ML18092A758

From kanterella
Jump to navigation Jump to search
Proposed Tech Spec Revs,Reflecting Increase in Rated Thermal power.Marked-up Draft FSAR Changes Encl
ML18092A758
Person / Time
Site: Salem PSEG icon.png
Issue date: 08/06/1985
From:
Public Service Enterprise Group
To:
Shared Package
ML18092A756 List:
References
NUDOCS 8509050151
Download: ML18092A758 (38)
v • d • e


Text

APPENDIX BB TECHNICAL SPECIFICATION REVISIONS HANO MARKED

    • ( --------: - - - -

8509050151 850806 -- ' -

PDR P

ADOCK 05000272 2737e:1d/032285 PDR 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.

. ~11 OPERATIOAAL MOOE 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 1ficat1ons.

  • 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

Tl\BLE 3.2-1 ONR PARAMETERS

. LIMITS . ' I i1 Loops In

.. Operation

' PARl\Mt:HR 58'l.

~>810f He.1ctor Coolant Sy-;tem Tavg

  • 2220 ps1a*

l 1 n~c.,'iuri zer Pressure

. 349,299 ljf.n w

A

  • 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:

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

Chapter 1 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).

- 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.

ter 5 - 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

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

DE\....t:: \t: -

TABLE 4.1-lA (Sheet l of 6)

REACTOR DESIGN COMPARISON TA3~E Salem Jni t 1 17xl7. Fuel As.semb1y Thenna 1 and With Densificution Effects  :: f fects Hydraulic Design Parameters 3338 3338

1. Reactor Core Heat Output, MWt 11,393 \ 10 6 11, 393 ;< ~'J 6 -
2. Reactor Core Heat Output, Btu/ht~

97 .4 37.4

3. Heat Generated in Fuel , ~:
4. System Pressure, Nominal, psi a aso
5. System Pressure, Min. Steddy 222J State, psi a
6. Minimum DNBR at Nominal Con-ditions Typical Flow Channel, 2.31 Thimble (Cold Wal 1) Fl ow Chann2l 1. 86
7. Minimum DNBR for Design

> 1. 30 >1. 30 Transients Cool ant Fl ow 132.3 ,{. 10 6 134.1  ;<. i0 6

8. Total Thennal
9. Eff ecti ve Transfer, i26.4 ,i.. 10° 128.0 x io 6
10. Effective for Heat

) J..

I.

51. 2 Along Fuel
15. 3 15.5 Rods, Velocity,
  • ~, 'J 6 2.47 x .L'.*; 2.50 x 10
  • SGS-UFSAR Re vision O July 22, 1~82

TABLE 4.1-lA (Sheet 2 of 6)

  • REACTOR DESIGN COMPARISON TABLE Sa 1em Uni t 1 Sa 1 em U i t 1 17xl7 Fuel Assembly 15xl5 F el Assembly
  • The nna l and With Densification Effects ification Effects Hydraulic De~ign Parameters Cool ant Temperature, °F

-544.4

13. Nominal Inlet 63.9
14. Average Rise in Vessel 66.6
15. Average Rise in Core 579. l
16. Average in Core 5 76. 3
16. Average in Vessel Heat Transfer
18. Active Heat Transfer, 52, 200 Area, ft 2 59, 700
19. Average Heat Flux, Btu/hr ft 2 185, ?GO 212,600
20. Maximum Heat Flux for Operation, Btu/hr-ft 2 430, 900[b J 580,000 kw/ft 5.33 6 .88
21. Average. Thennal 22.

Nonnal Operat* n, kw/ft 12.4[b] 18.8 23.

protection set-18.0[d]

  • points, w/ft.
24. Heat ux Hot Channel Factor 2.16 Revision O SGS-UFSAR July 22, 1982

TABLE 4.1-lA (Sheet J of 6)

REACTOR DESIGN COMPARISON TABLE Salem Unit 1 17xl7 Fuel Assembly Thennal and Hydraulic ~ith D~nsification Design Parameters Effects Fuel Central Temperdture, °F

25. Peak at 100 Percent Power 4250
26. Peak at Maximum Thennal Output for Maximum Overpower Trip Point Core Mechanical Design Parameters Fuel Assemblies
27. Design 1~CC Canless RCC Canless
28. Number of Fuel Assemblie 193 193
29. uo 2 Rods per Assembly 264 204
30. Rod Pitch, in. 0 .496 0 .SGJ *
31. 8.42G x 8.426 8.426 x 8.426
32. 222,739 215,400 2), pounds 3 3
  • Zi re a1 oy '../ e i g , 1b s
  • 50. 913 48,250
34. Number of Gr, ds per Assemoly 8-Type R 7-Type L 35.,Loading j region non-unifonn 3 region non-unifonn Fuel Rods
36. :J0,952 39,372 Diameter, in. 0.374 0.422

.J I 1..:. .. 1~ lw I ~ o ~c..A. ~ J I I I* ) - j 1 -I_.

l, 2, (and 3) L).0065 o.ou75 (O.uoasi RevisionO SGS-UFSAR July 22, 1982

TABLE 4.1-lA (Sheet 4 of 6)

REACTOR DESIGN COMPARISON iABLE Sa 1em Uni t 1 17xl~ Fuel Assembly With Densification Ef fee ts Effects Core Mechanical Design Parameters Fuel Rods (Cont'd)

39. Cl ad Thickness, in. 0.024 Zircaloy-4 40
  • Cl ad Mate ri al Fuel Pellets
41. *Material uo 2 Sintered
42. Density ( % of Theo re ti cal) 94
43. Diameter, in., Regions 1, 2, 0.3225 0.3659 (0.3649)

(and 3) 0 .530 0.600

44. Length, in.

Rod Cluster Control Ag-In-Cd Ag-In-Cd

45. Neutron Absorber Type 304 Type 304
46. Clad Material SS-Cold \.larked SS-Cold Worked 0.0185 0.019 47.

53 53

48. Number of 49 ** Number of per

-24 20

  • SGS-UFSAR Re vision 0 July 22, 1982

. ~ .-~.

  • TABLE 4.1-lA (Sheet 5 of 6)

REACTOR DESIG~ COMPARISON rABLE Salem Unit 1 17xl7 Fuel Assemoly Core Mechanical Design _With Densificdtion Parameters Efrects Core StructLi re

50. Core Barrel, I.D./O.D., in. 148. U/152. 5
51. The rma 1 Shiel d I. 0. /0. D. , i n. 158.5/164.0 Nuclear Design Parameters Structure Characteristic:
52. Core Diameter, in. (Equivalent) 132. 7 132. 7
53. Core Average Active Fuel Heigh ,

in. 143. 7 144 Reflector Thickness and

54. -10 -10
55. -10 -10
56. Side - *..idter pl s Steel, in. -15 -15
57. H2o;u, r.lolecu ar t<atio, Lattice (co J) L.41 2.52 Feed Enri c w/o 2.25 L.25
2. 80 2. 80

..) *WU ..J* ...JU Re vi si on 0 SGS-UFSAR July 22, 1982

TABLt 4.1-!A (Sheet 6 of 6)

REACTOR DESIGN COMPARISON TABLE Salem Unit 1 17xl7 Fuel Assembly Fuel Assembly With Densification Without tJuclear Design Parameters Ef_fects 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 Sal Elli l:fn i L 2 Sal e n1 U11 i t 2 17xl7- Fuel Assembly 15xl5 Fuel Assembly Thennal and With Densification Without Hydraulic Design Parameters Effects Densification Effects Reactor Core Heat Output, MWt 3411 3411 1.

Reactor Co re Heat Output, Btu/hr 11,642 x 10 6 11,642 x 10 6 2.

Heat Generated in Fuel,% 97.4 97.4 3.

System Pressure, Nominal, psi a 2250 2250 4.

5. System Pressure, Min. Steady State, psia 2220 2220
6. Minimum DNBR at Nominal Condi-2.24 2.oCaJ tions Typical Flow Channels,
  • Thimble (Cold Wall) Flow Channel. 1.80 1.30
7. Minimum DNBR for Design ->

1.30 ->

Transients Coolant Flow

8. Total Thennal Flow Rate, 1b/hr 132.2 x 106 134. 0 x 10 6
9. Effective Flow Rate for Heat Transfer, 1 b/hr 126.3 x 10 6 128.0 x 10 6
10. Effective Flow Area for Heat Transfer, ft2 51. l 51. 2
11. Average Velocity Along Fuel
  • Rods, ft/sec 15.4 15.6
12. Average Mass Velocity, 1b/hr-ft 2 2.47 x 10 6 2.50 x 10 6
  • SGS-UF~R Re vision 0 July 22, 1982

TABLE 4.1-lX (Sheet 2 of 6)

REACTOR DESIGN COMPARISON TABLE Sale.ff! YRit 2 Saleffl l:J11it 2 17xl7 Fuel Assembly 15xl5 Fuel Assembly Thennal and -With Densificati6n Without Hydraulic Design Parameters Effects Densification Effects Coolant Temperature, °F

13. Nominal Inlet 545.0 545.0
14. Average Rise in Vessel 65.8 65.1
15. Average ~ i se in Core 68.7 67.8
16. Average in Core 581.0 580.4
17. Average in Vessel 577. 9 577. 5
  • Heat Transfer
18. Active Heat Transfer, Surface Area, ft 2 59. 700 52, 200 2 189,700 217, 200
19. Average Heat Flux, Btu/hr-ft
20. Maximum Heat Flux for Nonnal Operation, ~tu/hr-ft 2 440,200[b] 580,000
21. Average Thermal Output, kw/ft 5.44 7 .03
22. Maximum Thennal Output for Nonnal Operation, kw/ft 12.6[b] 18.8
23. Peak 1 i near power for deter-mi nation of protection set-points, kw/ft 18. oCc J
  • SGs..:uFSAR Re vi si on O July 22, 1982
  • TABLE ~.1-1( (Sheet 3 of 6)

REACTOR Di:SIGN COMPARISON TABLE 17xl7 Fuel Assembly 15xl5 fuel Assembly Thennal and Hydraulic -With Densification Without Design Parameters Effects Jensification Effects Heat Transfer (Cont'd)

24. Heat Flux Hot Channel Factor, 2.32[cJ 2.40 FQ Fuel Central Temperature, °F
25. Peak at 100 Percent Power 3400 425U
26. Peak at Maximum Thennal Output for ~aximum Overpower Trip Point 4150 Core Mechanical Design Parameters Fuel Assemblies
27. Design KCC Canless RCC Canless
28. Number of Fuel Assemblies 193 193
29. uo 2 Rods per Assembly 264 ~04 30 . .Rod. Pitc11, in. 0 .496 0.563
31. Overall Dimension, in. 8.426 ,t.. 8.426 8.42G x 8.426
32. Fuel Weight (as UO?), pounds 222,739 215' 400
33. Zircaloy Weight, lbs. 50 '913 48,250
34. Numoer of Grids per Assembly 8- Type R 7-Type L
35. Loading Teclini que 3 region non-unifonn 3 region non-uniform_

Re vi si on 0 SGS-UFSA.R July 22, 1982

TABLE 4.1-1X (Sheet 4 of 6)

REACTOR DESIGN COMPARISON TABLE Saleffi URit 2 Saleffi URit 2 17xl7 Fuel Assembly 15xl5 Fuel Assembly With Densification Without Core Mechanical Design Parameters Effects Densification Effects Fuel Rods

36. Number 50,952 39,372
37. Outside Diameter, in. 0.374 0.422
38. Diametral Gap, in., Regions 1, 2, (and 3) 0.0065 0.0075 (0.0085)
39. Clad Thickness, in. 0.0225 0.0243
40. Clad Material Zircaloy-4 Zi real oy-4 Fuel Pellets -
41. Material . uo Sintered uo 2 Sintered 2
42. Density(% of Theoretical) 95 94, 93, 92
43. Diameter, in., Regions 1, 2, (and 3) 0.3225 0.3659 (0.3649)
44. Length, in. 0.530 0. 600 Rod Cluster Control Assemblies
45. Neutron Absorber .Ag-In*-Cd Ag-In-Cd 46.*Cladding Material Type 304 Type 304 SS-Cold Worked SS-Cold Worked
47. Clad Thickness, in. 0.0185 0.019
48. Number of Clusters 53 53
  • SGS-UFSAR 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 24 20 Cl uste*r Core Structure
50. Core Barrel, I.D./O.D., in. 148. 0/ 152. 5 148.0/152.5
51. The rma 1 Shi e.l d, I. D. /0. D. , in. 158.5/164.0 158. 5/164.0 Nuclear Design Parameters Structure Characteristics
52. Core Di_ameter, in. {Equivalent) 132. 7 132. 7
53. Core Average Active Fuel Height, i43.7 144 in.

Reflector Thickness and Composition

54. Top - Water pl us Stee 1, in .. -10 -10
55. Bot too - Water plus Steel, in. -10 -10
56. Side - Water plus Steel, in. -15 -15
57. H2o;u, Molecular Ratio, Lattice (cold) 2.41 2.52 RevisionO SGS-UFSAR July 22, 1982

TABLE 4.1-lx (6 of 6)

REACTOR DESIGN COMPARISON TABLE 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 Without Effects Oensification Effects Nu clear Design Parameters CeJ Feed Enrichment, wI o

58. Region 1 2.:i. sI 2.10 ~.25 z.eo/ 2. 60 2.80
59. Region 2 3.10 3.3l}
60. Region 3 '3.30/

[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. \

Revision 0 SGS-UF~R 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 uo melt temperature over the lifetime of the rod, including 2

allowances for uncertainties. The melt temperature of unirradiated U0 is 5080°F[l] and decreases by 58°~ per 10,000 MWD/MTU. From 2

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

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

and. anticipated 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 4.4-13 Re vision O SGS-UFSAR July 22, 1982

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 I_

TA:3LE ~ . .+-lA I 5,-:cct ~ Jf ~)

~EACTUR JESIGN CCMPARISJN ~~BL~ SA~~~ J~IT l 17 x 17 ..Jith Thenna l jnd Hydraulic Design Par-irneters :Jensification Re:lctJr Core Heat Output, MWt 333d 3338

- 11,393 ,( 1:, J9J ,( ~JG Reactor '.:ore Heat Output, Btu/hr

~e~ t Jt=nerated in Fue 1 ,  ? 7. 4 S;stem ?ressure, Nominal, psia 22:.0 System Pressure, Minimum Steady 2220 State, psi a

~inimu~ JNBR at Nominal Condittons T;pical ~low Channel 2.31 Thimble (Cold-..;all) Flow Channel 1.86 Mini~tirn JNBR f~r Desi~n Trdnsients > 1. 30 > l. 30 iJNa C0rre1ation " R- Gri d " ( ;.I - 3 "R-Grid" l..i-J with modified *~i th modi fi e'J spacer factor) Si:ldCcr factor-)

.:0:: 1 1n t Fl ow 1

132. 3 ..< 106 134 .1 x liJ 0

~ff2c:i~e ~low ~ate for

-rans fer, 1 o/hr 126.4 x 1Q6 128.0 ,( liJO

~ffecti 1e ~low Area Tr*a n s f e r , f t 2 51. l 51. 2 Fuel Rods, ft/sec 15.3 15.5 1 b/hr-f t2 2 . .+ 7 x 106 2.50 x 106 Coo: .: - - -

~~omi na ~ 544.4 544.4 inVessel,°F o..i.. 7 6 3. 'j in Core, °F 67.5 66.6 SGS-i.JFSAR Revision 0 July 22, 1982

TAaLE 4.4-lA (s~eet

)

  • REASTJR JES~GN CQMPAR!SO~

~I A

~7

I

. ~

,..i ::1

,, i : .

Ther::ia l *ind Hzdra J 1 i c 1 Desi~n Parameters J e 'l s i f i 1: 1 1: i c n Average in Core, OF Average in Vessel, 0 .-

r "J

i 'J

  • i

~ .

J .

.J

  • I

.J Heat - rans fer Active Heat Transfer, Surf ace Area, f t2 j2,2JJ 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 3.33

lax irnum Thermal ,J°utput, for rio~a1 o;Jeration, k
  • w/ft 12.i.+LDJ Peak l..i:-:ear ?ower for 1etennination

?rotection Setpoints, kw/ft is.oCcJ

?ea~ at 100 ?ower, 'F

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

>laximum Overpower Trip P int, °F Ac ro s 5 2~. 7 .. 2.sCd~

.;c.,..Js~:; psi 4-'J.:3 ... 3.J Jc.:. * ...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~ ::~.

Cb J This li, tis associated *"'ith the value o7 ~ = 2.32.

CcJ See Se ~ion 4.3.2.2.ci.

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

l.2 j Pre ously, a conservatively high value of pressure JroiJ NdS used to de ermine vessel loop flow rates.

Revision O SGS-UFSAR July 22, 1982

TAoLE 4.4-iX \sheet 1 of 2)

  • REACTOR DESIGN COMPARISON TABLE SAL.:. 11 17 x 17 ;.Jith

~.J :-;- c..

15 x ~ 5 ..Ji ti1ou t Thermal and Hydraulic Design Parafileters. Densification Densification Reactor Core i-ieat Output, MWt 3411 3411 Reactor Core Heat Output, Btu/hr 11,642 x 106 11,642 x 106 Heat Generated in Fue 1 , 97.4 9 7. 4 System Pressure, Nominal psia 2250 2250 System Pressure, ~~inimum Steady State, psi? 2220 2220 Minimum D~SR at ~ominal Conditions Typical Fl ow :hannel 2.24 2.0[aJ Thimble (Cold ..iall) Flow Channel 1.80 Minimum DN3R for Design Transients >1.30 > l. 30 DNB Cor"'el Jti on "R-Grid" (r-J-3 "?.-Grid" (.~-3 11i th modified with filOdi fi ed spacer factor) spacer factor)

Cooi ant Fl G'fl Total Therfilal F 1 ow Rate, lb/hr 132.2 x io6 134.0 x 106 Effecti v.c ;::-18'" 'ate for Heat T;a;1sFer*, 1 J,.*nr 126.3 x 106 128.J x 106 Effecti~e ~low Area for Heat Trarisfer, ft2 51. 1 51. 2 Avera*]e l~'ocit; Along Fuel Rods,  :,. s2c 15.4 15. 6 Avera~~ :1.J;~ .'elocitj, lb/hr-ft2 2.47 x 106 . 2. 50  :.<: 106 Nominal rn1et, °F 545.0 545.J Average i\ise  ;:1 '/essel, °F 65.8 65.1

'::- -~ '.l ;

SGS-UFSAR Revision O July 22, 1982

-,. . - ,. , *x

  • \

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

("'\,...) ..,.. ... ( s he et t. *Jf 2 !

17 :< 17 Aith l 5 x 15 ..ii thou t Ther:!lal a:;d ~ydr-3!.Jl ic Design ?:lrarneters . Jensification Je'lsification Avera::ie i 'l Core, °F 581.0 sao. 4 Aver::ige in '/esse 1, °F 577. 9 577. 5 Active Heat Transf~r, Surface Area, ft2 59,700 52,21.)l)

Average ~eat Flux, Btu/hr-ft2 189,700 217,200

.*1a,.< ii:iun -l21 t r-; 'j x ' for *10r~a1 00e."'at::;*1, 3:u/~r-ft2 440 , 200[0 J S80 ,OJO

,'.,ve:"l;e -:-~2~*:;ia1 .J:J:JJt, <*tJ/ft 5.H 7. '] 3

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

-- ~ J

, 2

  • b- 13.3
1 0 r~ J 1 0 '.) e r a : i 0 n <: "' / f t .L J2ci< ... i::~i*' Jower fJr :eten;1;r:ach., Qf

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

3400 4250

?ea~ d!; -,,er*'.,,: *~,_'::::..it *~aximum for

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

.~ c r::*:; s *= ') r~ ' "J 3 j 24.7 + 2.5[d] J2.s[e]

,.,._: -.-.

  • 0  : :e> *  : ncl udi ng nozzle, ps~ 49.'3 + *).0 J <:'.. J

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

  • G9 for a l irniting ':J:J'c3~ channel *,.,,.:is

~*~.:-.*.:*~ *~-:~./  :;.:~C-? ~fi-2 ':..'li:*:~Jle ;~oi*J wall) J:;d :ests*..;ere incomplete.

-), >': *-.-,;~ ':,  :;;soc;*~ted ..;i':.~ :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 u,

J y 2 2 ' l 98 2

TABLC: 4.4-ZA

  • THERi"1.~L-HYORAULlC lJES IGi~ PAK.Ai*lEiEi 1. 86 ana Anticipated Transients > 1. 30 SGS-UFSAR Revision 0 July 22, 1982

TABLE 4.4-3A VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS WITH DESIGN HOT CHANNEL FACTORS SALEM UNIT 1 Maximum Core Hot Subchannel 2.2

  • SGS-UFSAR Revision 0 July 22, 1982

TAdLC: 4.4-2X

\

THERMAL-HYDRAULIC OESIGN PARAHETtRS FOR ONE OF FOUR COOLANT LOOPS OUT OF SERVICE Total Core Heat Output, MWt SALEM UN If z 2388 Total Core Heat Output, 10 6 Btu/hr 8150 neat Generated in Fuel, 97 .4 ..

Nomi na 1 System Pressure, psi a 2250 Coolant Flo\-1 Effective Thermal ~low Rate for Heat Transfer, 10 6 1 bs/hr 90. 6 Effective ~l~w Area for Heat Transfer, ft 2 51. l Average Velocity along Fu~l Rods, ft/sec 10. 9 Average Mass Velocity, 10 6 2

1 b/hr-ft 1. 77 Coo 1 ant Temperature, °F Desi3n Nominal Inlet 539.1 Average Rise in Core 68.2 Ave rage in Co re 574.7 Ac:ive ~e3t fransferSurface 2

Ar::1. ft 59,700

~ver~;e ~edt Flux, Btu/hr-ft 2 132,900

~inimum ONB Rat~o dt Nominal Conditions > 1.80 1; ...,.: *~ *. _, I\:: ) ~-.;.,. - -., f ~~~*: ~..,

and Anticipated rransients > l.3U SGS-UFSAR Revision O July 22, 1982

TABLE 4.4 \

VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS WITH DESIGN HOT OiANNEL FACTORS mi" EM !ltll I 2 Average Maximum Core p.18%

Hot Subchannel 4.0% 13.6 Revision O SGS-UFSAR July 22, 1982

TAt3LE S.1-:.

SYSTEM DESIGN AND OPERATING PARAl~ETERS tt 11 i t 1 t:IFI ~ ~ "'}

Plant design life, years Number of heat transfer loops ,,..

~ 40 4

Design pressure, psig 2455 2435 Nominal operating pressure, psig 2239 2235 Total system volume including pressurizer and surge line (ambient conditions), ft 3 12,61~ *12,612 System liquid volume, including pressurizer and surge line (ambient conditions), ft 3 11, 8~2 11,892 Total heat output (100 percent power), Stu/hr u,q;;a 1' io 6 11, 68u x i06 Reactor vessel coolant temperature at foll power:

544.4 545.0 Inlet, nominal, °F 0

Ot1tl et, F' 609. 3 610 .2 Coolant temperature rise in vessel at fui 1 power, avg, °F eiJ.9 6:3.2 '

1l .Z

  • t X 10 G Total coolant flow rate, lb/hr
Z:::ld:t )(

11 6

"" ldd:9 x 10 Steam pressure at full ~ower, psia 805

  • SGS-UFSAR Revision 0 July 22, 1982

I -

I I

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

  • REACTOR VESSEL DESIGN DATA Design/Operating Pressure, psig Ul"li t 1 248S/22de U11i t 2 2485/2235 Hydrostatic Test Pressure, psig 3107 Design Temperature, °F 650 Overall Height of Vessel and Closure Heat, ft-in. (bottom head OD to top of control rod mechanism adapter H}. 43-10 Thickness of Insulation, ~in., in. + 3 Number of Reactor Closure Head Studs -&+- 54 Diameter of Reactor Closure Head Studs, in. +- 7 ID of Flange-, in. 172.S* 172.5

.OD of Flange, in. 205 173 ID at Shell, in.

Inlet Nozzle ID, in 27 1/2 27-1/ 2 Outlet Nozzle ID, in. ~ 29 Clad Thickness, min., in. 5/32 5/32 Lower Head Thickness, min., in. (base metal) 5 3/8 5-3/8 Vessel Belt-Line Thickness, min., in.

(base metal) -&: 8.5 Closure Heat Thickness, in. - 7 Reactor Coolant Inlet Temperature, °F 544.4 545.0 Reactor Coolant Outlet Temperature, °F 608.3 610 .2 132.2. ~ 10' Reactor Cool ant Fl ow, 1b/hr 134.1 )( 106 133 *. 9 }( 106-Total Water Volume Below Core, ft3 1050 Water Volume in Active Core Region, ft3 665

  • SGS-UFSAR 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 Unit 1 2164 U11i t 2 2164 Total Water Volume to Coolant Piping 2929 2959 Nozzles Centerline, ft3 Total Reactor Vessel Water Volume, (~lith -4~45 4945 core and internal sin place), ft3 Total Reactor Cool ant System Vo 1ume, ft3 12,612 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" I

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

I ls. 2q sq ~ s:~ G\.A u.....v.. r 1.'t. /

I

  • SGS-UFSAR Revision O July 22, 1982
  • .;;-... :r:. -~*.~-.;~-- *- --* .* ---
  • TABLE 5.2-5 (Sheet 1 of 2}

STEAM GENERATOR DESIGN DATA*

(Model 51) tin f t l <tfoitf Number of Steam Generators + 4 Design Pressure (Reactor coolant/ste~), psig 2485/1.005 2485/1085 Reactor Coolant Hydrostatic Test Pressure (tube side-cold), psi g 3107 3107 Design Temperature (reactor coolant/steam), °F 650/688 650/EiOO

'33.0S .ic 'O <.

Reactor Coolant Flow, lb/hr 33. 53 x 1:86 -li:47 )( 196..

Total Heat Transfer Surface Area, ft2 s1, sec 51,500 Heat Tran sf erred, Btu/hr 28§7 )( 1Q6 2920 x 106 Steam Conditi.ons at Full. Load, *Outlet Nozzle:

St~am ~.81 x l(J6 3.7 4 x 106 Flow, lb/hr Steam Temperature, °F ~ 519 -

Steam Pressure, psig ~ 805 Maximum Moisture Carryover, wt percent ~ 0 .25 Feedwater, °F ~ 435.

Overall Height, ft~n. 87 8 67-8 Shell OD (upper/lower), in. 175 3/4 I 135 175-3/4 I 135 Number of U-tubes ~ 3388 U-tube OD, in. 0.875 0.875 Tube.Wall Thickness (minimum), in. 9.0§9 o. 050 Number of Manways/ID in. .4;'16 4/16.

Number of handholes/ID, in. 4f+ 2/6

  • SGS-lJFSAR Revision 0 July 22, 1982

TABLE 5.2-5 (Sheet 2 of 2)

STEAM GENERATOR DESIGN DATA*

(Model 51) l:Jrd t 1 l:Jni t 2 Rated Load No Load Reactor Coolant- Water Volume, ft3 1080 1080 Primary Side Fluid Heat Content, Btu 28.7 x 106 27. 7 x 106 Secondary Si de Water Vo 1ume, ft3 1838 3524 Secondary Side Steam Volume, ft3 4030 2344 Secondary Si de Steam F1 ui d Heat Content, Btu S.738x 107 9.628 x 107 i

    • SGS-UF~R Revision o July 22, 1982

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

TABLE 5.2-7 REACTOR COOLANT PIPING DESIGN PARAr.fTERS Un i t 1 Reactor Inlet Piping ID, in. 27=1/f' 27-1/2 Reactor Inlet Piping Nominal Thickness, in. 2.58 2.38 Reactor Outlet Piping ID, in. ~ 29 Reactor Outlet Piping Nominal Thickness, in. ~ 2.50 Coolant Pump Suction Piping ID, in. t-- 31 Coolant Pump Suction Piping Nominal Thickness, in. 2.6& 2.66 Pressurizer Surge Line Piping ID, in. il. 500 (\)

Pressurizer Surge Line Piping nominal Thickness, in. -t:-2-5 (?..)_

Design/Operating Pressure, psig ~48~fh:::io 2485/2235 Hydrostatic -Test Pressure (Cold), psig 3107 Design Temperature, °F

  • 650 Design Temperature (pressurizer surge line}, °F 680
  • Water Volumej (all 4 loops including surge line} ft Design Pressure (pressurizer relief lines}, psig Design Temperature (pressurizer relief lines}, °F 1455 (3) t],}

~E Lt l:t::. "°"'~ ~o\e. * \-'.)~

"00,NC. ~au. ~l- ~~~ U,.E\J\ t\OtJ 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) ~ \,L\.~oo (1,) ~j_ . \. 1,'5 , ~'c-"L

\ bQ

~~

~)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 1y 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) ASr.E III, Class C (She 11 Si de) ASr.E VII I Residual Heat Removal Piping and Valves ANSI 831.1.0*

ANSI *s31. 7**

General 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 Refueling water storage temperature, °F Ambient 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 ( cn?t Nb
  • 11 72.1x10 6 -HJRit Ila. 2) .'\

H3so 3 concentration in refueling water storage tank, ppm boron -2000

  • Used for design.
    • 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:

No. 1 Unit No. 2 Unit

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

~ixiiin.11R G1:JaraRteee Rati "!t Gross Output, Mwe 1158 Anticipated Net Output, M#e 1115 Maximum L.a l cul ated Load Gross Output, Mwe 1176 1201 Anticipated Net Output, M#e 1130 1155 10.2.2 SYSTEM GESCRIPTION 10.2.2.1 Turbine~Generator 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 1e~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

Retrieved from "https://kanterella.com/w/index.php?title=ML18092A758&oldid=2506065"