall thickness of.026 inches would have a maximum faulted condition stress (i.e., due to combined LOCA and safe shutdown earthquake

. loads) that is less than the allowable limit.

This thickness is

.010 inches less than the minimum steam generator tube wall thickness.039 reduced to.036 inches by the assumed general corrosion and erosion loss of.0033 inches.

The corrosion rate is based on a conservative weight loss rate for Inconel tubing in flowing 6500 F primary side reactor coolant fluid.

The weight loss, when equated to a thinning rate and projected over a 40-year plant life with appropriate reduction after initial hours, is equivalent to.083 mils thinning.

The assumed corrosion rate of 3 mils leaves a conservative 2.917 mils for general corrosion thinning on the secondary side.

The steam generator tubes, existing originally at their minimum wall thickness and reduced by a very conservative general corrosion loss, still provide quite an adequate safety margin.

Thus, it can be concluded that the ability of the steam generator tubes to withstand accident loadings is not af fected by a lifetime of general corrosion losses.

\\

5.4-16a l

- _ _ _ _ _ _ = __

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6 l

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9 e.

B/B-FSAR and under startup and light load conditions when the preheater section is bypassed.

The water hammer preventive features are more fully described in the section that follows.

As shown on Figure 10.1-1, the valves and the piping downstream thereof are Safety Category I,.

Quality Group B.

The valves and their downstream sections of Category I main feedwater and tempering piping are located in the same Category I valve rooms which house the main steamline isolation valves described in Section 10.3.

10.4.7.3 Water Hammer Prevention Features Several water hammer prevention features have been designed into the feedwater system.

These features are provided to minimize the possibility of various water hammer phenomena in the steam generator preheater, steam generator main feedwater inlet piping and the steam generator upper nozzle feedwater piping.

The following discussion is typical for each of the four steam generators and their associated feedwater piping.

10.4.7.3.1 start-Up, Low Load conditions a.

Under start-up and low load conditions when NSSS rated flow is less than 15% and temperatures are less than 250' F, feedwater will only be admitted to the upper nozzle of the steam generator by the use of flow through the feedwater bypass tempering line and/or flow through the feedwater preheater bypass line via the feedwater bypass control valve and feedwater preheater bypass valve.

The 6-inch diameter upper nozzle is located on the upper shell of the steam generator, below the normal, full power water level.

Level control in the steam generator is provided by the feedwater bypass control valve at these conditions.

b.

Surface mounted resistance temperature detectors (RTD) are provided on each of the feedwater pipes, leading to and very near the steam generator's upper nozzle to detect during start-up and low load conditions as well as other operating conditions, possible back leakage of steam from the steam generator into the feedwater piping.

These RTD's are monitored by the plant proccas computer and alarmed in the main control room so that actions can be taken to initiate feedwater flow to the upper nozzle before potential feedwater hammer conditions may develop.

10.4.7.3.2 Increasing Load a.

As load increases about 15% of NSSS rated flow and feedwater temperatures rise above 250' F,

forward feedwater flushing of the main feedwater piping may be initiated by opening 10.4-11

4.

B/B-FSAR the feedwater isolation bypass valve.

A small controlled flow through the 3-inch feedwater isolation bypass line is provided to flush the main feedwater pipLng between the isolation valve and the steam generator.

b..

Three sets of three RTD's are provided on the main feed-water piping upstream and downstream of the feedvater isolation valve and near the steam generator feeilwater nozzle to detect when the feedwater flushing tem,aerature rises above 255 F.

Two out of three logic is pcovided for each set of three RrD's and all three must be satisfied to meet the forward flushing temperature require nents.

c.

If flow in the 3-inch feedwater isolation valve bypass line (forward flushing flow) remains above a preset.ainimum and below a preset maximum and the flushing temperatures remain satisfied, a timed period occurs after which a

. permissive signal is provided to automatically open the

.feedwater isolation valves.

Automatic opening of a feed-water isolation valve can be blocked by placing its control switch in the main control room in the closed position.

This automatic permissive to open occurs after a timed period to allow epproximately two volumes of water to be purged from the piping between the feedwater isolation valve and the steam generator main feedwater nozzle.

Feedwater flow at the main feedwater flow-element must also be above a preeet minimum in' order for the feedwater isolation valve to open.

1 d.

After the feedwater isolation valve has opened, the feed-water isolation bypass valve will be manually closed.

e.

Prior to opening of the feedwater isolation valve, transfer from the feedwater bypasc control valve to the feedwater control valve will occur in order to provide steam generator level control at the higher feedwater flow conditions.

f.

If flow to the steam generators remains continuous during a load transient and above a minimum flow rate, feedwater will not be terminated to the main feedwater nozzle even if. temperature of the feedwater has dropped below 250* F.

Interruptior. or a reduction in flow below the minimum rate however, will cause the feedwater preheater section of the steam generator to be bypassed.

g.

Steam generator low level trips are provided to close all of the feedwater isolation valves, feedwater isolation bypass valves and feedwater preheater bypass valves.

Steam generator low pressure trips are provided to close all of the feedwater isolation valves, feedwater isolation bypass valves, feedwater preheater bypass valves and the feedwater bypass tempering valves.

10.4-12

  • .3 B/B-FSAR 10.4.7.3.3 Split Feedwater Flow a.

Prior to opening of the feedwater isolation valve, the majority of feedwater flow at the lower power level is introduced to the upper nozzle of the steam generator by the preheater bypass pipe.

b.

At higher power leve?.s after the feedwater isolation

[

valve has opened, only a small portion of the feedwater flow bypasses the preheater, with the bypass portion contributing to approximately 10% of full feedwater flow at 100% power.

This split feedwater flow arrangement provides an approximate 90% of full flow limit'to the main feedwater nozzle at higher power levels in order to minimize the potential for tubing vibration in the steam generator.

The feedwater flow rate to the steam generator nozzle is monitored and alarmed, if flow rises above approximately 90%, in order for actions to be taken to reduce flow.

c.

The preheater bypass valve remains open throughout the start-up and low load conditions, as well as up to and including full power operation.

i 10.4.7.3.4 Other Upper Nozzle Feedwater Line Uses Innsmuch as there is water flowing to the upper nozzle of the steam generator during normal operation, and it is the required location for introducing cold fluid into the steam generator, i

auxiliary feeedwater and chemical feed are connected to the upper nozzle feedwater lines rather than to the main feedwater lines.

The chemical feed lines are used to add chemicals directly to the steam generators under low load conditions prior to wet layup.

The chemical feed and auxiliary feedwater lines are Safety category I, Quality Group B out to, and including their isolation valves.

10.4.7.4 Safety Evaluation The condensate and feedwater systems are not safety-related except l

as described in Subsection 10.4.7.1.1.

If it is necessary to remove a component such.as a feedwater heater, pump,-or control valve from service, continued operation of t1e system is possible by use of the multistream arrangement and the provisions for removing from service and bypassing equipment and sactions of the system.

An abnormal operational transient analysis of the loss of a feed-water heater string is included in Subsection 15.1.1.

(

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%*4 B/B-FSAR o

errors in the determination of the steady-state power level are made as described in Section 15.0.3.2.

The thermal power values used for each transient analyzed are given in Table 15.0-2.

In all cases where the 3579 megawatt thermal (MWt) rating is used in an analysis, the resulting transients and consequences are conservative compared to using the 3425 MWt rating.

The values of other pertinent plant parameters utilized in the accident analyses are given in Tables 15.0-3 and 15,0-4.

15.0.3.2 Initial Conditions For most accidents which are DNB limited, nominal values of initial conditions are assumed (including an appropriate tem-perature margin to compensate for steam generator tube fouling).

The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-8567 (Reference 10).

This procedure is known as the " Improved Thermal Design Procedure," and is discussed more fully in Section 4.4.

For accidents which are not DNB limited, or in which the Improved Thermal Design Procedure is not employed the initial conditions are obtained by adding the maximum steady state errors to rated values.

The following conservative steady stete errors were assumed in the analysis:

a.

Core Power

+2% allowance for calori-metric error b.

Average Reactor i 4.90F allowance for l

Coolant System controller deadband and temperature measurement error and steam generator fouling penalty c.

Pressurizer i 30 pounds per square inch pressure (psi) allowance for steady state fluctuations and mea-surement error Table 15.0-2 summarizes initial conditiens and computer codes used in the accident analysis, and shows which accidents employed a DNB analysis using the improved thermal design pro-cedure.

15.0.3.3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution.

The nuclear design of the reactor core ninimizes adverse power distribution through the placement of control rods and operating instructions.

Power distribution may be characterized by the radial factor (FAH) and the total peaking factor (F ).

The peaking Q

factor limits are given in the technical specifications.

15.0-G l

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