ML18092A758

Proposed Tech Spec Revs,Reflecting Increase in Rated Thermal power.Marked-up Draft FSAR Changes Encl

ML18092A758
Person / Time
Site:	Salem
Issue date:	08/06/1985
From:	Public Service Enterprise Group
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Shared Package
ML18092A756	List: ML18092A755 ML18092A757 ML18092A758
References
NUDOCS 8509050151
Download: ML18092A758 (38)

Text

APPENDIX BB TECHNICAL SPECIFICATION REVISIONS HANO MARKED

( --------: ----

8509050151 850806 PDR ADOCK 05000272 P

PDR 2737e:1d/032285 8-4

1.0 0-EFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear 1n capitalized type and are appl 1cabl e throughout these Technical Specifications.

THERML POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the ructor coolant.

RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be.,a_ tota1 reactor core heat transfer rate to the reactor coohnt of ~MWt.

OPERATIOAAL MOOE

~11 1.4 *An OPERATIO~L MODE shall correspond to 1ny one inclusive combina-tion of core reactivity cond~t1on, power 1evel and average reactor coolant tenperature specified in Tab1e l.1.

ACTION 1.5 ACTION shall be those additional requirenents specified as corollary statements to each principle specification and shall be part of the spec 1 ficat1ons.

OP~ABLE -

OP~ABILITY 1.6 A system, subsystl!n, train, component or device shall be OPERABLE or have OPERABILITY when 1t is capable of perfonning 1ts specified function(s). Implicit in this definition shall be the assumption that 111 necessary attendant instrumentation, controls, nonnal and energency electrical power sources, cooling or seal water 1 lubrication or other auxi1 equipnent that are. required for the systen, subsystem, train, component or device to perform its funct1on(s) are also capable of perfonning their rela support funct1an(s)

• REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURREr<<:E shall be any of those 1:ond1t1ons specified 1n Spec1fitat1ons 6.9.1.8 and 6.9.1.9.

SALEM - UNIT 1 1-1 Amencinent Ho. 9

w -

A PARl\\Mt:HR He.1ctor Coolant Sy-;tem Tavg l 1 n~c.,'iuri zer Pressure Tl\\BLE 3.2-1 ONR PARAMETERS i1 Loops In Operation 58'l.

~>810f

. LIMITS

• 2220 ps1a*

. 349,299 ljf.n I

• f1iiliCnot applicable d-urTniJ either a THERMAL POWER ramp 1ncease in excess of s: RATED THERMAL POWER 1*er 11d11ule or a TllCRMI\\!. l'OW[R step increase in excess of lM RATED THERMAL POWER.

ATTACHMENT # 2

ATTACHMENT # 2 DRAFT FSAR CHANGES TO BE INCORPORATED UPON ISSUANCE OF. THE SALEM UNIT 1 POWER UPRATE AMENDMENT The following changes have been identified:

Chapter 1

- Change page 1.0-1 as per the attached marked-up page.

chapter 2&3 - No Changes Chapter 4*

- Delete Tables 4.1-lA (six pages), 4.4-lA (two pages) 4.4-2A (one page, and 4.4-3A (one page).

ter 5 Change Tables 4.1-lB (six pages), 4.4-lB (two pages),

4.4-2B (one page), and 4.4-3B (one page) as per the attached marked-up pages.

Change throughout section 4.4 any reference to *Tables 4.4-lA and a* to "Table 4.4-1*.

Similarly for Tables 4.4-2A and B, and Tables 4.4-3A and B.

Change pages 4.4-5, 4.4-13, and 4.4-54 as per the attached marked-up pages.

- Revise Tables 5.1-1 (one page), 5.2-3 (two pages),

5.2-5 (two pages), 5.2-7 (one page) and 5.5-1 (one page) per the attached marked-up pages.

Chapters 6, 7, 8, and 9 -

No Changes.

Chapter 10

- Revise pages 10.2-1 and 10.3-2 per the attached marked up pages.

Chapters 11, 12, 13, 14 and 15 -

No Changes.

• Section 4.3, Nuclear Design, is based on Cycle 1 core design.

since the uprate will not be implemented until cycle 7, it is not appropriate to change section 4.3 to reflect the uprated conditions.

Revisions to the Nuclear Design are addressed by the reload analyses, section 4.5.

1.0 - INTRODUCTION AND

SUMMARY

This Updated Final Safety Analysis Report is submitted pursuant to the requirements of 10 CFR 50.71 by Public Service Electric and Gas Company (PSE&G) for the two nuclear power units at its Salem Generating Station.

PSE&G and Westinghouse Electric Corporation have jointly participated in the design and construction of each unit.

The plant is operated by PSE&G.

Each unit employs a pressurized water reactor nuclear steam supply system furnished by Westinghouse which is similar in design con-cept to several other projects licensed by the Nuclear Regulatory Cam-mi sion.

The only systems shared by the two units are Compressed Air, Demineralized Water and the Solid Radwaste Handling System.

There are a minimum of shared components; chemical drain and laundry hot shower tanks and pumps are the only components in common.

CCl\\R. ~ov..AA- ~

'DoY... ~

~~ -' ~ 31.f \\ \\ Tf\\\\l t. ~

The Jicensed"rati Fl§S ef U1e tw*e 1:u~its are as felle*,.*s: l:JF1it 1 :n3a MHt, aF1a l:JF1it 2 3411 MWt.

The

• • • al"l"!l'lte~ gross and approximate net elec-trical outputs* are 1132 mJe-aflti-+/-090 MWe respect~ly fer Urlit 1 aAe 1)58 MWe and 1115 MWe respectively fe1* UF1it 2 *. The reactors are expected to be capable of outputs of approximately 3494 MWt (Unit 1) af'le!

3570 MWt (U1dt 2), which corresponds to the valves-wide-.open rating of the turbine generators of 1176 M'.Je ~reB and 1130 M'we net fo1 Unit 1,

~

1201 MWe gross and 1155 MWe net~fe12 l:JAit 2.

The containment and engineered safety features for both uni ts have been designed and eva l u-ated at the Unit 2 maximum power rating of 3570 Mwt.

Most postulated accidents have been evaluated at 3423 MWt.

Loss-of-coolant accidents and those postulated accidents having offsite dose consequences have been analyzed at the power rating of 3570 Mwt.

SGS-UFSAR 1.0-1 Re vision 0 July 22, 1982

TABLE 4.1-lA (Sheet l of 6)

REACTOR DESIGN COMPARISON TA3~E Thenna 1 and Hydraulic Design Parameters

1.

Reactor Core Heat Output, MWt

2.

Reactor Core Heat Output, Btu/ht~

3.

Heat Generated in Fuel, ~:

4.

System Pressure, Nominal, psi a

5.

System Pressure, Min. Steddy State, psi a

6.

Minimum DNBR at Nominal Con-ditions Typical Flow Channel, Thimble (Cold Wal 1) Fl ow Chann2l

7.

Minimum DNBR for Design Transients Cool ant Fl ow

8.

Total Thennal

9.

E ff ecti ve

Transfer,

10. Effective

Rods, SGS-UFSAR for Heat Along Fuel

Velocity, Salem Jni t 1 17xl7. Fuel As.semb1y With Densificution Effects 3338 11,393 \\ 106 97.4 2.31

1. 86

> 1. 30 132.3,{. 106 i26.4,i.. 10°

) J.. I.

15. 3 2.47 x

• ~, 'J

.L'.* ;

D E\\....t:: \\t:

f fects 3338 11, 393 ;<

~'J 6 -

37.4 aso 222J

> 1. 30 134.1 ;<. i06 128.0 x io6

51. 2 15.5 2.50 x 106 Re vision O July 22, 1~82

TABLE 4.1-lA (Sheet 2 of 6)

REACTOR DESIGN COMPARISON TABLE

• The nna l and Hydraulic De~ign Parameters Cool ant Temperature, °F

13. Nominal Inlet

14. Average Rise in Vessel

15. Average Rise in Core

16. Average in Core

16. Average in Vessel Heat Transfer

18. Active Heat Transfer, Area, ft2

19. Average Heat Flux, Btu/hr ft2

20. Maximum Heat Flux for Operation, Btu/hr-ft2

21. Average. Thennal

22.

Nonnal Operat* n, kw/ft

23.

kw/ft protection set-

• points, w/ft.

24. Heat ux Hot Channel Factor SGS-UFSAR Sa 1 em Uni t 1 Sa 1 em U i t 1 17xl7 Fuel Assembly 15xl5 F el Assembly With Densification Effects 59, 700 185, ?GO 430, 900[b J 5.33 12.4[b]

18.0[d]

ification Effects

-544.4 63.9 66.6 579. l 5 76. 3 52, 200 212,600 580,000 6.88 18.8 2.16 Revision O July 22, 1982

TABLE 4.1-lA (Sheet J of 6)

REACTOR DESIGN COMPARISON TABLE Thennal and Hydraulic Design Parameters Fuel Central Temperdture, °F

25. Peak at 100 Percent Power

26. Peak at Maximum Thennal Output for Maximum Overpower Trip Point Core Mechanical Design Parameters Fuel Assemblies

27. Design

28. Number of Fuel Assemblie

29. uo2 Rods per Assembly

30. Rod Pitch, in.

31.

32.

2), pounds 3 3

• Z i re a 1 o y '../ e i g

, 1 b s *

34. Number of Gr, ds per Assemoly 35.,Loading Fuel Rods

36.

Diameter, in.

.J I 1..:... 1~ lw I

~ o

~c..A. ~ J I I I* )

j 1 -I_.

l, 2, (and 3)

SGS-UFSAR Salem Unit 1 17xl7 Fuel Assembly

~ith D~nsification Effects j

1~CC Canless 193 264 0.496 8.42G x 8.426 222,739

50. 913 8-Type R region non-unifonn

J0,952 0.374 L).0065 3

4250 RCC Canless 193 204 0.SGJ

• 8.426 x 8.426 215,400 48,250 7-Type L region non-unifonn 39,372 0.422 o.ou75 (O.uoasi RevisionO July 22, 1982

TABLE 4.1-lA (Sheet 4 of 6)

REACTOR DESIGN COMPARISON iABLE Core Mechanical Design Parameters Fuel Rods (Cont'd)

39. Cl ad Thickness, in.

40

• Cl ad Mate ri al Fuel Pellets

41. *Material

42. Density ( % of Theo re ti cal)

43. Diameter, in., Regions 1, 2, (and 3)

44. Length, in.

Rod Cluster Control

45. Neutron Absorber

46. Clad Material

47.

48. Number of 49 ** Number of SGS-UFSAR per

. ~

.-~.

Sa 1 em Uni t 1 17xl~ Fuel Assembly With Densification E f fee ts 0.3225 0.530 Ag-In-Cd Type 304 SS-Cold \\.larked 0.0185 53

-24 Effects 0.024 Zircaloy-4 uo2 Sintered 94 0.3659 (0.3649) 0.600 Ag-In-Cd Type 304 SS-Cold Worked 0.019 53 20 Re vision 0 July 22, 1982

TABLE 4.1-lA (Sheet 5 of 6)

REACTOR DESIG~ COMPARISON rABLE Core Mechanical Design Parameters Core StructLi re

50. Core Barrel, I.D./O.D., in.

51. The rma 1 Shiel d I. 0. /0. D., i n.

Nuclear Design Parameters Structure Characteristic:

52. Core Diameter, in. (Equivalent)

53. Core Average Active Fuel Heigh,

in.

Reflector Thickness and

54.

55.

56. Side - *..idter pl s Steel, in.

57. H2o;u, r.lolecu ar t<atio, Lattice (co J)

Feed Enri c w/o SGS-UFSAR Salem Unit 1 17xl7 Fuel Assemoly

_With Densificdtion E frects 132. 7 143. 7

-10

-10

-15 L.41 2.25

2. 80

..) *WU 148. U/152. 5 158.5/164.0 132. 7 144

-10

-10

-15 2.52 L.25

2. 80

..J*...JU Re vi si on 0 July 22, 1982

TABLt 4.1-!A (Sheet 6 of 6)

REACTOR DESIGN COMPARISON TABLE tJuclear Design Parameters Salem Unit 1 17xl7 Fuel Assembly With Densification Ef_fects Fuel Assembly Without Densification Effects a Previous

, the value of 2.09 for a limiting ~ypical channel was quoted

,only si ce the thimble (cold wall) DNB tests were incomplete.

[b] This mit is associated with the value of Fo = 2.32.

[c] Incl des the effect of fuel densification.

[d] Se Section 4.3.2.2.6.

SGS-UFSAR Revision 0 July 22, 1982

TABLE 4.1-~ (Sheet 1 of 6)

REACTOR DESIGN COMPARISON TABLE Thennal and Hydraulic Design Parameters

1.

Reactor Core Heat Output, MWt

2.

Reactor Co re Heat Output, Btu/hr

3.

Heat Generated in Fuel,%

4.

System Pressure, Nominal, psi a

5.

System Pressure, Min. Steady State, psia

6.

Minimum DNBR at Nominal Condi-tions Typical Flow Channels, Thimble (Cold Wall) Flow Channel.

7.

Minimum DNBR for Design Transients Coolant Flow

8.

Total Thennal Flow Rate, 1b/hr

9.

Effective Flow Rate for Heat Transfer, 1 b/hr

10. Effective Flow Area for Heat Transfer, ft2

11. Average Velocity Along Fuel

• Rods, ft/sec

12. Average Mass Velocity, 1 b/hr-ft2 SGS-UF~R Sal Elli l:fn i L 2 Sal e n1 U11 i t 2 17xl7-Fuel Assembly 15xl5 Fuel Assembly With Densification Without Effects Densification Effects 3411 11,642 x 106 97.4 2250 2220 2.24 1.80

> 1.30 132.2 x 106 126.3 x 106

51. l 15.4 2.47 x 106 3411 11,642 x 106 97.4 2250 2220 2.oCaJ

> 1.30 134. 0 x 106 128.0 x 106

51. 2 15.6 2.50 x 106 Re vision 0 July 22, 1982

TABLE 4.1-lX (Sheet 2 of 6)

REACTOR DESIGN COMPARISON TABLE Thennal and Hydraulic Design Parameters Coolant Temperature, °F

13. Nominal Inlet

14. Average Rise in Vessel

15. Average ~ i se in Core

16. Average in Core

17. Average in Vessel Heat Transfer

18. Active Heat Transfer, Surface Area, ft2

19. Average Heat Flux, Btu/hr-ft2

20. Maximum Heat Flux for Nonnal Operation, ~tu/hr-ft 2

21. Average Thermal Output, kw/ft

22. Maximum Thennal Output for Nonnal Operation, kw/ft

23. Peak 1 i near power for deter-mi nation of protection set-points, kw/ft SGs..:uFSAR Sale.ff! YRit 2 Saleffl l:J11it 2 17xl7 Fuel Assembly 15xl5 Fuel Assembly

-With Densificati6n Without Effects Densification Effects 545.0 65.8 68.7 581.0 577. 9

59. 700 189,700 440,200[b]

5.44 12.6[b]

18. oCc J 545.0 65.1 67.8 580.4 577. 5 52, 200 217, 200 580,000 7.03 18.8 Re vi si on O July 22, 1982

TABLE ~.1-1( (Sheet 3 of 6)

REACTOR Di:SIGN COMPARISON TABLE Thennal and Hydraulic Design Parameters Heat Transfer (Cont'd)

24. Heat Flux Hot Channel Factor, FQ Fuel Central Temperature, °F

25. Peak at 100 Percent Power

26. Peak at Maximum Thennal Output for ~aximum Overpower Trip Point Core Mechanical Design Parameters Fuel Assemblies

27. Design

28. Number of Fuel Assemblies

29. uo2 Rods per Assembly 30..Rod. Pitc11, in.

31. Overall Dimension, in.

32. Fuel Weight (as UO?), pounds

33. Zircaloy Weight, lbs.

34. Numoer of Grids per Assembly

35. Loading Teclini que SGS-UFSA.R 17xl7 Fuel Assembly 15xl5 fuel Assembly

-With Densification Without 3

Effects Jensification Effects 2.32[cJ 3400 4150 KCC Canless 193 264 0.496 8.426,t.. 8.426 222,739 50 '913 8-Type R region non-unifonn 3

2.40 425U RCC Canless 193

~04 0.563 8.42G x 8.426 215' 400 48,250 7-Type L region non-uniform_

Re vi si on 0 July 22, 1982

TABLE 4.1-1X (Sheet 4 of 6)

REACTOR DESIGN COMPARISON TABLE Core Mechanical Design Parameters Fuel Rods

36. Number

37. Outside Diameter, in.

38. Diametral Gap, in., Regions 1, 2, (and 3)

39. Clad Thickness, in.

40. Clad Material Fuel Pellets

41. Material

42. Density(% of Theoretical)

43. Diameter, in., Regions 1, 2, (and 3)

44. Length, in.

Rod Cluster Control Assemblies

45. Neutron Absorber 46.*Cladding Material

47. Clad Thickness, in.

48. Number of Clusters SGS-UFSAR Saleffi URit 2 Saleffi URit 2 17xl7 Fuel Assembly 15xl5 Fuel Assembly With Densification Effects 50,952 0.374 0.0065 0.0225 Zircaloy-4

. uo 2 Sintered 95 0.3225 0.530

.Ag-In*-Cd Type 304 SS-Cold Worked 0.0185 53 Without Densification Effects 39,372 0.422 0.0075 (0.0085) 0.0243 Zi real oy-4 uo2 Sintered 94, 93, 92 0.3659 (0.3649)

0. 600 Ag-In-Cd Type 304 SS-Cold Worked 0.019 53 Revision 0 July 22, 1%2

TABLE 4.1-1~ (5 of 6)

REACTOR DESIG~ COMPARISON TABLE

~aleffi URit 2 Saleffi UMit 2 17xl7 Fuel Assembly 15xl5 Fuel Assembly

-INith Densification Without Core M.:cl1ani cal Design Parameters Effects Densification Effects Rod Cluster Control Assemblies (Cont'd)

49. Number of Absorber Rods per Cl uste*r Core Structure

50. Core Barrel, I.D./O.D., in.

51. The rma 1 Shi e.l d, I. D. /0. D., in.

Nuclear Design Parameters Structure Characteristics

52. Core Di_ameter, in. {Equivalent)

53. Core Average Active Fuel Height, in.

Reflector Thickness and Composition

54. Top - Water pl us Stee 1, in..

55. Bot too - Water plus Steel, in.

56. Side - Water plus Steel, in.

57. H2o;u, Molecular Ratio, Lattice (cold)

SGS-UFSAR 24 148. 0/ 152. 5 158.5/164.0 132. 7 i43.7

-10

-10

-15 2.41 20 148.0/152.5 158. 5/164.0 132. 7 144

-10

-10

-15 2.52 RevisionO July 22, 1982

TABLE 4.1-lx (6 of 6)

REACTOR DESIGN COMPARISON TABLE Nu clear Design Parameters I CeJ Feed Enrichment, w o

58. Region 1

59. Region 2

60. Region 3 UtJ\\i 1./A..\\N\\l' ZI UN\\°'t"~ l. -4. 2' Sale~ YAit 2 Saleffi UAit 2 17x17 Fuel Assembly 15x15 Fuel Assembly

-With Densification Effects 2.:i. s I 2.10 z.eo/ 2. 60

'3.30/

3.10 Without Oensification Effects

~.25 2.80 3.3l}

[a] Previously, the value of 2.09 for al imiting typical channe.l was quoted only since the thimble (cold w*all) DNB tests were incomplete:

[b] This limit is associated w-ith the value of ro = 2.32.

[c] Includes the effect of fuel densification.

[d] See Section 4.3.2.2.6.

[ej G~~.h. i &

u'Z. \\

SGS-UF~R Revision 0 July 22, 1982

4.4.2.l Summary Comparison The design o~ the Salem Unit 1 and Unit 2 reactors with the 17 x 17 fuel rod array per. assembly has the following identical thennal and hydraulic parameters-as the 15 x 15 fuel rod array reactor design.

1. Core power

2.

System pressure

3.

Coolant inlet temperature

4.

Open_ lattice fuel rod array The vessel loop flow rates for both Unit 1 and Unit 2 thermal design are approximately 1.4 percen~ less than the 15 x 15 design valves.

The basis for this change is dis~ussed in Chapter 5 ** This change in flow also results in small changes in the core and vessel coolant average temperature and core and vessel coo_l ant exit temperatures.

\\~ -4..+-l Values of each parameter 3re presented in Tfles q.4 lA* aAE! 8 for all

..,...~ 4*.

-~

coolant loops in service an~ in Taeles 4.4 2A aAE! B for all but one coolant loop in service. It is also noted that in this power capability evaluation, there has not been any change* in the design criteria. The reactor is still designed to a minimum DNBR ~ 1.30 as well as no fuel centerline melting during nonnal operation, operational transients and faults of moderate frequency.

Whe1e applicable th~ Figu1es and fable! in t~is-section con!ist ef twa f:'al"ts laeeleet "A" afle "B", *111c!ielc! l"efel" te UAits 1 aAE! 2: re Sf:'eetively

• SGS-UFSAR 4.4-5 Revision 0 July 22, 1982

adequate heat transfer is provided between the fuel clad and the reactor.

coolant so that the core thermal output is not limited by considerations of the clad-temperature.-

Figure 4~4-4 shows the axial variation of average clad temperature for the average power rod both at beginning and end-of-life.

Treatment of Peaking Factors The total heat flux hot channel factor, FQ, is defined by the ratio of the maximum to core average heat flux.

As presented in Table 4.3-2 and discussed in Section 4.3.2.2.1, the design value FQ for normal opera-tion is 2.32, including fuel densification effects.

This results in peak local power~ of 12a4 k*wlft arH! 12.6 kw/ftx ~

tll'lits 1 aAel 2 Fes~eti"ily, at full power conditions.

As described in Section 4.3.2.2.6 the peak local power at the maximum overpower trip point is 18.0 kw/ft.

The centerline temperature at this kw/ft must be below the uo2 melt temperature over the lifetime of the rod, including allowances for uncertainties.

The melt temperature of unirradiated U02 is 5080°F[l] and decreases by 58°~ per 10,000 MWD/MTU.

From Figure 4.4-2, it is evident that the centerline temperatures at the maximum overpower trip points for both units are far below those re-quired to produce melting.

Fuel centerline and average temperatures at rated ( 100 percent). power and at the maximum overpower trip point are

~

44-l...

presented in Tables 4.4 lA aAEi Ba

• I T'~p\\~

4.4.2.3 Critical Heat Flux Ratio or Departure from Nucleate Boiling Ratio and Mixing Technology The minimum DNBR 1 s for the rated power, design overpower and anticipated

-""\\'"'<l.>-&_," 44-.

transient conditions are given in Tael-e54:Q ~. al"la lB.

The core aver-age DNBR is not a safety related item as it is not directly related to the minimum DNBR in the core, which occurs at some elevation in the limiting flow channel.

Similarly, the DNBR at the hot spot is not SGS-UFSAR 4.4-13 Re vision O July 22, 1982

I main parameter which affects the DNBR.

If the Salem Units 1 and 2 were ope:rating at full power and nominal steady state conditions as specifiea

\\~

44-\\...

in Taeles ~.4-lA aRe 8-; a reduction in Jocal mass velocity of 72 ~eiceeRt

~

6~ percerit res13e&tivelyX would be required to reduce the DNBR frolil Lbe aAEi 1.80 to 1.30.

The above mass vela:city _effect on the DNB cor-relation was*based on the assumption of fully developed flow along the full channel length.

In reality a lo1;al flow blockage is expected to promote turbulence and thus would likely not effect DNBR at all.

Coolant flow blockages induce local crossflows as well as promote turbulence.

Fuel rod behavior is changed under the influence of a sufficiently high crossflow component.

Fuel rod vibration could occur, caused by this crossflow component, through vortex shedding or turbulent mechanisms.

If the crossflow velocity exceeds the limit established for fluidelastic stability, large amplitude whirling results.

The.limits for a controlled vibration mechaRism are established from studies of vortex shedding and turb~lent pressure fluctuations.

Crossflow velocity above the established limits can lead to mechanical wear of the fuel rods at the grid support locations.

Fuel rod wear due to flow inducea vibration is considered in the fuel rod.fretting evaluation (Section 4.i).

<t.4.4 TESTING AND VERIFICATION 4.4.4.1 Tests Prior to Initial Criticality A reactor coolant flow test, as noted in Item 5 of Tdble 13.3-1,. is performed following fuel load1ng but prior to initial criticality.

Cool~nt loop pressure drop data is obtained in this test. This aata in conjunction with coolant pump perfomance information allows determina-tion of tl1e coolant flow rates at reactor operating conditions.

This test verifies that proper coolant flow rates have been used in the core thermal and hydraulic ilnalysis.

SGS-UFSAR 4.4-54 Re vision 0 July 22, 1982

TA:3LE

~..+-lA I 5,-:cct

~ Jf ~)

~EACTUR JESIGN CCMPARISJN

~~BL~ SA~~~ J~IT l Thenna l jnd Hydraulic Design Par-irneters Re:lctJr Core Heat Output, MWt Reactor '.:ore Heat Output, Btu/hr

~e~ t Jt=nerated in Fue 1,

S;stem ?ressure, Nominal, psia System Pressure, Minimum Steady State, psi a

~inimu~ JNBR at Nominal Condittons T;pical

~low Channel Thimble (Cold-..;all) Flow Channel Mini~tirn JNBR f~r Desi~n Trdnsients iJNa C0rre1ation 1.:0:: 1 1 n t Fl ow

~ff2c:i~e ~low ~ate for

-rans fer, 1 o/hr

~ffecti 1e ~low Area T r*a n s f e r, f t 2 Fuel Rods, ft/sec 1 b/hr-f t2 Coo:.: - - -

~~omi na ~

inVessel,°F in Core, °F SGS-i.JFSAR 17 x 17..Jith

Jensification 333d 11,393,(

2.31 1.86

> 1. 30

" R - G ri d " ( ;.I - 3 with modified spacer factor) 132. 3..< 106 126.4 x 1Q6

51. l 15.3 2..+ 7 x 106 544.4 o..i.. 7 67.5 3338 1:, J9J,(

~JG

? 7. 4 22:.0 2220

> l. 30 "R-Grid" l..i-J

• ~i th modi fi e'J Si:ldCcr factor-)

134.1 x liJ 0 128.0,( liJO

51. 2 15.5 2.50 x 106 544.4 6 3. 'j 66.6 Revision 0 July 22, 1982

TAaLE 4.4-lA (s~eet '

).

REASTJR JES~GN CQMPAR!SO~

Ther::ia l *ind Hzdra 1J 1 i c Desi~n Parameters Average in Core, OF Average in Vessel, 0.-r Heat -rans fer Active Heat Transfer, Surf ace Avera3e Heat Flux, 3 tu/hr-ft2 i*1ax i rilur11 !-lea t F"l ux, for no nna l operation 3tu/nr-ft2 Aver-~ge Tnermal uut,JUt, k'.-1/ft

lax irnum Thermal,J°utput, for rio~a1 o;Jeration, k

• w/ft

Area, Peak l..i:-:ear ?ower for 1etennination

?rotection Setpoints, kw/ft

?ea~ at 100

?ower, 'F

?eak dt 7he~a1 'J1jtput Maxi m for

>laximum Overpower Trip P int, °F Ac ro s 5

.;c.,..Js~:;

f t2 psi I

. ~. '

~I A ~7,..i ::1 J e 'l s i f i 1: 1 1: i c n

~

"J i 'J

• i 3.33 12.i.+LDJ is.oCcJ 2~. 7.. 2.sCd~

4-'J.:3... 3.J

,, i :.

I J.

.J *.J j2,2JJ Jc.:. *...J Cb J CcJ

[d]

l.2 j the *1alue of 2.09 for a.li~itin*; :/;JiCdi :::i;a11-:-=:

.... l; qJ~:.::: en y since 1:he thimble (cold *"al 1 ) J.d *~-::s:s ;1ere i.11.~);:ip~ ::~.

This li, tis associated *"'ith the value o7

~ = 2.32.

See Se

~ion 4.3.2.2.ci.

on best estimate reactor flow rate of 95,600 j,Jm/looo.

Pre ously, a conservatively high value of pressure JroiJ NdS used to de ermine vessel loop flow rates.

SGS-UFSAR Revision O July 22, 1982

TAoLE 4.4-iX \\sheet 1 of 2)

REACTOR DESIGN COMPARISON TABLE SAL.:. 11 ~.J :-;-

c..

Thermal and Hydraulic Design Parafileters.

Reactor Core i-ieat Output, MWt Reactor Core Heat Output, Btu/hr Heat Generated in Fue 1,

System Pressure, Nominal psia System Pressure, ~~inimum Steady State, psi?

Minimum D~SR at ~ominal Conditions Typical Fl ow :hannel Thimble (Cold..iall) Flow Channel Minimum DN3R for Design Transients DNB Cor"'el Jti on Cooi ant Fl G'fl Total Therfilal F1 ow Rate, lb/hr Effecti v.c ;::-18'" 'ate for Heat T;a;1sFer*, 1 J,.*nr Effecti~e ~low Area for Heat Trarisfer, ft2 Avera*]e l~'ocit; Along Fuel

Rods,

,. s2c Avera~~ :1.J;~.'elocitj, lb/hr-ft2 Nominal rn1et, °F Average i\\ise ;:1 '/essel, °F SGS-UFSAR 17 x 17 ;.Jith Densification 3411 11,642 x 106 97.4 2250 2220 2.24 1.80

>1.30 "R-Grid" (r-J-3 11i th modified spacer factor) 132.2 x io6 126.3 x 106

51. 1 15.4 2.47 x 106.

545.0 65.8

-~ '.l 15 x ~ 5..Ji ti1ou t Densification 3411 11,642 x 106 9 7. 4 2250 2220 2.0[aJ

> l. 30

"?.-Grid"

(.~-3 with filOdi fi ed spacer factor) 134.0 x 106 128.J x 106

51. 2

15. 6

2. 50

.<: 106 545.J 65.1 Revision O July 22, 1982

-,. -,, *x

. **Lt.-.."-'*

. ("'\\,...)

( s he et t. *Jf 2 !

Ther:!lal a:;d

~ydr-3!.Jl ic Design ?:lrarneters.

Avera::ie i 'l Core, °F Aver::ige in '/esse 1, °F Active Heat Transf~r, Surface Area, ft2 Average ~eat Flux, Btu/hr-ft2

.*1a,.< ii:iun -l21 t r-; 'j x ' for *10r~a1 00e."'at::;*1, 3:u/~r-ft2

,'.,ve:"l;e

-:-~2~*:;ia1.J:J:JJt, <*tJ/ft

a:<i;:iu~n ir:'=r:-11 0Jtput-, for

1 0 r~ J 1 0 '.) e r a : i 0 n <: "' / f t J2ci<... i::~i*' Jower fJr :eten;1;r:ach., Qf

?ro':::*:o:..~on 3etpoin:s, :<*,.,,/ft

?ea~ d!; -,,er*'.,,:

• ~,_'::::..it *~aximum for

"\\L<i;:i:~:n -:!ver*: _..;-::r' -;-rip Point, °F

.~ c r::*:; s *= ') r~ '

"J 3 j 0

:e>

ncl udi ng nozzle, ps~

17 :< 17 Aith Jensification 581.0 577. 9 59,700 189,700 440 200[0 J 5.H

, 2 -- ~ J

.L

• b-

~a.oCcJ 3400 24.7 + 2.5[d]

49.'3 + *).0 l 5 x 15..ii thou t Je'lsification sao. 4 577. 5 52,21.)l) 217,200 S80,OJO

7. '] 3 13.3 4250 J2.s[e]

J <:'.. J

~a ~ _, *

~-o ' 1 ~*.., e / 3 1 *~ e of 2

• G 9

~*~.:-.*.:*~

• ~-:~./

.

~C-?

~fi-2

':..'li:*:~Jle

-), >': *-.-,;~ ':,

soc;*~ted..;i'

.~

for a l irniting ':J:J'c3~ channel *,.,,.:is

~oi*J wall) J

;d :ests*..;ere incomplete.

he value of FQ = 2.32.

_CJ Se.c:

~~c~ion ~-~.2.2... ).

Cdj

)d;~*

i _*:o jes: es':.i;.:a*:e r'edcto~* Flow rate of 35,600 ::i1n/lJ0p.

_c~ ?r<? 1~ :~*:;;y, a '.:Jnservdti vely rii ~n 1abe of ;Jressur'? ~drop '""as used to Je:*~!"".1i ~e '1'25se1 l ccp

~* c*,... ra:es.

SGS-IJ~SAR Re vision 0 J u, y 2 2 ' l 98 2

\\

TABLC: 4.4-ZA THERi"1.~L-HYORAULlC lJES IGi~ PAK.Ai*lEiEi 1. 86

> 1. 30 Revision 0 July 22, 1982

Core Hot Subchannel SGS-UFSAR TABLE 4.4-3A VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS WITH DESIGN HOT CHANNEL FACTORS SALEM UNIT 1 Maximum 2.2 Revision 0 July 22, 1982

TAdLC: 4.4-2X THERMAL-HYDRAULIC OESIGN PARAHETtRS FOR ONE OF FOUR COOLANT LOOPS OUT OF SERVICE SALEM UN If z Total Core Heat Output, MWt Total Core Heat Output, 106 Btu/hr neat Generated in Fuel, Nomi na 1 System Pressure, psi a Coolant Flo\\-1 Effective Thermal ~low Rate for Heat Transfer, 106 1 bs/hr Effective ~l~w Area for Heat Transfer, ft2 Average Velocity along Fu~l Rods, ft/sec Average Mass Velocity, 106 1 b/hr-ft2 Coo 1 ant Temperature, °F Desi3n Nominal Inlet Average Rise in Core Ave rage in Co re Ac:ive ~e3t fransferSurface 2388 8150 97.4..

2250

90. 6

51. l

10. 9

1. 77 539.1 68.2 574.7 Ar::1. ft 2 59,700

~ver~;e ~edt Flux, Btu/hr-ft2

~inimum ONB Rat~o dt Nominal Conditions 1;...,.: *~ *. _,

I\\::

) ~-.;.,.

-., f

~~~*: ~..,

and Anticipated rransients SGS-UFSAR 132,900

> 1.80

> l.3U Revision O July 22, 1982

\\

Core Hot Subchannel SGS-UFSAR TABLE 4.4 VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS WITH DESIGN HOT OiANNEL FACTORS mi" EM !ltll I 2 Average p.18%

4.0%

Maximum 13.6 Revision O July 22, 1982

\\

TAt3LE S.1-:.

SYSTEM DESIGN AND OPERATING PARAl~ETERS Plant design life, years Number of heat transfer loops Design pressure, psig Nominal operating pressure, psig Total system volume including pressurizer and surge line (ambient conditions), ft3 System liquid volume, including pressurizer and surge line (ambient conditions), ft3 Total heat output (100 percent power), Stu/hr Reactor vessel coolant temperature at foll power:

Inlet, nominal, °F Ot1tl et, 0 F' Coolant temperature rise in vessel at fui 1 power, avg, °F Total coolant flow rate, lb/hr Steam pressure at full ~ower, psia SGS-UFSAR tt 11 i t 1

~

2455 2239 12,61~

11, 8~2 u,q;;a 1' 544.4 609. 3 eiJ.9 io6 116

Z:::ld:t )( ""

t:IFI ~ ~

'}

40 4

2435 2235

• 12,612 11,892 11, 68u 545.0 610.2 x i06 6:3.2 1 l.Z

• t X 10 G ldd:9 x 10 805 Revision 0 July 22, 1982

I I

I TABLE 5~2-3 (Sheet 1 of 2)

REACTOR VESSEL DESIGN DATA Design/Operating Pressure, psig Hydrostatic Test Pressure, psig Design Temperature, °F Overall Height of Vessel and Closure Heat, ft-in. (bottom head OD to top of control rod mechanism adapter Thickness of Insulation, ~in., in.

Number of Reactor Closure Head Studs Diameter of Reactor Closure Head Studs, in.

ID of Flange-, in.

.OD of Flange, in.

ID at Shell, in.

Ul"li t 1 248S/22de H}.

+

-&+-

+-

172.S*

Inlet Nozzle ID, in 27 1/2 Outlet Nozzle ID, in.

~

Clad Thickness, min., in.

5/32 Lower Head Thickness, min., in. (base metal) 5 3/8 Vessel Belt-Line Thickness, min., in.

(base metal)

-&: Closure Heat Thickness, in. -

Reactor Coolant Inlet Temperature, °F 544.4 Reactor Coolant Outlet Temperature, °F Reactor Cool ant Fl ow, 1 b/hr Total Water Volume Below Core, ft3 Water Volume in Active Core Region, ft3 SGS-UFSAR 608.3 134.1 )( 106 U11i t 2 2485/2235 3107 650 43-10 3

54 7

172.5 205 173 27-1/ 2 29 5/32 5-3/8 8.5 7

545.0 610.2 132.2. ~ 10' 133 *. 9 }( 106-1050 665 Revision O July 22, 1982

TABLE 5.2-3 (Sheet 2 of 2)

REACTOR VESSEL DESIGN DATA Total Water Volume to Top of Core, ft3 Total Water Volume to Coolant Piping Nozzles Centerline, ft3 Total Reactor Vessel Water Volume, (~lith core and internal sin place), ft3 Total Reactor Cool ant System Vo 1 ume, ft3 Unit 1 2164 2929

-4~45 12,612 U11i t 2 2164 2959 4945 12,612 i>ELTi\\E ~~ NO'TE w ~E~ DO\\ "l~ ht,""t'U.A-L F~~ c.,~Cl.N\\..~

f&J.~'t&.~~

"'CAr4_ ~~~~~I 29 '2 ~ ~

l I 0"1.A-~ fY\\d ()~ JU..\\)~-t.& ~(YV"

~\\-t.NA.l A.. ~fOcpl\\9\\\\\\i~ -LM01.* T~ ~

~

ls. 2q sq ~

s:~ G\\.A u.....v.. r 1.'t.

I SGS-UFSAR Revision O July 22, 1982 I

I

' /

.;;-... :r:. -~*.~-.;~--

TABLE 5.2-5 (Sheet 1 of 2}

STEAM GENERATOR DESIGN DATA*

(Model 51) tin f t l Number of Steam Generators

+

Design Pressure (Reactor coolant/ste~), psig 2485/1.005 Reactor Coolant Hydrostatic Test Pressure (tube side-cold), psi g 3107 Design Temperature (reactor coolant/steam), °F 650/688 Reactor Coolant Flow, lb/hr Total Heat Transfer Surface Area, ft2 Heat Tran sf erred, Btu/hr Steam Conditi.ons at Full. Load, *Outlet Nozzle:

St~am Flow, lb/hr Steam Temperature, °F Steam Pressure, psig Maximum Moisture Carryover, wt percent Feedwater, °F Overall Height, ft~n.

33. 53 x 1:86 s1, sec 28§7 )( 1Q6

~.81 x l(J6

~

~

~

~

87 8

<tfoitf 4

2485/1085 3107 650/EiOO

'33.0S.ic 'O <.

-li:47 )( 196..

51,500 2920 x 106 3.7 4 x 106 519 805 0.25 435.

67-8 Shell OD (upper/lower), in.

175 3/4 I 135 175-3/4 I 135 Number of U-tubes U-tube OD, in.

Tube.Wall Thickness (minimum), in.

Number of Manways/ID in.

Number of handholes/ID, in.

• Quantities are for each steam generator SGS-lJFSAR

~

0.875 9.0§9

.4;'16 4f+

3388 0.875

o. 050 4/16.

2/6 Revision 0 July 22, 1982

TABLE 5.2-5 (Sheet 2 of 2)

STEAM GENERATOR DESIGN DATA*

(Model 51)

Reactor Coolant-Water Volume, ft3 Primary Side Fluid Heat Content, Btu Secondary Si de Water Vo 1 ume, ft3 Secondary Side Steam Volume, ft3 Secondary Si de Steam F1 ui d Heat Content, Btu

• Quantities are for each* steam generator l:Jrd t 1 Rated Load 1080 28.7 x 106 1838 4030 S.738x 107 l:Jni t 2 No Load 1080

27. 7 x 106 3524 2344 9.628 x 107 SGS-UF~R Revision o July 22, 1982

---*:.... -~---=---"'---

TABLE 5.2-7 REACTOR COOLANT PIPING DESIGN PARAr.fTERS Reactor Inlet Piping ID, in.

Reactor Inlet Piping Nominal Thickness, in.

Reactor Outlet Piping ID, in.

Reactor Outlet Piping Nominal Thickness, in.

Coolant Pump Suction Piping ID, in.

Coolant Pump Suction Piping Nominal Thickness, in.

Pressurizer Surge Line Piping ID, in.

Pressurizer Surge Line Piping nominal Thickness, in.

Design/Operating Pressure, psig Hydrostatic -Test Pressure (Cold), psig Design Temperature, °F

• Design Temperature (pressurizer surge line}, °F Water Volumej (all 4 loops including surge line} ft Design Pressure (pressurizer relief lines}, psig Design Temperature (pressurizer relief lines}, °F

~E Lt l:t::.

"°"'~ ~o\\e. * \\-'.)~

"00,NC.

~au. ~l-

~~~ U,.E\\J\\ t\\OtJ Un i t 1 27=1/f' 2.58

~

~ t--

2.6&

il. 500

-t:-2-5

~48~fh:::io 2 7-1/2 2.38 29 2.50 31 2.66

(\\)

(?..)_

2485/2235 3107 650 680 1455 (3) t],}

t:cll~ol Jt(}~*. '~\\o\\oi ~

""°~ 01."(Nl~ ~M

~ UW.T-~ J 2.LlSS It~ I.W()A A., ~11¥~

~01.. TN. (.~

~

~ \\'-\\~S-ft3.) ~

.s:~ lU ~r2'.s.

f!\\) ~

1.. ~e) ~

\\\\.~oo ~ ~

(1,) ~j_. \\. 1,'5, ~'c-"L \\,L\\ b Q

~)From pressurizer to safety valve 2485 psig 650°F From safety valve to pressurizer relief tank 600 psig 600°F.

SGS-UFSAR Re vision O Ju 1 y 2 2, 198 2

TABLE 5.5-1 (Sheet 1 of 3)

RESIDUAL I-EAT REMOVAL SYSTEM DESIGN PARAMETERS Code Re qui reme nt s Residual Heat Exchangers (Tube Side)

(She 11 Si de)

Residual Heat Removal Piping and Valves General ASr.E III, Class C ASr.E VII I ANSI 831.1.0*

ANSI *s31. 7**

Plant design life, years 40 Component cooling water supply temperature design, °F 95 Reactor cool ant temperature at startup of decay "heat removal °F 350 Time to cool Reactor Coolant System from 350°F to 140°F, starting at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after shutdown, hr 16 Ambient Refueling water storage temperature, °F Decay heat generation at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shutdown, Btu/hr 1' 70

• 0 x 1@A ( c n?t Nb

• 11 72.1x106 -HJRit Ila. 2)

.'\\

H3so3 concentration in refueling water storage tank, ppm boron Used for design.

-2000

• • For piping not supplied by the NSSS suppl.ier, material inspection fabrication and quality control conform to ANSI 831.7.

Where not possible to comply with ANSI 831.7, the requirements of ASt<E III-1971, which incorporated ANSI 831.7, were adhered to.

SGS-UFSAR Revision 0 July 22, 1982

10.2 TURBINE GENERATOR 10.2.l UESIGN BASES The Steam and Power C.onversion System is designed to convert the heat produced in the reactor to electrical energy.

Heat absorbed by the Reactor Coolant System is transferred to the feedwater in four steam generators.

Ttie feedwater system pro'1f des suffi_fi ent feedwater fl ow to the four steam generators where removal of heat from the Reactor Coolant System results in sufficient steam formation to*drive the turbine genera-tor units as follows:

~1' lOO"lo Q.EPtt:\\O(. ~OlO~

~ixiiin.11R G1:JaraRteee Rati "!t Gross Output, Mwe Anticipated Net Output, M#e Maximum L.a l cul ated Load Gross Output, Mwe Anticipated Net Output, M#e 10.2.2 SYSTEM GESCRIPTION 10.2.2.1 Turbine~Generator No. 1 Unit 1176 1130 No. 2 Unit 1158 1115 1201 1155 The turbine is a four-cas_ing, tandem-compound, six flow exhaust, 1800 rpm unit with44-inch long last stage.blades.

The turbine shaft i.*s direc*tly connected to the ac generator.

A brushl ess exciter is coupled to tile

'*

• generator.

The generator is hydrogen cooled with water-cooled stator windings.

It is rated at 1,300,000 KVA at 75 "psig hydrogen pressure, 0.90 PF, 0.48 SCX, 3 phase, 60 cps, 25 KV, and 1800 rpm.

Generator SGS-UFs.\\R 10.2-1 Revision O July 22, 1982

ANSI-831.7, Nuclear Power Piping.

w11ere not possible to comply w1ta AN~i ~J1.1, tne requirements or A~ME lll-1Y71, wnicn incorporated ANSI 831. 7, were adhereJ to.

(b) Principal System Valves:

Main Steam Safety Valves - ASfofE*aoiler and Pressure Vessel Code,Section III, Class A.

Main Steam Relief Valves - ASME aoiler and Pressure Vessel Code,Section III, *Class II (Glass I for 111aterhls, inspections, faorication and quality control).

Main Steam Stop Valves - ASME doiler and Pressure Vessel Code,Section III, Class II {Class I for materials, inspections, fabrica-tion and qualitt control).

Feedwater Isolation Valves - ASME Boiler anJ Pressure Vessel Code,Section III, Class II {Class I far materials, inspecti~ns, faurica-tion and quality control).

10.3.2 SY~TEM DESCRIPTION 10.3.2.1 Main Steam System The Main Steam System is shown in Fi~ure 10.J-1.

The Main Steam System far edch unit conveys saturated steam from four steam ~enerdtors to the ili ~i1 press4r~ turbi~~ with 1 e~s t11~r1 40 psi Tl\\e.

s;.h~.flrl'W\\

(.o"r\\&1M~ ~

~U,\\\\ \\o/\\.(i.

~

pressure _d_rf>p,.

'ffle1i1~fl eaen~e;rn:ra to1 for t11e Ne. 1 111it ; !.-

L..

A.-f>P"'111*~1!:\\_ ~.~O,COO ~

~

'r.01-V\\/ en;. r50 P..S..(~ S13a.,.

S\\-~o...>tt"\\

Gfiliigi:iea*, ta ftu*Risfl approx;rndtelJ J,60tl,d00 potrnds pc:r 1hJtff tif 7SO ~sij, 513 F stealff te t1'1e tt1reiRe, the l"t;ghe1 de3ign flow 1aLe For.tne.~o. 2 U~i t, approximately J, 7t:lt:l,OOO po..i-Ads per-floo-1"- -te tliE curbi He ror each

~team ~eRePd~iF; ~~~used for the system design of botn units.

~eheat is SGS-UFSAR 10.J-2 Revi sian 0 July 22, 1982