ML20040E121

Response to Round One Question 440.40 on CESSAR Fsar. Nonproprietary Version

ML20040E121
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Site:	05000470
Issue date:	01/31/1982
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ML19297F308	List: LD-82-009, Forwards Proprietary & Nonproprietary Copies of Response to Round 1 Question 440.40 on CESSAR Fsar,To Be Used as Part of Cesec Review ML20040E121
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Download: ML20040E121 (75)

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'

l RESPONSE TO ROUND ONE QUESTION 440.40 ON THE CESSAR FSAR ECCS LICENSING ANALYSIS AND PLANT SYSTEMS ANALYSIS JANUARY, 1982 i s POWER i

SYSTEMS 8202030141 820129 COMBUSTION ENGINEEPING, INC PDR ADOCK 05000

LEGAL NOTTCE This report was prepared as an account of work sponsored by Combustion Engineering, Inc.

Neither Combustion Engineering nor any person acting on its behalf:

a.

flakes any war anty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the irformation contained in this report, or that the use of any information.,

apparatus, method, or process disclosed in this report may not infringe privately owned rights; or c.

Assumes any liabilities with respect to the use of, or damages resulting from the use of, any information, apparatus, method or process disclosed in this report.

RESPONSE TO ROUND ONE QUESTION 440.40 ON THE CESSAR FSAR ECCS LICENSING ANALYSIS AND PLANT SYSTEMS ANALYSIS JANUARY, 1982

1 I

ABSTRACT In response to NRC Round One Question 440.40 on the CESSAR FSAR, this report provides information on C-E's System 80 design to enable the NRC to model feedwater line breaks, steam line breaks, and large and small break LOCAs.

m, e

t CRITERIA FOR PROPRIETARY INFORMATION Information contained in this report which is delimited by means of surrounding brackets is proprietary to Combustion Engineering, Inc.

Code numbers 1-6 have been placed in the vicinity of such brackets to classify this proprietary information. The following list identifies the classification criteria associated with these code numbers.

Code Criteria 1

The information reveals privileged cost or price information, commercial strategies, production capabilities, or budget levels of Combustion Engineering. Inc., its customers or suppliers.

2 The information reveals data or material concerning Combustion Engineering or customer funded research or development plans or programs of substantial present or potential competitive advantage to Combustion Engineering, Inc.

3 The use of the information by a competitor would substantially decrease his expenditures, in time or resources, in designing, producing or marketing a similar product.

4 The information consists of test data or other similar data concerning a process, method or component, the application of which results in a substantial competitive advantage to Combustion Engineering, Inc.

5 The information reveals special aspects of a process, method, component or the like, the exclusive use of which results in a substantial competitive advantage to Combustion Engineering, Inc.

6 The information contains ideas for which patent protection is likely to be sought.

i

TABLE OF CONTENTS Page Criteria for Proprietary Information i

Table of Contents 11 Introduction iii - v Reactor Vessel 1 -1 to 1-43 Steam Generators 2-1 to 2-24 Reactor Coolant Pumps 3-1 to 3-15 Pressurizer 4-1 to 4-10 Emergency Core Cooling Systems 5-1 to 5-10 Primary Coolant Piping 6-1 to 6-21 System Valves 7-1 to 7-13 Fuel Rod Design 8-1 to 8-9 Containment Data 9-1 to 9-5

• At the end of each section the order of presentation of the information is as follows:

1) Qualifing Remarks and Comments

2) Schematics

3) Tables

4) Figures ii

Question 440.40:

As part of the CESEC review, the NRC intends to perform audit eval-uations of feedwater line breaks, steam line breaks, and large -

and small-break LOCAs (as part of the FSAR and TMI Action Plan Item II.K.3.31 and II.K.3.31 reviews).

In order to perform these audits we require the following data, as outlined in the "PWR Infonnation Request Package."

J

Response

This document provides the response to NRC Round One Question 440.40 for a typical System 80 plant.

The format of the response is similar to that of the Question. The '

responseis divided into nine sections. They are:

1.0 Reactor Vessel 2.0 Steam Generator 3.0 Reactor Coolant pumps 4.0 Pressurizer 5.0 Emergency Core Coolant System 6.0 Primary Coolant Piping 7.0 System Valves 8.0 Fuel Rod Design 9.0 Containment Data

• Next to where the information is requested in the original question, the reader is referred to either a table, figure, or schematic. These tables, figures, or schematics contain the information requested,and are located at the end of each section.

It should be noted that, for the most part, this is the data used in C-E's Safety Analysis. Many of these numbers include an appropriate amount of uncertainty. Therefore, these numbers may not be totally self-consistent.

In many situations,the amount of uncertainty included in the data has been identified. However, in some cases, uncertainties have been applied during the development of these numbers, and therefore, the exact amount of uncertainty included in thc3e numbers is not given.

• The information in this section, while tot requested in the package, was provided to model the pressure feedback effects of the containment during the Large Break LOCA transient.

iii

DCLDS UR E

( A c cl.se.d +o G M C 4 0 )

WR PFORy.: TION RE MEIT ::.e: !

.ntroduction The purpose of tnis package is to reouest s:e:-fic Pressurized Water Reactor (PWR) informa: ion, coth pnysical and operati:nal, :na: will allow tne plant to be modeled using advanced computer codes, suen as TRAC or RELAP. These plant models will be used in the study of various transients of concern to the nuclear industry.

For organizational purposes the plant modeling nas been subdivided into 7 main components:

1.

Reactor Vessel 2.

Steam Generator 3.

Reactor Coolant Pumps 4

Pressurizer 5.

Emergency Core Coolant Systecs 6.

Primary Coclant Piping

[

7.

System Valves O.

Fuel Rgd Design The forward and reverse flow energy loss coeffients, recuired in this package, are used to describe tne influence of coolaat, volume ge: metry upon coolant flow energy. For example, a 90' elbow will r.ot only enange the directicn cf cc: lent flew but will cause the coolart to 1:se e.9ergy as well. Tne coefficient is dimensionless and is a function of the frictic,n factor and the ecuivalent L/0 of the piping features at the junction or in

ne volume. The basic ecuation is:

K = f L/0 witn f

friction factor

=

L, pipe length

=

0 pipe diameter

=

l iv i

b Features with equivalent.L/Ds to be considered. i :t ce 'but are not restricted tal abrupt area changes, p'enum volumes, m:'sture se:*r: tors, The terms ' forward' and ' reverse' a:cly to the normal valves and elbews.

and reversed direct on of ficw, re pect vely.

i i

Operational information such as controls, operating conditions and There are alarm setpoints are critical to the correct modeling of the PWR, many factors that affect this information. Of prime cor.cern, are the delays in system actuation caused by instrument deadbands, uncertainties and attached electronics. These are not specifically written in this Mcwever, this package due to the plant specific nature of these factors.

information is needed and, is therefore, requested.

Due to the generalized nature of this questionaire, some of the In information requestad may not be applicable for a particular plant.

these cases the requested parameter should be ignored.

In addition to the requested data in this package, schematic drawings depicting each of the major compcnents should be included.

l l

v

~

SECTION 1.0 REACT 0k VESSEL VALUE l

1.

n12: nozzles 1

~

f.

Insics diameter at no::le inlet Schematic 1.1 3.

Inside dit.eter at no:zle outic; Schematic 1.1

,C.

Distance frcm nozzle inle: to Schematic 1.1 nozzle cutlet D.'

Forward flow energy loss coefficient Table 1.1

'E.

Reverse flow energy loss c:sfficient Table 1.1 F.

Inside surface roughness

, Ylf * 'k #')

II. Downeccer A.

Flow area as a functicn of elevatien, relative to inlet no::1c centerline Table 1.1 S.

Full power inlet te=perature Table 1.1 C'.

Full pcwer inlet pressure Table 1.1 D.

Elcyctica of top of dcGnce=er relative in inlet no: le centerline Schematic 1.1 E.

Forward ficw energy loss coefficient Table 1.1 F.

Reverse flow energy loss coefficient Table'1.1 G.

Surface rougnness N/A H.

Hydraulic diameter as a function of

~

clevatica, relative te inlet noz:lc centcrline Table 1.1 III. Lower Plenum (below ficw dist.-ibutor)

A.

Ficr.s area as a functien of elevatien,'

relative to inlet no::1c cen: rline Table 1.1 S.

Totc1 volume including structural f

material

• l Table 1.6 & 1.7 C.

lietal-te-water volume ratio 8

hmark #15 1 -1

.t.

VALUE D.

For sr: flow energy '::s ::cffi:ier:

Table 1.1

~

E.

Revs-se flow energy

s coeffi:ie..:

Table 1.1 J

F.

Avert;c rougnness N/A G.

Frac:icnal co=:ositien of structural com::enents (e.g., 53-3C5 25.4%, e::)

Table 1.11

~

H,.

Hydraulic diameter as a function of '

elevation relative to inlet nozzle centerline Table -1.1 IV. Lower plenum flow distributor A.

Flew area Table 1.1 B.

Forward flow energy loss coefficient Table 1.1 Table 1.1 C.

Reverse flow energy.Tess coefficient D.

Cc: position Table 1.11 E.

Axial elevation at center and at edge Schematic 1.1 Schematic 1.1 F.

Thickness Y.

' Lower plenum between distributor and lower core plate A.

Flow area Table 1.1 Table 1.1 B.

Hydraulic dia;::eter C.

Forward flow energy less coefficient Table 1.1 D.

Reverse flow ener7y. less coefficient Table 1.1 E.

Roughness '

N/A F.

Material'cc= position Table 1.11 6.

Total volume' including structural material Table 1.6 & 1.7 H.. Metal-to-water volu=e ratio See Remark #15 t

0 0

,e 3

1-2 1

If!.

Reac::r ::re VALUE A.

Fuel asse:oly Table 1.5

' 1., Flow area Table 1.5 2.

Hydraulic diameter i

l 3.

Forward flow energy loss coefficient at grid spacer '

Table 1.1 4.

Reverse ficw energy loss coefficient at grid spacer Table 1.1 N/A 5.

Roughness 6.

Material co=;:esition Table 1.n 7.

Total volume Table 1.5 8.

Metal-to-water volu=e ratio see Remark #15 9.

Axial elevations of center of grid spacers, relative to inlet nozzle centerline Table 1.2 8.

Control rod asse. coly 1.

Flow area Table 1.1

~

2.

Hydraulic diameter Table 1.7 (CEASECHON)

3. "' Forward ficw energy lo:s coefficient Table 1.1 at grid spacer 4.

Reverse flew energy loss coefficient, l

4t grid spacer Table 1.1 5.

Rougnness N/A l

6.

Material cc=;osition Table 1.11 7.

Total volu=e Table 1.7 8.

Metal-to-water volume ratic Remark #15 9.

Axial elevations of center of grid scacers, relative to inlet noz:le centerline Table 1.2 l

l

~

.e

.. _.,.. ~

1-3

.e C.

Fuel assemoly wita instrumor.:

VALUE 1.

" Flow area SEE 2.

Hydrault: diameter FUEL ASSEMBLY 3.

Forward flow energy loss coefficien:

(SE Q 0N A) 4.'

Reverse flow energy loss coefficien:

SEE QUALIFYING 5.

Roughness 6.

Katerial co= position 7.

Total volume 8.

Metal-to-water volu=e r,atio

^

D.

Core bypass flow path (s) 1.

Flow area Table 1.3 2.

Hydraulic diameter Table 1.3 3.

Forward flew energy loss coefficient Table 1.3 '

4.

Reverse flow energy loss coefficient Table 1.3 5.

Roughness N/A 6.

Total volume Table 1.3 7.

Percentage bypass of core full power flow Table 1.3 E.

Core pcwer distributions (axial and radial for eacn of the following conditions)

Tables 1.8, 1.13, 1.16 1.

Normal full power

& Figure 1.8 2.

Control rods 5C% inserted Not used in C-E's analysis,

3.

Control rods fully insertad wita cost significant control rod asse=bly stuct cut Table 1.16

~

F.

Reactor protective system interac:fons wita core Table 1.12 S.

Engineered safeguards protective system intar-Table 1.12 actions with core Remark #18 H.

Coolant temperature at core inlet as 4 function of core power Figure 4.1 8

l-4

--e,--n-

l

.o VALUE Oce* ant pressure at core inla; as 1 Remark #18 fun: tion of core ::wer Table 1.1 J.

Coolan: temperatu.e a: core ::o as a '- -

Remark #18 function of core pewer Figure 4.1

l.,,

K.

Coolant pressure at core tcp as a Table 1.1 function of ccre power Remark #18 L.

Quantity of gama heating in core as a function of core pcwer and axial elevatfor.

Table 1.16 M.

Reactor kinetics--beginning of life Remark #17 Table 1.18 1.

Scram rod reactivity insertion as a function of time for:

Tables 1.9 & 1.17

=

a.

all rods drop Remark #12 & 16 b.

all but most reactive red drops 2.

Reactivity enange as a function of Table 1.14 &

moderator density Figuro 1.5 3.

Density reactivity change as a function of boron concentration N/A 4.

Reactivity enange as a function of Table 1.15 &

fuel temperature (Doppler)

Figure 1.6 5.

Boren worth as a function of bor,on concentration and moderator temperature Remarks #13 & 14 N.

Reactor kinetics - end of life Table 1.18

~

Remark #17 1.

Scram rod reactivity insertien as a function of time for:

a.

all rods drop Remarks #12 & 16

.. Table 1.1($ 1.17

.b.

all but most reactive rod drops 2.

Reactivity enange as a functicn of TWe 1.14 &

moderator density Figure 1.5 3.

Density reactivity enange as a functi:n of boron concantration N/A 4.

Reactivity enange as a function of Table 1.15 &

fuel temperature (Deppler)

Figure 1.6

,,. ~

s 1-5

1 l

VALUC 5.

Eer:n w:rtn a: a func-ion of ::r:n

.concentra:icn ano mocera::r e.:grt:ura Remarks #13 & 14 VII. Upser c:re plenum frem tco of outlet no:zles to tot.cs of vessel head Table 1.4 A.

F c,< area Table 1.4 8.'

Hydraulic diameter Table 1.4 C.

Fhrwardflevenergylesscoefficient D.

Rpverse flew energy less c: efficient Table 1.4 E.

Rcugnness N/A Table 1.11 F.

Material cc positi:n Table 1.7

. tal voic=e c.

Remark #11 H.

Metal-to- <ater vole =e ratio Remark #15 I.

2nside diame:cr of plenum shroud ' orifices N/A VIII.

U;;::r nead A.

Flew area Table 1.4 Table 1.4 B.

Hydraulic diameter--

Table 1.4 C.

Forward ficw energy loss coefficient,

D.

Reverse flow energy loss c: efficient Table 1.4 N/A E.

Rcugnness Table 1.11 F.

Material c:=;:esition Table 1.'6 G.

.stal volume Remark fil-l' H.

Metal-to-water volume ratio Remark #15 4

l 1-6

II. Outlet nozzles VALUE A.

Inside diameter at no::le inlet Schematic 1.1 8.

Inside diameter at nc::le outlet Schematic 1.1 C.

Distance frca no::Te 1nlet to nc::1e outlet Schematic 1.1 D.

Forcrd ficw enere_y loss coefficient Table 1.1

7..

'. cyc rs e fic.< :..r : e v icss c:efficier.:

Table 1.1 F.

Rouchness N/A 3

Elevation,vetativ'e to inlet no::le Schematic 1.1 See Table 1.7 for Outlet Nozzle Volume l

l l

1-7 i

REACTOR VESSEL The following remarks qualify the data presented:

Table 1.,1 provides flow amas, adjusted. pressure drops at stated 1.

reactor conditions, and fluid properties. Note that pmssure losses are provided in lieu of flow coefficients. This is as agreed to by #4L representatives that are perfoming the analyses for the NRC.

2.

Table 1.2 provides the position of each spacer grid and is given relative to the bottom of the active core.

3.

Table 1.3 provides the flow areas (average), channel flow length, and adjusted flow rates as well as controlling pressure loss areas for each bypass path.

The bypass paths are illustrated in Figure 1.2 and described below.

a.

Through the clearance gap between the outer face of the core support barrel extension and in the inner face of the outlet nozzle.

b.

From the vessel entrance region, around the alignment keys into the dome region.

c.

Through flow holes in the core support cylinder, up into the core shroud - CSB annulus and into the core region imediately below the fuel alignment k. ate.

l d.

The corner (outer) guide tubes have flow entering the flow holes l

located above the LEF plate. The rodded corner tubes dump their l

flow into the UGS region via the tie tubes. The empty corner tubes dump their flow into the core region just below the FAP.

e.

The center guide tubes dump their flow into the core region be. low the FAP. The empty center tubes have flow entering tt: rough the hollow center leg of the LEF. The instrumented center tubes have their flow entering the instrumentation nozzle' located in the lower plenum.

4.

The core shroud has no seams for leakage flow.

5.

Table 1.4 provides adjusted famard and reverse pressure losses in the outlet plenum - dome flow network shce in Figure 1.3.

These losses are given together with the minimum flow area for each flow path and the percentage of the pressure loss due to friction. The reverse flow pressure losses are based on the same flow rate and fluid pro-perties as the fomard flow losses.

6.

Table 1.1 refers to Figure 1.1.

7.

Table 1.3 refers to Figure 1.2.

l 8.

Table 1.4 refers to Figure 1.3 9.

Table 1.5 provides selected gaometrical and themal infomation which may be required forinput data.

1-8

REACTOR VESSEL j

'10.

The instruments are located in the center guide tube. They have no themal hydraulic effect on the fuel assembly. Flow is accounted for in the bypass flow paths.

' lume is depicted as Top Head Volume in Table 1.7

11. Upper Hi i

and Figure i.4.

12. Table: 1.9 and 1.19 show the fractional reactivity inserted during a scram for the range of allowable ASI's. The curves am nomalized to a total inserted reactivity of 1.0.

The actual reactivity inser:ed during a scram at HFP (Hot Full Power) can be obtained by multiplying the mactivity values on these curves by the minimum net rod worth at HFP (with the most significant; rod stuck out) which is 10% t.o.

13. Reactor Coolant System Boric Acid Concentrations The nominal value of the hot full power critical soluble boron concentration is shown in Figure 1.7.

The maximum soluble baron concentrations at hot full power, hot zero power and cold shutdown are: 1002 PPM,1120 PPM and 2365 PPM, respectively. These values, which include uncertainties, are valid for the conditions of ARO.

(All Rods Out). The minimum soluble boron concentrations at hot full power, hot zero power and cold shutdown are: O PPM, 94 PPM and i

451 PPM, respectively. Once again, these values are only for the conditions of A30, and include uncertainties.

14. Critical Baron worth as a function of moderator temperature is given in Table 1.16. Critical boron concentration is given as a function of burnup (MWD /T) in Figure 1.7.

15. Metal to water Volume ratios am not used in C-E's analysis and are themfore not generated. However,these ratios could be fomulated, for the most part, from the infomation which is supplied by way of j

water volumes and metal wall heat transfer infomation.

l

16. Scram rod reactivity is applicable only to Small Break LOCA*

l analyses. Large tinak LOCA* analyses shutdown on voiding.

17. The physics information presented in this section is chosen from more than one fuel management scheme and is used in C-E's safety analysis.

(Some of the infomation is taken from the Arizona extended cycle, while other information is relevant to the CESSAR 12 month cycle). The data was chosen to conservatively envelope these fuel management schemes.

Therefore, it is not reasonable to assume that this particular set of parameters will be self consistent. Because of this, a detailed core analysis employing this data may not yield meaningful results.

18.

In LOCA analyses

• temperatures and pressures are input for full power conditions only.

• CESSAR-F Analysis perfomed by C-E.

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TABLE 1.l System 80 Forward & Reverse Flow Pressure losses (2)

Forward Flow Reverse Flow Station F}ow Areas, ft Station Appsi a p psi A1

= 19.635 1-2

.90

.96 A2

= 27.16 90*

7.41 8.23 A3

= 30.09 2-3

.07

.36 A4,5

= 36.96 3-4

.52

.56 A6,7

= 40.53 4-5

.65 0

5-6

.03

.16 A8

= 33.75 6-7

.55

.55 Ask

= 37.97 7-8

.40

.13 A9

=115.62 8-9 10.30 7.28 A10

= 31. 91 9-11 3.67 8.56 All

= 90.21 11-15 1.95 1.98 A15

=100. 49 15-20 Fuel Assembly

• 17.50 19.09 A16

= 44.39 A17,18

= 60.80 (Total) 20-22 9.78 14.74 AGAP

=103.48 22-23 7.31 11.17 A19

= 46.31 23-24 1.30 3.56 A20

= 87.39 A21

= 20.61 Totals 62.34 77.33 A22

=105.66 A23

= 42.89 Conditions A24

= 19.24 565.0F Core Power 3876 MWT (102%)

Teore in

=

Operating Pressure 2250 psi

( AT PRESSURIZER)

Teore Avg = 594.5F Yessel Flow Rate 163.9x10 lbm/hr (minimum)

Teore out = 624.0F Core Flow Rate 159.0x10 lbm/hr Fluid Procerties Stations Specifie Volume Viscosity 3

(ft /lbm)

(1bm/hr-ft) 1-17

.021804

.22513 17-18

.022918

.21232 18-24

.024467

.19708

~

• Percentage o* Total Pressure Drop Across Fuel Assembly Station I.ocation Forward (%)

Reverse (%)

~

15-17 Lower End Fitting (LEF) h 17-18 Friction 3

Grids 18-20 UpperEndFitting(UEF) 1-11

Table 1.1 (Continued)

Notes:

(1) This table refers to Figure 1.1.

(2) All pressure losses have been increased by +10% above best estimate values. -(Conservative Direction).

I e

1-12

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TABLE 1.2 Location of Spacer Grids, System 80 Spacer Grid Distance from Centerline of Grid Tabs to Active Core Bottom, Inches 2.52 1

l 2

16.83 l

3 32.55 4

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REACTOR VESSEL ASSEleLY

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ASSEMBLY

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. UPPER GUIDE

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. STRUCTURE.E

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N0ZZLES i

l 1-39 s

FIGURE 1.5 CORE REACTIVITY (%ae) vs MODERATOR DENSITY USED IN LOCA ANALYSIS

+0.4 g

D n

s

+0.2 0.0 O

4

-0.2 C

3' E

C

-0.4 For ARD conditions these curves e

correspond to:

W

-O-MTC=0.0 x 10-4

- (use for smaii areak)

-0.6 ae

-O-MTC=+.5 x 10-4ae: (use for targe Break)

-0.8

• MTc - Moderator Temperature coefficient

.i l

l I

50 45 40 35 30 3

CORE AVERAGE f10DERATOR DENSITY (LBM/FT )

FIGURE 1.6 CORE REACTIVITY vs CORE AVERAGE FUEL TEf1PERATURE USED IN LOCA ANALYSIS 1.8 1.4 1.0

-O-BEGINNING 0F LIFE

-O-END OF LIFE 0.6

^w 0.2 3

T' -

SE t;

22 24 26 28 30 32 34 36 38 4'O 42

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I FIGURE l7 l

CRITICAL BORON CONCENTRATION vs BURNUP (MUD /T) 3 4

D 1-42 I.

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FN E 1.8 i

i LA3GE BREAK LOCA AXI AL POWER SH APE 5

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SECTION 2.0 STEAM GENERATORS All elevati:ns sn:cle :e relative to inle: cf :ne steam geners::r rea:::r coolan: inle: nonie, wnese elevati:n relative to :ne centerline of tne reac::r vessel cold leg inle: snoule se included.

If tne steam genera::rs differ frem ea:n otner tne re:uested information snould De proviced for ea:n generator.

All elevations are relative to reactor vessel cold leg inlet no::le centerline.

I.

Primary Leop VALUE A.

Primary coolant inlet nonle

'1.

elevation of inlet no::le centerline at the entrance to the inlet plenum Schematic 2.1 2.

inside dia=eter at nc=le inlet Schematic 2.1 3.

inside diameter at plenum inlet Schematic 2.1 4.

length of nonle at n:=le centerline Schematic 2.1 5.

angular orientatien of nonle centarline relative to nori: ental Schematic 2.1 6.

forward flew energy loss coefficient Table 6.2 7.

reverse flow energy loss coefficant Table 6.2 8.

inside surface r ugnness N/A 9.

flowrate at nonle entrance at full load Table 6.2 10.

coolant temperature at nonle entrance at full load Table _6.2

, 11.

coolant pressure at nc=le entrance l

at full load Table 6.2 l

S.

Primary c:clant inlet plenum 1.

elevatten of tute snest bundle entrance Schematic 2.1 Table 2.1 2.

plen'um volume 24

3.

arel of ;*e um at ?ntr:nce to tuce our.:le Table 2.1 4

forward fic.< energy loss ::sfficien:

Table 6.2 5.

rever.se flow energy loss esefficient Table 6.2 6.

rougnness M/A 7.

wall tnickness *

. Table 2.1 8.

width of plenum divider (UTSG)"

at tube sheet entrance Table 2.1 C.

Tube bundle

.1.

number of flow tubes Table 2.1 2.

tube ID Table 2.1 3.

tube 00 Table 2.1 4.

elevation of tube bundie exit once througn steam generator (OT55)b Does not apply 5.

lengtn of flow tubes (OT5G)

Does not apply average length u-tube steam generator (UT3G)

Tab 1e 2.1 6.

Heat transfer area Tab 1e 2.1 7.

Heat transfer area including curved section (UTSG only)

ITEM 6 8.

forward flow energy loss coefficient Table 6.2 9.

reverse flow energy loss coefficient

• Table 6.2 10.

intem al reughness N/A 11.

total volume in tubes Table 2.1

12. maximu:.-Jminic:m and average tube elevation (UT5G) schematic 2.1 a.

U-tube steam generator D.

Once througn steam generator

~

~

,.. ~

2-2

.~

G l.

3, Sri =ary Coolar.: Outle: ?lenum VALUE l

1.

elevaticn of tud: sneet ext:

schematic 2.1 2.

plenum volume Table 2.1 3.

area of plenum at tune exit Table 2.1 4.

forward flow energy loss coefficien Table 6.2 5.

reverse flow energy loss coefficient Table 6.2

~

6.

rougnness N/A

~

7.

wall tnickness Table 2.1 9

i E.

Primary Coolant Outlet No::le 1.

elevatien of in1et no::le centerline at the entrance to the inlet plenum schematic 2.1 2,

inside diameter at no::le putlet schematic 2.1 3.

inside diameter at no::le inlet schematic 2.1 4.

lengtn of no::le at no::le centerline schematic 2.1 5.

angular orientation of no::le centerline relative tihorizontal schematic 2.1 6.

forward ficw energy loss coefficnent Table 6.2

~

7.

reverse flow energy loss coefficnent Table 6.2 i

8.

inside surface rougnness

.N/A 9.

flowrate at no::le entrance at full load Table 6.2 los coolant tec;erature at em::le entrance at full load Table 6.2

11. coolant pressure at nc::le entrance at full load Table 6.2 1

(

8 e.

g l

2-3 l

n.

se::acacy.:::

val.UC A.

Festas;er Sc;p1y 1.

feed flow at full leze Table 2.3

~

2.

feed f1cw as a fun::ica cf Comment 2.1 a.

load b.

mixture level 3.

feedwater temperature Ta ble 2.3

'4.

feedwater pressure Comment 2.9 5.

auxiliary feed syste a 4.

initiating set;0ints Ta ble 2.3 b.

flow rate ea:n type as a function Table 2.3 of pu=ps running C.

aux feed tC perature Table 2.3 (each type) d.

nu=cer of pu=;s (eacn type)

Table 2.3 e.

aux feed pressure Ta bl e 2.3 (eachtype) f.

aux feed inlet elevation Schematic 2.1 6.

main feed inlet elevation Schematic 2.1 B..

Downc::er(Prencater)Section l'.

downcocer(OT5G)

THIS SECTION DOES NOT APPLY TO A UTSG a.

flow area as fun:tien of.

elevation above ::p of 1ower tube plate b.

roughness c.

fon<ard ficw energy loss ccefficient d..

reverse ficw energy less c: efficient e.

dcwnce er shr: d ID f.

downce=er shr: d 00

.2-4

_....n.

g.

taffle snroud I; n.

taffic snroud C3 1.

snell ID j.

,snell CD t.

elevation of baffle snreud bottc= acove lower tube plate 1.

elevation of downce=er shroud tcp above lower

~

tube plate m.

total volume in sectien n.

heat transfer area.

2.

preheater (UTSG) a.

elevation of (1) top cf preheater section Schematic 2.1 (2) bottem of preheater S' hematic 2.1 c

section b.

flow area as function of Figure 2.3 height of section c.'

forward flow energy loss coefficient Figure 2.3 d.

reverse flow energy loss coefficient Not Available 4.

roughness external tubes /

Table 2.4 baffles f.

heat transfer surf ace Comment 2.2 9

Coolant volume of section Table 2.4 h.

metal volume in secticn Not Available 3.

operating conditiens Comment 2.2 a.

outict temperaturc b.

outlet pressure l

c.

outlet flow if different frem inlet flew 4.

outlet to tube bundle (toiler) a.

flow area Figure 2.3 i

b.

• forward ficw energy less c cfficiant Figure 2.3 c.

reverse flew energy less c: efficient Not Available 2-5

_,.......,,,m,..

VALUE Tuse' Bundle (3siler) 1.

fic area tne:ugn tubes as a Figure 2.3 function of naf gn:

2.

total nea: t.ansfer area Table 2.4 3.

forward ficw energy loss coeff fcian:

Figure 2.3 4

reverse flow energy loss coefficient Not Available

~

5.

.r:ugnness (external tubes)

/

Table 2.4 6.

elevatiens Comment 2.3 a.

tcp of baffle assa.t.bly b.

bottom of upper tube plate c.

top of baffle shroud 7.

heignt of top of nucleate boiling Comment 2.4 region as a function of load (above lower tuce plate) 4.

heignt of tcp of film boiling Comment 2.4 region as a function of' Toad" (acove Icwer tube plate) 7-t O

e l

9 l

2-6

9.

f1:- losses in :tifie regicn Corment 2.5 i

• 10. metal volume in.s; ten Not Available

~

11. coerating c:nci:icns Comment 2.6 a.

, outlet tem:erature k.

outlet pressure

'c.

cuality as a function of

~

heignt in regicn

12. total free volume in region Table 2.4 D.

Superneat Steam Downc mer (OTSG)a THIS SECTION DOES NOT APPLY TO A UTSG 1.

steam senerater ID 2.

rougnness 3.

forward flew energy loss coefficien:

4.

reverse flow en'ergy loss coef.ficient 5.

elevations:

a.

bottom cf dcwnc:=er b.

steam outlet centerline 6.

flow area ts. a function of height' 7.

steam outlet no::le ID 8.

steam ccnditions at exit

a., outlet temperature b.

outict pressure c.

flow rate as a function of load' 9.

heat transfer area 10.

total free vole =e in regien E.

Steem De=e (UTSG)D e.

Once tarcugn steam generator G.

U-tube stans generater e

2-7

l 1.

f1:- area as a fun:tien :f Figure 2.3 heignt (t:o of tuses :s 3,e:rl vene moisture sisaratsr(s) (SVP.5) 2.

eleva icns schematic 2.1

,a.

too of tuce our.dle D.

bott:m of SVPS C.

top of SYMS d.

bott m of steam dryers e.

top of steam dryers f.

steam outlet 3.

SVP.S (Steam Separators - CE)*

a.

forward flow energy loss coefficient Figure 2.3 b.

reverse flow energy loss coefficient Not Available C.

roughness Not Available d.

flow area througn SYM.S Figure 2.3 e.

recirc. flow as a Comment 2.7 function of load f.

number of swirl vanes Not Applicable 9

nu.i.bcr of stcam separaturs (CE)

Table 2.4 4.

ste'an dryers a.'

forward flow energy loss cocfficient comment 2.8 b.

reverse flow cncrgy loss coc'fficient Not Available C.

roughness Comment 2.8 d.

flow area through dryers comment 2.8 e.

racire. flow as a comment 2.7 l

function of load 5.

total free volume in region Ta bl e 2.4 6.

total metal volume.

Not Available 7.

operating conditiens a.

'stcam outlet pressure Table 2.4 b.

steam outict tc perature Ta bl e 2.4' c.

steam flow a: a function comment 2.6 of load

1.,Comoustien Er.gineering

\\

f e

2-8

D.

Steam ou!Ict n ::le.3 Figure 2.5 F.

SG Material C:co siti:n Table 2.5 1.

SG vessel

~

2.

baffles Not Applicable 3.

downcocer shr:cd (0753)

Not Applicable As tube suppcrt p1'ates

, 5.

primary tubes 6.

SYMS (steam separaters) 7.

stcam dryers G.

Yalves THE INFOR?tATION AVAILABLE ON SECONDARY SYSTEM VALVES WILL BE PRESENTED IN SECTION 7. SYSTEM VALVES.

1.

c:ain steam isolation valve a.

valve diameter b.

control setpcints distance frem SG cutict ne:ble c.

I.

vertical II.

horizontal 2.

relief valves e

a.

valve dia:cters k.

contr:1 sctpoints distance frcm SG outict nd::le,

c.

~

I.

vertical II.

hori: ental 3.

atmospheric de:p valves a.

valve diameters b.

control setpcints c.

distance fr:m 53 outlet ne::le 1.

vertical l.,

II.

nori:catal 0

y.. -

e 2-9

Comments on Section 2. Steam Generators I

Comment 2.0 The seconda'ry side shock loss coefficients and pressure drops presented in Table 2.4 and Figure 2.3 are best estimate, i.e., there are no uncertainties on them. This should be taken into consideration when using these numbers.

It is recommended that a 20% uncertainty be added to these numbers in the conserva-tive direction.

Comment 2.1 The feed flow is controlled by the feedwater control system. See Figure 2.2 for a diagram of the feedwater control system and its logic. Feedwater flow as a function of load is dependent on feedwater temperature. The downcomer level is normally controlled to a constant level of 37.3 feet above the top of the tube sheet for all power levels.

Comment 2.2 C-E's steam generators utilize an integral economizer. Because of this, the data on the heat transfer surface area of the preheater is included in the section on tne tube bundle (boiler). Also, data on the operating conditions (temperature, pressure and flow) will not be providad.

Figure 2.3 indicates the various flow areas and shock losses 1.1 the secondary side of the steam generator.

s I

Comment 2.3 l

The terminology used to define these specific locations in the steam senerator is not clear. For the elevations of any point in the steam generator see Schematic 2.1.

Comment 2.4 These two questions are not applicable to the C-E design of steam generator. The height of the top of nucleate boiling is never reached, as throughout the tube region only nucleate

(

boiling is occurring. There is never any film boiling occurring in the tube region during normal operation.

2-10

Comment 2.5 The 'nformatiori that is requested in this section is not clear.

For shock losses in the secondary side of the steam generator see Figuie 2.3 Comment 2.6 Operating conditions at various points inside the steam generator are not available. Steam outlet conditions and the steam flow at 100% power are provided in Table 2.4.

Values for the tube bundle exit flow quality can be extracted from the information on circulation ratio versus load presented in figure 2.4.

The full power steam finw is ennsistent with a feedwater temoerature of 450'F. Steam flow at other loads is dependent on the temperature to which the feedwater is controlled.

Coninent 2.7 The recirculation ratio as a function of load is not available separately for the steam dryers and the steam separators. Figure 2.4 provices a graph of circulation ratio versus percent load in terms of total flow through the tube bundle (evaporator) divided by the steam flow leaving the tube bundle.

Convent 2.8 The pressure drop across the steam dryers and the steam outlet nozzles were determined from experimental data. These pressure drops will be reported in Table 2.4 Comment 2.9 i

The feedwater pressure will vary with steau. generator pressure.

Table 2.4 provides the total pressure drop across the secondary side of the steam generator under full power conditions.

i l

l 2 11 i

(

y r

l l

l W

M C &

=

N < Q C C3 W

U.

-== e m a &

C4 Wv y

":C R C4 u w C4 &Co I

2-12

TABLE 2.1 Steam Generator Primary Side Infomation 3

Primary Coolant Inlet Plenum Volume 337.11 ft 3

Area of olenum at entrance to tube bundle Wall thickness (inlet plenum)

Width of plenum divider at tube sheet entrance Nu.iber of flow tubes Tube ID 0.0555 ft Tube OD 0.0625 ft 3

Average length U-tube Heat transfer area (tube bundle)

Total volume in tubes (active)

Total Volume in tubes (inactive)

Elevation of tube sheet exit 3

Primary coolant outlet 91enum volume 337.11 ft 3

Area of plenum at tube exit Wall thickness (outlet plenum)

Installed number of tubet with no tubes plugged

• • Based on t):e number of installed tubes 2-13

i O

TABLE 2 2 INTENTIONALLY LEFT BLANK 2-14

TABLE 2.3 FEEDWATER SUPPLY 6

1.

Feed Flow at Full Load 8.59 x 10 lb/hr 3.

Feedwater Temperature at full load. 450 F 5.

Auxiliary Feed Systems The specific design of the auxiliary feedwater system is beyond the scope of Combustion Engineering's System 80 design. In place of the specific data requested the following interface requirements will be presented:

1.

The emergency feedwater system shall maintain adequate inventory in the steam generator (s) for residual heat removal and be capable of the following:

a.

Maintaining the NSSS at hot standby with or without normal offsite and normal onsite power available.

b.

Facilitating NSSS cooldown at the maximum administratively controlled rate of 750F/hr. from hot standby to shutdown cooling initiation with or without normal offsite or onsite power available.

(The Shutdown Cooling System Mcomes available for plant cooldown when the RCS temper-ature and pressure are reduced to approximately 350 F and 400 psia.)

2.

The Emergency Feedwater System shall be available to deliver flow to the steam generator (s) automatically upon receipt of an EFAS as follows:

Withir,19 seconds when normal offsite or normal power a.

is available.

b.

Withir. 45 seconds when both normal onsite and normal offsite power are not available.

(

3.

The required emergency feedwater flow, based on residual heat removal requirements is 875 gpm delivered to the steam generator (s) downcomer feedwater nozzle. Maximum expected steady state steam generator pressure at the downcome, nozzle is approximately 1275 psia.

4.

Emergency feedwater temperature shall be at least 40F and no greater than 180F.

5.

A minimum of 300,000 gallons of secondary quality makeup water shall be available to the Emergency Feedwater System for delivery

(

to the intact steam generator (s). This amount ensures sufficient l

feedwater to allow an orderly plant cooldown to shutdown cooling initiation conditions.

• See table 1.12 for analysis setpoints.

l 2-15

TABLE 2.4 ADDITIONAL SECONDARY SIDE STEAM GENERATOR DATA Preneater and Tube Bundle (Boiler)

_3 Coolant volume of economizer Region Total heat transfer area I Roughness external tubes Total Free Volume in evaporator Region Total Free Volume in Steam 05me Region 3 Total Volume in downcomer Region 3 Stea5LDome Numter of Steam Separators Steam outlet pressure 1070 psia Steam outlet temperatura 552.9 F 6

Steam flow at full load 8.59 x 10 l b/hr Secondary Side Pressure Orc 0 (Full Power Conditiond2 3

Pressure Drop across steam dryers Pressure Drop across nozzle outlet Total AP across secondary side of S.G.

(Feed ir:let to steam outlet)

Notes: 1) This heat transfer area is the minkum required area based on less than the installed number of stun generator tubes, and is consistent with the given outid.t conditions.

2) See coment 2.0.

3) The given downcomer volume is the volune in the downcomer i

from the top of the tubesheet to the normal water level.

(37.3 ft shove top of tubesheet). The volume in the annulus between the shell and the shroud above this normal water level is included in the steam dome region.

l 2-16 l

m

-ms

TABLE 2.5

' STEAM GENERATOR MATERIAL COMPOSITION

~

Steam Generator Vessel Carbon steel with surfaces in contact with Primary Coolant clad with stainless steel Tube Support plates Stainless Steel Primary Tubes Ni-Cr-Fe Alloy Stcam Separators Carbon Steel Steam Dryers Carbon Steel I

t l

I l

2-17

as.

-us s.

FIGURE 2.1 INTENTIONALLY LEFT BLANK l

d t

l l

l 1

2-18

=

TO MAIN STEAM TUABtNE

<l STk AM F LOW TRANSMITTERS STEA U FLOW S.GA w --

TR ANSP. tit 1 ERS S.G.B.

v v

L LT LT LEVEL TRANSMITTER g

/

'Lg VEL 5.GA TRANS?. TITTER D *LEVE L NSt.HTTER

/

LEVEL g

S.G A S.G.B.

TR ANSt.tTTE R S.G.B.

g q

f>OWNCOMER g

DOWNCOMER S.G A FEEDWATER FLON FEEDWATER FLOW TRANS*AITTER S.GA TR AN"M:TTF R I'

$.g.g' S.G.B.

\\m

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ECONCf.9tZE R OO WNCOfAER F E E D*./ ATE R FEEDWATER DOWNCOP.*E R FCONO *.*!ZE R CONTROL CONTROL F E E C'.'/AT E R F E ED?l ATE R - %

COf;1ACL CON 1HOL VALvh S.G.A.

VALVE $ GA

.1 VALM S G.B.

VALVE S.G.B.

TOTAL

', f

) f TOTAL FEEDWATEH FLOW TRANSMITTER p7

}

L gy FEEDWATER FLOW i

S.GA

(

) i TRANSMITTER g

S.G.B.

s g

rtEOwATER Puup g Figure 2.2 FEEDWATER FLOU PATHS WITH FLOW AND LEVEL TRANSMITTER LOCAT10iis liiDICATED l

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. T A

tO G

f4 W

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M N

• W t:

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t Esf 7

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t f

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=

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r

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C

=

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/

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NNtL Nt

=

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T '. t. n I0 Ii E3 I0oi S

tts l I g,'

lI Il !L

$2 OOtu R-

=

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i I T

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AA,,

I T:!O TEg e

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)

i

i\\

I

l PROGRAMS IN FWCS FULLOPEN

---

I l

l-o wz

.x

~N>$

w

~Ea*

l 002 zeo O

I o

C w

m r

00 I

o I

0 100%

07' FWCS FLOW DEMAND SIGNAL o

z MAX l

E l

w c

l

~o l

n.

b l

m I

o ww Q

• MIN 1.

a.E

=a 1

n.

100%

07" FWCS FLOW DEMAND SIGNAL 1

FIGURE 2.2 (Cont'o) 1 l

2-2T I

ll. -..

Figure 2.3 System 80 Steam Generator 4

i Shock Loss Factors and Flow Areas t

(See Consnent 2.0) a l

e

.1 1

2-22

l t

l.... !.

a.._... _._._i.

. J..

g t.

g.

i..

l..,. 1

.. _ _..,.m..g._....i..___._.__.;...

. i.

..s...

_..._..._.t

.i i...........i..... _.. _..........i........ _..

1...

. GGNG R ATOR.

.i..

.. '._i.* '._ F I G U R E *2.4 SYS Tl~M 80 STCA M

.q..;. i~

.i

.... 3 81 1 MWt Pi.AMT

.- +.-

1

..... C.s R Ct4 L AT.l0 Al. R ATI C i

._...s.

.S.......... _ __

t....

4...

. 4..,.

_.ty. s g c a ga r g_ o p o

-i

....q i

._. 3.

.a.

_.._.I

..a.

,..i i

... i.

... _ '..l

\\

..s 1

1 1

.....s.

...I m....... _ _

.....0.._r_..___

........i.

i......... _....

....:......t..

...........i..

t

....._..__._....._._i...._...._2._._..

..:........:.._.....i...._....i.

..r..............._...i i.,...._.. _.....

...t.

.... _.....i..,..

.l.

,.. i..

.... _...,...__1.,.t.._......

... _... _..... _ _.......t1...

.c...,....._._....

t i.

~

_._.s

. :...:.,.t._

.:.a.

2-23

-4 m 1

{,M i

3 5

N k

m%

4 i

w%rn 14,5 T3 u.

i l

2-24

- -- - "~

.. = *

- = ~

l SECTION 3.0

~

~

REACTOR COOLANT PLMPS b

],

GE0fETRY VALUE j

A.

Pump vole =e Table 3.1 i

8.

Effective puts volume flow area Remark #1 l

C.

Effective pu=p volume hydraulic diameter Remark #1 D.

Pu=p volu=e flow leng'th Table 3.1 Remark #2

.E.

Pu=p vole =e height Table 3.1 Remark #2 F.

Pump volume elevation Item E, Above G.

Pumpinlet(suction) 1.

flow area Table 3.1 2.

hydraulicdiameter(ft)

Table 3.1 3.

elevation Table 3.1 4.

forward flow energy loss coefficient Nbkf 3

5.

reverse flow ener,gy loss coefficient Taole 3.1 Remark #3 H.

Pumo outlet (disena ge)

?

l l

1.

flow area Table 3.1 2.

hydraulicdiameter(ft)

Table 3.1 3.

elevation Table 3.1 4.

fordard ficw energy loss coefficient Table 3.1 5.

reverse flew energy los; coefficient Neb.i Remark #3 3-1

II. PERFORMANCE VALUE A.

Ra:ed angular velocity Table 3.1 B.

Rated volumatric ficw Table 3.1 I

.,v C.

Rated head Table 3.1 D.

Rated pu=p tor:ue Table 3.1 E.

Rated pump motor tor ue Table 3.1 Remark #4 F.

Rated density Table 3.1 G.

Operating parameters for normal. steady state at 100%

rated plant conditions 1.

angular velocity See Above 2.

volumetric flow See Above 3.

head See.Above

4.

• pump torcue See Above i

5.

pump meter tercue Sca Above 6.

density See Above

,H.

Pump i.nd pumo motor cement of inertias Table 3.1 I.

Pump mater torcut vs. pump =otor speed table Table 3.6' J.

Pump fricticnal t:rque coefficients as Table 3.1 And a k #7 a function of pump angular velocity l

\\

~

.. ~

VALUE K.

Maximum forward and revers ptmp rotational velocities Table 3.1 e

Remark 15

~

~

L.

Single pnase nc=ologous pe=s data l.

Reautre 16 data ta:1es of tne independer.: -

variable vs. each dependent variable wita Table 3.2 &

definition of ter:s in variables given in Table 3.3 Table 1:

e e

1 e

e e

G G

O O

e e

e e

e e

O G

e e

e g

e 3-3 e

l o

TAELE 1.

7'.'"7.. *1030.*3 C'J'4VI DI.:In*TIONS De:endent Varia31e Regime Regime Mode indsoendent Nummer 10 Name a

y v/a Variaele Head Toreus

~

1 MAN Normal

>0

>0

< 1 v/=

h/=2 s/s2 2

HYN Pu=p

>0 70 7 1 a/v h/v2 s/v2 3

tiAD Energy

>0

<0

> -1 v/a -

h/a2 2

~

,f,2 4

HYD Dissication

>0

<0 "i - 1 a/v h/v2 s/v 5

HAT Hermal

<0

<0

< 1 v/s h/a2 2

2 s/=2 4

HYT Tursine

<0 70 31 m/v h/v '

s/v 7

HAR Reverre TO

> -1 v/s h/s2 s/a2 8

HYR Pu=p

[<0

>0 7 -1 s/v h/v3 s/v2 Rotational velocity) ratio.

(actual rotational velocity / rated o =

rotational velocity.

y 'o Vole =etric flow ratio.

(actual volumetric flow / rated volu=etric flow).

h o Head ratic.

(actual head / rated head).

so Tercue ratio.

(actual torcue/ rate torcue)

L M.

Four cuadrant curves (recuired only if single pnase homologous curves of Item 2-L of above are not availaole)'.

Not Required 1.

recuire data tables (or plots if data VALUE l

are not available) describing the pu=p characteristics in tar =: of:

volu=etric flow Not required rotational velocity head nnd torcue 4.

Two pnase pe=a data over 3-4

• ~

~

Reh._ aire fully ::egraded too :na:e -: :::;;;ut

'( 1)

. pues data (::nsisting of 15 cat

.:*as n :=t same fc =ct as tna: cescrice: i.,Ma.2-:.,

Table 3.4 witn a specific c:rrelation de: een voic 1

fracti:n and two pnase nead and

r:ue relative to single pnase nead and ::rcue (2)* If the data an'd correlation of Ita= 2-N-1 of above are not available, provide any available two phcse pu=p data and related c:rrelat'icn(s)

~

~

(

with ec=plete explanatcry infor=ation

- III. THETd'ALHYORAULIC (CONDITIONS FOR NORMAL STEADY ST TE AT 100% RATED PLANT CONDITIONS A.

Pucp volume VALUE 1.

average pressure Remark #8 2.

average temperature Remark #8 3.

average cuality SUBC00 LED LIQUID 5.

Pump suction and disenarge jynctions 1.

mass flow Table 3.1

! IV.

CONTROL LOGIC A.

Recuire all trip set;oints, logic, and interlocxs associated with the tripping off of eacn pu.:p Re m rk #6 1

o 8.

Recuire all reactor syste:s infor=atien needed to interpret the re:uested infer nation of Ite: 4-A Remark #6 l

r*

l i

3-5

-,,m.

,y._~..

REACTOR COOLANT PLMPS The following ren,trks qualify the data presented for this section:

4,5 1.

2.

3.

Fomard and reverse flow energy loss coefficients are not applicable to the rwactor coolant pumps. Pump homologous curves define RCP perfomance. Locked rotor K factors are supplied. (Table 3.1) 4.

Rated pump torque and Rated purnp motor torque are identical.

5.

Maximum fomard rotational velocity is considered to be equal to Rated Angular Velocity. When the hydraulic torque exceeds the design antireverse torque the pump is then allowed to rotate in the reverse di rection.

6.

Large and Small Break LOCA:

Off-site power is assumed unavailable at time of LOCA. Reactor Coolant Pumps trip at t = 0, for Large Break LOCA. For a Small Break LOCA* the RCP's are credited until a Reactor trip (pressure shutdown = 1600 psia). Figure 3.1 is used in our analyses only for Small Break.

Feedline Breaks:

It has been determined that, for feed line breaks.* the most limiting time for loss of offsite power is at the time of Turbine Trip.

Therefore, the RCP's are credited until the time of Turbine Trip.

Steam line Breaks:

For steam line break analyses

• at Hot Full Power, it has been determined that the loss of offsite power concurrent with the break of the steam line is the limiting case in terms of post-trip return to power.

For steam line break analyses

• at.iot Zero Power it has been detemined that the limiting case uit1 respect to return to power is the one without loss of offsite power.

3-6

REACTOR C00i. ANT PUMPS 7*

Pump frictional torque is supplied in Table 3.1 as Friction and g dg 8.

System pressure at the pump can be calculated from pressure losses provided in other sections. System temperatures are also specified elsewhere.

• CESSAR-F analysis performed by C-E.

1 3-7 l

REACTOR COOLANT PUMPS TABLE 3.1 3

1.

Liquid Volume (Per pump) 134 ft 2.

Elevation Difference between Inlet and Outlet Centerline 4.9583 ft 2

3.

Minimum Cross Sectional Flow Area at Pump Inlet and Outlet 4.91 ft 4.

Diameter, Inlet and Outlet 2.5 ft 5".

Elevation of Suction Nozzle Relative to Centerline of Cold Leg at Reactor Vest.11 Inlet Nozzle

-3'4" 6.

Elevation of Discharge Nozzle Relative to Centerline of Cold Leg at Reactor Vessei Inlet Nozzle 0 ft 7.

a) Forward Flow locked rotor K factor

-9.9 b) Reverse Flow locked rotor K factor

+16.2 8.

Rated Angular Velocity 1190 RPM 9.

Rated Volumetric Flow 111,400 gpm

10. Rated Head 365 ft

11. Rated Pump Torque 38786 ft - lb 3

12. Rated Density 45.87 lb/ft 2

13. Pump and Pump Motor moment of Inertia 136, 990 lb-ft

14. Maximum Forward Rotational Velocity 1190 RPM

15. Friction and Windage Torque 1866,0 ft - lb f

16.. Anti-Reverse rotation device failure torque 400,000 ft-lb f 3 __

TAGLE 3.2 l

2 HAN -

hla as a function of (vla) for vla i l nomal operation, po:itive flow, positive speed l

HVN -

hlv2 as a function of (mlv) for lv/al > 1 nomal operation, positive flow, positive speed HAD -

hla2 as a function of (vla) for lv/al 1 1 energy dissipation, negative flow,' positive speed MVD -

hlv2.as a function of (alv) for lv/al > 1 energy dissipation, negative flow, positive speed HAT -

hla2 as a function of (vla) for lv/al - 1 turbine operation, negative flow, negative speed HVT -

hlv2 as a function of (alv) for lv/al > 1 turbine operation, negative flow, negative speed HAR -

hja2 as a function of (vla) for lv/al 1 1 abnomal pump, positive flow, negative speed HYR -

hlv2 as a function of (alv) for lv/al > 1 abnomal pump, positive flow, negative speed BAN -

8/a2 as a' Enction of v/a* for v/a < 1 nomal operation, positive flow, positive speed SVN -

S/v2 as a function of a/v for v/a > 1 nomal operation, positive flow, positive speed BAD -

8/a2 as a function of v/a of lv/al 1 1 energ,v dissipation, negative flow, positive speed BVD -

8/a2atafunctionofa/voflv/al>1 energy dissipation negative flow, positive speed BAT -

8/s2 as a furiction of v/a of lv/al 3 1 turbine operation, negative flow, negative speed BVT -

8/v2 as a function of (a/v) for lv/al > 1 turbine operation, negative flow, negative speed BAR -

8/v2 as a functior, of (v/a) for lv/al 1 1 abnormal pump, positive flow, negative speed BVR -

8/v2 as a function of (a/v) for lv/al > 1 abnomal pug, positive flow, negative speed

• v/a is positive,v/a has values 0.0, 0.1, 0.2,......for all tables.

v/a is negative.v/a has values 0.0, -0.1, -0.2,......for all tgbles.

3-9

l TABLE 3.2 (Continued) 2 MANANC (h/a ) p = f(v/a) for v/a i 1 nomal ope-ation, positive flow, positive speed 2

HVNANC (h/v ) p = f(a/v) for v/m > 1 normal operation, positive flow, positive speed HADANC 2

(h/m ) p = f(v/a) for lv/al 1 1 energy dissipation, negative flow, positive speed (h/J)y.e f(a/v) for lv/al > 1 HVDANC energy dissipation, aegative flow, positive spaed 2

HATANC (h/a ) p = f(v/a) for lv/al 1 1 turoine operation, negative flow, negative speed 2

(h/v)p=f(a/v)forlv/al>1 HVTANC turbine operation, negative flow, negative speed l

2 HARANC (h/a ) p = f(v/a) for lv/al 1 1 atr.ormal pump, positivs flow, negative speed 2

(h/v ) p = f(2/v) for lv/al > 1 HVRANC abnomal pump, positive flow, negative speed l

3-10

TABLE 3.2 (Continued) v = Ratio of volumetric flow rate to the rated volumetric flow rate v = Q/QR a = Ratio of the pump speed to the rated pump speed a a w/wR S = Ratio of hydraulic torque to the rated hydraulic torque s a Th/TR h = Ratio of the pump head to the rated pump head h = H/Hg

(

l (h/a ) p Difference between single phase and 2

- degraded two-phase heads 2

i (h/v ) p l

l 3-11

TAkE 3.3

\\

SINGLE PHASE HOMOLOG 0US PLMP DATA i

t-l Representative of:

System 80^-

t l* lori"l 0

.1

.2

.3

.4

.5

.6

.7

.8

.9 1.0 i

HAN (h/32) i 1.30 1.30 1.29 1.28 1.27 1.24 1.21 1.17 1.12 1.08 1.00 HVN (h/v ). -1.08 -

.92

.75

.56.

.36

.17

.06

.27

.50

.7o 1.00 1

2 BAN (8/a )

.68

.71

.73

.76

.80

.83

.86

'. 89

.92

.96 1.00 1

2 BVN(8/v)

.82

.58

,.40

.24

.08

.08 -

.26

.44

.62 80 1.00' i

l HAD (h/a )

1.30 1.38 1.56 1.75 1.98 2.24 2.52 2.80 3.10 3.40 3.72 z

HVD (h/v )

j,77 j,94 2.b8 2.16 2.24 2.32 2.40 2.52 2.70 3.30 3.72 2

w 2

l

.L BAD (s/a )

68

.82 1.00 1.23 1.48 1.76 2.D6 2.34 2.66 -

2.98 3.32 f

ro

,.99 2.12 2.19 2.20 2.24 2.39 2.58 2.80 3.04 3.32 BVD(s/v) 1.78' 1

2 i

HAT (h/a )

.290

.300

.310

.34

.35

.400

.48

.58 69

.87 1.08 2

t l

2 HVT(h/v) 1.77 1.58 1.40 1.26

• 1.19 1.14 1.09 1.07 1.05 1.06 1.08 s

z BAT (8/a )

-1.34

-1.14

.95

.74

.54

.37

.20

.02

,.16

.34

.52 l

~

.84

..70

.52 j

8VT (s/v )

1.78 1.58 1.42 1.30 1.20 1.12 1.06

.86 2

l HAR (h/a )

.290

.280

.26

.20

.130 0

.168

.41

.60

.907

-1.207 2

HVR (h/v )

-1.08

-1.18

-1.26'

-1.32

-1.34

-1.35

-1.36

-1.35

-1.31

-1.26

-1.207 2

2 BAR (S/a )

-1.34

-1.54

-1.82

-2.09-

-2.42

-2.70

-3,03

-3.36

3.74

-4.10

-*.51

' 1.38

-1.68

-2.00-

-2.33

-2.70

-3.07 ^

-3.50

-3.97

-4.51 BVR (8/v )

. 11 2

-1.07 2

- u -.....

- 3 Tabla.3,4 HOMOLOCOUS PUMP DATA FOR THE DIFFERENCE BETWEEN SINGLE-PHASE AND DECRADED TWO-PHASE HEADS j

lh or lh 0

.1

'2

.3

.4

.5

.6

.7

.8

.9 1.0 1.017 1.014 1.012 0.983 0.940 1.000 HANANC 0.000 0.833 1.089 1.098 1.048 HVHANC 0.000

-0.040 0.003 0.095 0.210 0.231 0.449 0.560 0.674 0.804 1.000 iMDANC 0.000

-0.495

-1.295

-2.090

-2.667

-2.909

-2.786

-2.360

-1.773

-1 242

-1.111 H'JDANC '

O.110 0.083 0.054 0.020

-O.023

-0.084

-0.174

-0.307

-0.501

-0.776

-1.111 iMTANC 0.000

-0.178

-0. 345

-0.504

-0.652

-0.792

-0.927

-1.058

-1.188

-1.323

-1.468

- HVTANC 0.110 0.134 0.151 0.152 0.130 0.071

-0.043

- 0.230

- 0.512

- 0.915

-1.467 HARANC 0.000 0.667 0.811 0.785 0.761 0.734 0.701 0.655 0.591 0.507 0.380 3

w HVRANC 0.000

-0.043

-0.066

-0.071

-0.058

-0.027 0.021 0.086 0.168 0.266 0.380 e

e h

e e

e e

0 U

gr, k

y

'[j Table 3.5 PLHP LEAD DECRADATION MULTIPLIER VS VOID FRACTION Void Fraction (a)

Degradation Multiplier *

~

0.0 0.0 0.10 0.0 0.15 0.05 0.24 0.8C 0.30 0.96 0.40 0.98 0.60 0.97 0.80 0.90 0.80 0.90 0.36 0.50 1.00 0.0 Y

e 3-14 k_-

m Table 3.6 Pump Motor Speed vs. Pump Motor Torque Speed Electrical Torque Rad /Sec lb - ft 0

46522 37.38 48460 74.76 53500 87.22 60090 117.1 96920 117.7 98858 118.4 93043 119.6 85290 122.1 62029 124.6 0

l 127.1

-62029 129.6

-85290 130.8

-93043 131.5

-98858 132.1

-96920 162.0

-60090 174.4

-53500 211.8

-48460 249.2

-46522 3-15 ad

SECTION 4.0 PRESSURIZER

/

1.

Tant VALUE A.

CD Schematic 4.1 3.

ID '

Schematic 4.1 C.

Heignt Schematic 4.1 D.

Total in:ernal vole =e including structural =aterials Table 4.1 1

E.

Flow area as a function' nf neigne Table 4.1 F.,

Cc=;:csitica table 1.11 IT. Surge Tine A.

CD pipe Table 4.1 8.

ID pipe Table 4.1 C.

Rougnness N/A D.

Forward 11cw energlf 1 css coefficient Table 4.1 /

E.

Reverse ficw energy loss ccefficient Table 4.1 F.

Pipe length G.

Nu=:er of elbews Table 4.1 H.

Elevattens 1.

Not leg connection Table 4.1 2.

Pressurf:er connectica Table 4.1

[II. Surge line ne::le Remark #1 A.

E1evation of no::1e inlet cantsritne at entrance to pressuri:gr H.2 Above I.

ID no::le inlet Same as Table 4.1 C.

ID prussuri:er inlet IIB above Table 4.1 0.

. Lang:n of no::le at no::le centerfine Part of IIF g.

Fer,ard ficw energy loss coefficient at above pressurizer inlet PART OF SUR2 LINE K FACTOR 4-1 7

VAlljF.

T.

K,ever:t f*:w energy lo:s ::cfficier.: a; Same as Above pres:.ri:gr ints:

l G.

Inske surf ace r:ugnness N/A l '/.

Safety r,c::les - REFER TO SAFETY VALVES A.

Flew area Tabie 4.1

• 8.

Flow resistance Table 4.1 C.. Maxi==:n no::le capacity Table 4.1 D.

Operatienal setpoint Table 4.1 1

Y.

Relief ne::le System 80 A.

Flew area does not 8.

Flcw resistance empi y relief valves.

C.

Maxt=u.m n:::le capaci:y D.

Operati nal setpoin 11.

Prc:surizer heater?,

A.

Nu:::::er of r:ds Tabie 4.1 E.

Outer dia. eter of red Table 4.1

~

l C.

Total heat transfer surface area Table 4'.1

.D.

Pcwer input Table 4.1 E. '.Ceg :siticn Table 4.1 F.

Heater setpoints Table 4.1 & 4.2 1

1.

Mor:al coera:1en l

2.

Trans f ere: cpera:1en Table 4.2 VII. Operati:nal A.

Water vole =e as a functi:n cf lead Figure 4.1, 4.2 8.

Operating c:ndittens as a fun::icn of load 1.

Pressure Tabie 4.1 2.

Te-;erature T

f r above SAT pressure 3.

Quality N/A l

l 4

Eer:n c:ncentrati:n Rema rk #2 t

4-2

4 VALUE J.

.=

i i

l Spray line volume flow as a function 'of AP-l I*

Table 4.1 l

i

. Spray line st:: ir.::

Table 4.1 & 4. 2 1.

1 r.a.. 1.., e r a. 4..n e.

e.

4 i

J l.

2.

transien: ::tet:f=n l

• W e

e e

1a i

J.

i!

L

]

I 1

5 e..e

.m.

4M 8

4 i

, e i

1 I

1 e

4 e

e t

,e i

e o

e 0

O s

b 4

e n 9.,

e 4

O e

e 4-3

P ressurizer The following remarks qualify the data presented for this section:

1) The Surge Line nozzle is modeled as part of the Surge Line.

~"' '..-

Therefore, the nozzle ID's and length, as well as the nozzle K Factor, will be included in the data presented on the Surge Line.

2) The Baron Concentration of the pressurizer will be equal to the l

l equilibrium concentration of boron on the RCS due to the continuous l

pressurizer spray flow of 1.5 gpm.

I l

\\

l S

l l

?

4-4

- --n.n.. _

1 l*lJ' i

I i..

?{ I!!!

lI

'ili-v

. f ' w'"'33

[.

s.

mv

• s

'h i i

I l

N;Nh.

)

'ii.W.I u.l.

.,..v;:..i

..Lh I

e s i

g. - g Jas

.18 l

it i >,

il ' 1 I

'I l.2 e

l.,

{

3 1 t.

e r

t s

._j +

i i

. l I

l I t.'

l l

3

}

l.

1 s

i r

l. *

.,. - _--t !.

. [

,.l.

3 r.

i 3,

l

/

g I

l'

]l 8

' 8 I

I k

4

s I

'II r

.?

/.

'j l

• a:tl"

(.o F-9 E

f.4 C4

(

3

}.

5-u.J L.J J '>-

a I

2*

..l th

=: ct: to 4

tJ C

~,l

- f Cn s,

5

,m

[

7 i-w l

!..T l-l 1;. C l

e I

.I,,

r

,ia l;

I

- 't.E, g

g

+

ir It is -

i

(

• • • • • e l

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. If

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TABLE 4.1 Pressurizer Tank 3

900 ft Liquid Volume (100" Power) 3 1800 ft Net Free Volume 2

50.53 ft

. Cross Sectional Area 8.96 ft Tank Outside Diameter 8.02 ft Tank Inside. Diameter 42.47 ft Tank Height Tank Pressure Nominal 2250 psia Pressurizer Surge Line l

1.0625 ft Pipe OD 0.844 ft Pipe ID 5

Surge Line K-Factor 0

Friction Factor (Based on Reynolds number of 10 )

0.0089 "

Hot Leg Connection Relative to Centerline of Reactor 3.89 ft Vessel Cold Leg Pressurizer Connection Relative to Centerline of Reactor 9.08 ft Vessel Cold Leg 3 - 90*

Elbows in Surge Line 1 - 120*3*

34.48 ft Surae Line Volume 61.635 ft Surge Line Length Pressurizer Safety Valves 4

2 0.030 ft

. Maximum Flow Area I

Flow Rate (Per Valve)

Max-159.72 lb/sec Min-127.78 lb/_sec 2500 psia 21%

0Rening Pressure 3% opening pressure Accumulation 5% opening pressure Blowdown Pressurizer Heaters l

l 36 Number of Rods 0.104 ft.

Rod OD 1800 kw Power Input (Nominal Load) 68.75 inches Heated Length of heater

' Heater material composition ~

Type 316 stainless steel 4-6

Table 4.1 (continued)

Pressurizer Sorays Maxt=um Proportional Sp' ray Flow P. ate 375 Spa Mt.nimum spray flow rate 1.5 gpm

. Opening pressure of proportional spray valves 2275 psia Full open pressure of proportional spray valves 2300 psia Nominal Set Pressures for Proportion Heaters:

Full ON at 2225 psia Full 0FF at 2275 psia Set Pressures for Backup Heaters:

Backup heaters ON below 2200 psia Backup heaters 0FF above 2225 psia

• These numbers are plant specific and will vary.

• The initial pressurizer water volume used in LOCA 3 (1) analysis is 950 ft In steam and feedline break analysis,(I) it is assumed that the initial pressurizer water level can vary between 26% and 60% of the distance between the upper and lower level instrument taps. (See Schematic 4.1)

• *

• These values are used in C-E's Safety Analysis. (I)

(The forward and reverse surge line K factors are assumed to be equal.)

(1) CESSAR-F analysis performed by C-E.

e 4-7 l

y Table l l Pressurizer Heater Failures All pressurizer heaters excluding 300 KW of backup heaters are supplied by a non-vital buss (i.e. only 300 KW of backup heaters have emergency power). All heaters fail off on a loss of power. A low pzr. level signal also securas all heaters. A high level error signal turns all heaters on. Therefore a single failure causing a high level error can give 100%

heater output and a failure causing a low level error can give 0% heater output.

Pressurizer Sprays The pressurizer sprays are inoperable any time reactor coolant pumps are not operating and following pump cavitation caused by a loss of system pressure.

The spray valves also fail closed on a loss of actuating signal. Spray flow rate is controlled by the Pressurizer Pressure Control Systen. Therefore 100; spray flow can occur with a single failure in a Pressurizer Pressure Control Char.nel.

4-8

4 4

e O

621.2 620 O

e' 600 E

TH 592.85 m

h580 T

M AVG

@ 564 564.5 (T

" 560 g

c au 5@

520 0

20 40 60 80 100 STEAM GENERATOR POWER, PERCENT OFWARRANTY

=

(

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6 e

50 46 44 M 42 O

$ 40 a.

r 38 W

."J 36 u,

34 a::

US 32

~

Og 30

~

28 26 24 0*

REAC OR C00 NT AVER GE TEMPERATUR OF l

TYPICAL PRESSURIZEy LEVEL SETPOINT PROGRAM 4.2

~

. ['..,

SECTION 5.d 2

EMERGENCY CORE COOLMIT SYSTEMS

.,I, Accumula:Or VALUE

)

A.

Tants Table 5.1 &

1.

Number of tant:

Remark 01 2.

CD Schematic 5.1 3.

ID~

Schematic 5.1 4.

Total volume Table 5.1 5.

Heignt schematic 5.1 6.

Flow area as a function of neign:

Table 5.1 7.

Composition Table 1.11 8.

Surge line 1.

Junction flow area Remark #4 Table 5.1 2.

Pipe 00 3.

Pipe ID Table 5.1 N/A 4.

Total lengtn 5.

Forward flow energy loss c: efficient Table 5.1 6.

Reverse flow energy loss coefficient N/A l

7.

Roughness N/A 8.

Elevations a.

Tant connection c

Table 5.1 b.

Cold leg connection C.

Operatio'nal (Botn ncminal and upper and 1cwer limits) l 1.

Licuid level use volume

,2.

Licuid volume Table 5.1 6

e e

5-1

vntut 3.

0:erating conditions a.

ressure Table 5.1 3.

se=cerature-Table 5.1 c.

Baron concensration Table 5.1 A.

F$11 gas a.

Composition Table 5.1 D'.

Volume Table 5.1 HPIS 1.

Injection licuid conditions 4.

Pressure Table 5.1 b.. Te :perature Table 5.1 Table 5.1 c.

Boron concentration 2.

Flow rate as a function of primary Tables 5.2

• 5-system pressure and 4 pumps

& Remark #3 3.

Number of. pu=ps.

Table 5.1 4.

Operational setpoints Table 5.1 III. LPIS l

1.

Injecticn licuid conditions 4.

Pressure Table 5.1

n. Temperature Table 5.1 c.

boron concentration Table 5.1 2.

Flow rate as a function of primary Table 5.3 &

Remark #3 system pressurc and ! pu=ps 3.

Number of pumps Table 5.1 4.

Operational setpoints a.

actuation time delays Table 5.1 4

e e

e

~

5-2

=

T.'.

Cnarping system VALUE e

1.

Injecti:n licuid conciti:ns a.

P,ressure o

b.

Te:::erature e

c.

Entnalpy 2.

Flow rate as a function of primary Remark #2

~

system pressure and i pt=ps 3.

Number of pu.~ps 4.

Operational setpoints,

e 9

e e

e e

s e

O

• Me e

ee a

e e

e s

9 e

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e

Emerginey Core Cooling System The following remarks qualify the data presented in this section and, addition-ally, describe the code modeling assumptions made by C-E with respect to the Safety Injection System.

1.

There are 4 Safety Injection Tanks, (SIT) however, only 3 are taken credit for in a LOCA analysis *.

In the CESSAR Steam Line Break analysis, all four Safety Injection Tanks would be credited, however, the system pressure does not approach the SIT pressure.

2.

There are 3 charging pumps (2 on line), each of which delivers 44 gpm.

Although credit is not taken for these pumps in C-E's analyses *, the charging pumps can be started manually by the operator.

3.

In perfonning safety analysis calculations *, conservative assumptions are made concerning the availability of safety injection flow.

It is assumed that offsite power is lost and all pumps must await diesel startup before they can begin to deliver flow. For breaks in the RCS pump discharge leg, it is also assumed that all safety injection flow delivered to the broken cold leg spills into the containment.

Large Break LOCA:

A C-E analysis of the possible single failures that can occur within the Safety Injection System (SIS) has shown that the worst single failure for the large break spectrum is the failure of one of the low pressure pumps to start. This results in a minimum amount of safety injection water, available to the core, without affecting the operation of the containment

~

spray system.

Therefore, based on the above assumptions, the following safety injection flows are credited for the las ge break analysis:

Two high pressure safety injection pumps (HPSIP's) are piped so that each one can feed all four cold leg injection points. Thus:

a.

for a break in the pump discharge leg, the safety injection flow credited is 75% of the flow from two HPSIP's since it is assumed that all injection in the broken cold leg is spilled.

b.

for breaks in other locations, the safety injection flow credited is 100% of two HPSIP's.

Two low pressure safety injection pumps (LPSIP's) are piped so that each one feeds two cold leg injection points. Thus:

a.

for a break in the pump discharge leg, the safety injection flow credited is 50% of the flow from one LPSIP. The bases for this i

l flow is that only one LPSIP is operable (worst single failure) and one of the two injection points for the operable pump is located in the broken loop, and thus, that flow is spilled.

l

• i.e. the CESSAR-F analysis performed by C-E.

5-4

b.

for breaks in other locations, the safety injection flow is 100% of one LPSIP.

Four safety injection tanks (SIT's) are piped so that each SIT feeds a single cold leg injection point. Thus:

4 for a break in the pump discharge leg, the safety injection flow credited is 100% flow from thme SIT's since it is assumed that all injection in the cold leg is spilled.

b.

for breaks in other locations, the safety injection flow credited is 100% flow from four SIT's.

Small Break LOCA:

/

A C-E analysis of the possible single failures that can occur within the SIS has shown that the worst single failure for the small break LOCA spectrum is the failure of one of the emergency diesels to start. This failure causes a loss of both a high pressure pump and a low pressure pump and results in a minimer6 of safety injection water being available to cool the core. There-fore, based on the above assumptions, the following safety injection flcws are credited for the small break analysis.

Since each high pressure safety injection pump (HPSIP) is piped so that it can feed all four cold leg injection points:

a.

for a break in the pump discharge leg, the HPSIP flow credited is 75%

of the flow from one HPSIP. The remaining 25% is assumed to spill out the break.

b.

for breaks in other locations, the HPSIP flow cmdited is 100% of one HPSIP.

Since each low pressure safety injection pump (LPSIP) is piped so that it feeds two of the cold leg injection points:

for a break in the pump discharge leg,'the LPSIP flow credited is 50%

a.

of the flow from one LPSIP. The remaining 50% is assumed to spill out the bmak.

b.

for breaks in other locations, the LPSIP flow credited is 100% of one LPSIP.

The four safety injection tanks (SITS) are piped so that each SIT feeds a single cold leg injection point. Thus:

a.

for a break in the pump discharge leg, the SIT flow credited is 100%

of the flow from three SITS. The remaining SIT is assumed to spill out the break.

b.

for bmaks in other locations, the SIT flow credited is 100% of four SITS.

5-5

Steam Line Break:

A C-E analysis of the possible single failures that can occur has shown that the failure of one HPSI pump to deliver flow

• is the most adverse single failure for all the SLB cases except the SLB initiated at full power with no loss of offsite power (for this case, the failure of one MSIV to close on the unaffected steam generator is the limiting sir.gle failure).

4.

For the junction flow area, use the surge line area.

• The system pressure during a steam line break remains well above.

the shutoff head of the LPSI pumps.

If system pressure was to approach the shuteff head of the LPSI pumps, it would be appropriate to consider the failure of a diesel generator, as this would disable both a HPSI and a LPSI pump.

l 5-6 l

Sch:matic S.1 SYSTEM 80 SAFETY INJECTf0N TANK 4Md MF d*4/FTAW4 M

+

ASM t at,La*ficAL ndAO ~ 2;l RATN;>

W 1,

y

_us,e_mu, nuo ma sm TAN LI R l.

=

o mg J.g, 3434 Md74L / penc,eed (non now)

P,~

aumus. g nw w

i o

s I

l n

.-<, m_

. -s MYg

{ted) 000 l

._n 0

nsist c.,>

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• T ao 6 I

rm i~e-

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W'b W~Y

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,.. n' L0-

.,-.3 BOfM 34063 Cl3 6094 Add ALL DIMENSIONS IN INCHES ELEVA T/0N V/EW 5-7

-- - _ _- _ J

TABLE 5.1 Nunber of Tanks 4

3 Total Volume (1 Tank) 2406 ft 2

Cross Sectional Area of Tank 63.62 ft Liquid Volume (1 Tank)

Min:

1790 ft3 Max:

1927 ft3 Nom:

1858 ft N Gas Pitssure Min: 608 psia

• 2 Max: 647 psia Nom: 626 psia Temperature of Fluid 60 - 120*F Elevation of Discharge Mczzle of Safety Injection Tank Relative to Reactor Vessel Outlet Nozzle Centerline O ft assumed for conservatism I

Forward Flow K-Factor 7

Reference Diameter ID 0.9323 ft Safety Injection Tank Boron Concentration (nominal) 4400 ppm Nunter of HPSI/LPSI Punps 2 of each type Delay time for System Availability after

$ 30 seconds - small break SIAS is initiated and assuming LPSI & HPSI l

the loss of offsite power.

LOCA and steam line break.

l Large break LOCA: No pumped injection until SIT empty.

l l

Safety Injection Tank Line OD 14" SCHEDULE 140 (Conditions given below are for the source of the HPSI and LPSI HPSI & LPSI pumps, i.e., the REFUELING WATER TANK (RWT))

pressure atmospheric temperature max - 120*F SIAS for Low Pressurizer pressure See table 1.12 Boron Concentration (RWT)

Nominal 4000-4400 ppm Minimum 4000 Maximum 4400

• These values are used in the CESSAR-F Analysis perfonned by C-E.

I 5-8

Table 5.2 Minimum HPSI Flow vs. Back Pressure

• Pressure of RCS (psia)

Flow Rate to RCS from 1 Pisnp (No Spillage) (gpm) 14.7 1000.0 324.7 900.0 619.7 800.0 879.7 700.0 1109.7 600.0 1284.7 500.0 1454.7 400.0 1559.7 300.0 1664.7 200.0 1744.7 100.0 1789.7 0.0 The data is the total flow for one HPSI pump.

5-9

Table 5.3 Minimum LPSI Flow vs. Back Pressure

• Pressure of RCS (psia)

Flow Rate to RCS from 1 pump (no spillage) (gpm) 14.7 4500.0 64.7 3840.0 114.7 3000.0 164.7 1760.0 194.7 600.0 214.7 0.0

• The data is the total flow for one LPSI pump.

l l

5-10

SECTION 6.0 PRIMARY COOLANT P RING SYSTEM The primary coolant system consists of the piping leading from the reactor vessel to the steam generators, from the steam generators to the reactor coolant pumps, and from the coolant pumps to the reactor vessel inlet nozzles. To adequately model the piping, infomation concerning flow direction changes, the presence of valves, elbows, tees and changes in flow areas must be adequately described.

In the attached tables the locations where flow conditions change in a piping section are requested using a cylindrical coordinate system. The origin of the coordinate system is at the intersection of the rea:: tor-vessel axial centerline with a utility-designated mactor vessel inlet nozzle horizontal centerline. Angular referunces are counterclockwise and will be with respect to the referenced inlet nozzle. The relationship of the cylindrical coordinate system to the reactor vessel is shown schematically in Figure 1.

An example of how piping component locations would be specified is as follows. A section of primary piping connecting a reactor coolant ptsnp to a rwactor inlet nozzle is shown schematically in Figure 2.

For simplicity, assume the piping does not have any flow area reductions, changes in inner surface roughness, valves, or piping penetrations. The angular orientation of the inlet nozzle with respect to the reference. inlet nozzle is 270*.'

The various coordinate locations should be systematically specified, preferably starting where the coolant enters this section of piping and ending at the inlet to the reactor vessel inlet nozzle. The convention of starting at the normal coolant flow inlet and ending at the normal coolant flow exit should be followed throughout the reactor coolant system piping dsscription.

The first coordinate location to be specified is at the coolant pump discharge nozzle exit. This location is also at 270* orientation, and is 30.0 ft from the reactor vessel axial centerline and 15.0 ft below the inlet nozzle centerline.

Its coordinate location is therefore 270* -

30.0 ft. -15.0 ft.

6-1 l

The next coordinate location of interest is the 90' elbow where the coolant flow direction changes from horizontal to vertical. As shown in the figure, the only coordinate that has changed is the distance, r, from the reactor vessel centerline. The location of this 90' elbow is therefore 270*, 22.5 ft, -15.0 ft.

The coordinate location of the next elbow is 270*, 22.5 ft 0.0 ft.

This is assuming all inlet nozzles on the reactor vessel are at the same I

elevation. Note that the only coordinate value to change was the elevation

(-15.0 ft to 0.0 ft).

The final coordinate location is the inlet to the reactor vessel inlet nozzle, which is at 270', 12.5 ft. 0.0 ft.

For changes in coolant flow direction of greater than 90*, the section of piping should be divided into two or more sections, such that no one section represents more than a 90'* change in direction.

l l

l l

6-2

I g

L h

T=J

<k i

'~

q k k_*~

reference inlet nc: le P

e s

s_)

8 r

l

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V 1

1.

\\

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Figure k Piping systes cylindrics1 ceerdinate sys:c.

1 6-3 I

1 l

I

.i s

1 I

L__.12.sft 10 0 ft

~

j l

.,F~"

3-i 9

r.-

15.b'ft l

t

)6."

l Q

- 7.5 ft mj l

c 1

l.

l Figure 2.

Piping to reactor vessel inlet noz:le.

i l

i 6-4

1 lay:ut of tne S :cition to in"or=::icn :en:c-nin; :ne :ays':: to a::urately c.::c; primary :ising, several c:ner items of ca:a 2-1. te:t:

Tnese are:

l tae.sys::: at the soecifice points cf interes.

l l

1.

Flow area

2., Pipe inside rougnness, 3.

Hydrau1(c diameter 4.

Forward flow energy loss coefficient (s) 5.

Reverse.flev energy loss coefficient (s) 5.

If an area change, abriapt or s=coth 7.

$omal condition.ccolant pressure 8.

Nemal cenditieri coolant temperature 9.

Nomal condition ecolant fluid cc=penent flow rata r

10. Homal conditicn coolant vapor ce=ponent ficw rate l

11. Piping material. e.g. 55-306, inconel X750, etc

12. Piping tnicxness

13. Reasen for descriptien, e.g. cotor valve, pipe penetratien, piping tee, etc The Table 2 is provided to sicolify the presentatien of tne data.

first column is for a user-supplied reference nus:er 'in :he event there is a need fer furtner infomatien. Fcr exz ple, if tn,ere is a primary icep 6-5 1

\\

2 e

In: 1::a*i:n *f soit; ion valve, 1::iti:nal ' inform::i n seule =e recuire:.

llec i:E: eou:cer.

It is

ne valve woule ce s:1:ified using :nc user-seco

.ecues ad sna :ne t a: nucocrs de unicuely ;ecified to eliminate ;:ossicle sisintrapre a i:ns c,f cata.

e e

e 9

9 e

9 4

e e

e S

e e

O O

e o

e e

e e

g e

e e

e g

e e

e e

en0 e

e

(

l e

e e

o e

O t

e e

l O

e e

9 9

O e

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e M

4

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=

6-6 e

ta

s.

1 ls PRIMARY COOLANT PIPING RESPONSE Physical Layout

• The physical layout of _ the System.80 primary piping is presented in Figures 6.'l-6.9 and Figure 6.11.

From these fi ures, direction changes, nozzles, (Table 6.1 givss nozzle identification and elbows can be located.

Additional Information The additional information requested about the Primary Coolant Piping System is contained in Tables 6.'2 and 6.3,and Figure 6.10.

Table 6.2 and Figure 6.10 provide informat!on on RCS pressure losses. Table 6.3 provides the coolant volumes in each pipe' leg; 6-7

Table 6.1 N0ZZLE DESCRIPTION (used in Figures 6.2 - 6.9)

Nozzle Size Use A

12" SCH 160 surge B

14" SCH 140 safety injection C

16" SCH 160 shutdown cooling outlet D

2" SCH 160 charging inlet E

2" SCH 160 letdown and drain F

3" SCH 160 spray G

0.75" SCH 160 pressure measurement or sampling l

H 1" SOCKET RTO (Resistance Temperature Detectors) 6-8

TABLE 6. 2 I

Standard Plant RCS Data SECTION*

tRICTIONAL FORWARD FLOW REVERSE R0W FLOW FLOW DIAMETER DENSITY FLOW

'sEMP 6

LOSS ***

• SH0CK LOSS *** SH0CK LOSS *"* LENGTH AREA (ft)

(Ibs/ft)

(x10 1b/hr)

(*F) 2 (PSI)

(PSI)

(PSI)

(f t)

(ft )

4 l

-5

_s 1-2 7.03 9.62 3.5 41.00 82 622.6 2-3 7.03 9.62 3.5 41.00 82 622.6 1

3-4 0.0 l

11.45 3.82 41.00 8.!

622.6 4-5 0.0 l

26.61**

5.82**

41.00 82 633.6 5-6 63.2 26.61**

5.82**

43.68 82 593.8 6-7 0.0 i

26.61**

5.82**

45.86 82 565 l

,7-8 0.0 4.91 2.5 45.86 41

'565 d> 8-9 12.16 4.91 2.5 45.86 41 505 9-10 12.16 4.91 2.5 45.86 41 565 10-11 11-12 9.65 4.91 2.5 45.86 41 565 12-13 9.65 4.91 2.5 45.86 41 565 1

I

• See Figure 6.10

• • Equivalent for 11,000 tubes ID = 0.666 inches

• • • 20% uncertainty added to all shock losses

• • • • 10% uncertainty added to all frictional losses.

NOTES: 1) Absolute pressures are not presented in this table, however, by using this table such process point data can be determined.

2) This table uses maximum design flow for conservatism.

8 9

l TABLE 6.3 PRIMARY (X)0LANT PIPING VOLLME 3

Hot Leg 1 (includes elbow) 135.27 ft 3

Hot Leg 2 (includes elbow) 135.27 ft 3

Suction Legs 1, 2, 3, 4 (including elbows) 119.38 ft 3

Cold Leg 1 (includes elbow) 94.74 ft 3

Cold Leg 2 (includes elbow) 94.74 ft 3

Cold Leg 3 (includes elbow) 94.74 ft 3

Cold Leg 4 (includes elbow) 94.74 ft 6-10

FIGURE 6.1 SYSTEM 80 TOP VIEW m ~. o o cold leg 4 cold leg 3 pump suction s *,

ide elbow

~

\\

suction leg 3 suction leg 4 9

9 pump section team gen-steam genera or side elbow side elbow erator sid N

30*

w-cibow sN hot eg 1 e'

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hot leg 2.

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cold leg 1 pump suction elbow

,a side cibow cold leg 2 nwn es l

au i

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O FIGURE 6.2 HOT LEG 1 A64.29*

NC&!l! 6 If5

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mi V.)

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I 6-13

FIGURE 6.4 SUCTION LEGS 1, 2, 3, 4 ist orr.m*

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sr.or* aan

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i FIGURE 6.5 COLD LEG 1 1

w$,

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FIGURE 6.6 COLD LEG 2 t

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=

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l FIGURE 6.7 COLD LEG 3

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FIGURE 6.8 COLD LEG 4 164b*sogg

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FIGURE 6.9 COLD SIDE ELB0WS l

.wesnarer - pm It

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PLNP SUCTION SIDE ELBOW SUCTION LEG av _

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SIDE ELBOW SUCTION LEG 6-19

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FIGURE 6.10-t PIPING LOSSES E

• SYSTEM 80 so

}

PR

%)

PUMP gSURGE l

HOT LEG 5

6 l

RV suction coto tec 8

i LEG IOC I

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11 12 13 j

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OUTLET RV OUTLET N0ZZLE 8.

OUTLET SG OUTLET N0ZZLE 2.

MIDPOINT HOT LEG 9.

MIDPOINT SUCTION LEG f

3.

INLET SG INLET N0ZZLE

10. PUMP INLET l

l 4.

MIDPOINT SG INLET PLENUM

11. PUMP OUTLET i

,l S.

INLET TUBES

12. MIDPOINf COLD LEG 6.

DUTLET TUBES 13.

INLET RV INLET N0ZZLE

[

7.

MIDPOINT SG OUTLET N0ZZLE e

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6-21 I

i 7.0 S v 57 E *'. Y r..~ E :

A secara a sec f n is provide: f:r ces:ri:-i:n :? :ne stricus vaiees in :ne crimary and se::ntary systems. Inis nas seen ::ne s :na: :ne a:di:1:nal infcrma:icn recuired f:r descri:i g :ne vari:us valves c:es nc:

cause unnecessary clu::er.

.Tne casic information recuired for all valves is as follows:

1.

Location of valve in system a.

c:mponent name, or b.

item number, if located in primary piping (refer :s secti:n on primary piping) 2.

Valve type a.

eneck valve b.

inertial check valve (flapper) c.

motor valve d.

servo valve 3.

Valve flow area in full open position 4

Forward flow energy less coefficient (s) i 5.

Reverse flow energy loss c: efficient (s) j 6.

Presence of any flow area enange 1

7.

Succ:oled discharge c: efficient l

l 7-1

5.

T :-Snase :isenarge coefficien:

9.

.i:rmal c:nditions fluid f1:w rate

10..ior.a1 conditions vapor flow rate Specific information related to a particular type of valve is given below.

1.

Check valves L

a.

presence or absence of hysteresis b.

normal valve position--open or. closed c.

closing backpressure

.d.

leak ratio--fraction of valve area wnen valve is normally closed 2.

Inertial enect valves (see Figure 3) a.

repeatability of operation n.

initial valve position--open or closed c.

closingDackpressure(P) d.

leak ratio--fraction of valve area wnen valve is normally closed initial flapper angle (e,)

e.

f.

minimum flapper angle (emin) g.

maximum flapper angle (e,,,)

h.

mcment of inertia of flapper 1.

initial angular velocity (w) j.

moment arm lengtn of flapper (L) k.

radius of flapper disk l

1.

mass of flapper (W) 7-2

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n

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--- - - - - - ----^--

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3.

tiotor valve a.

conditions initiating motion b.

conditions terminating motion c.

valve change rate--either (1) rate of change of the normalized valve area as the valve opens and closes, or

~

(2) rate of change of the normalized valve steam position d.

initial position of the val.e e.

If 3.c.2 is given--normalized valve area as a function of normalized steam position 4.

Servo valve--use one of the following a.

normalized valve area as a function of the controlling parameter (s) b.

normalized steam position as a function of the controlling parameter (s) 5.

Motor and Servo valves--for smooth area changes only a.

forward flow energy loss coefficient (s) as a function of normalized steam position b.

reverse flow energy loss coefficient (s) as a function of normalized steam position See Enclosure i response to Section 7, System Valves.

7-4

1 i

t 0

ENCLOSURE 1 RESPONSE TO SECTION 7. SYSTEM VALVES p

7-5 l

[

The level of detail requested in Section 7, System Valves was found, after discussions with Argonne National Laboratory, to be unnecessary for their analysis. The information required about system valves will be presented by valve in the following order:

1) Main Feedwater System Valves

2) Main Steam Safety Valves

3) Main Steam Isolation Valves

4) Main Steam Atmospheric Dump Valves

5) Turbine Admission Valves

6) Turbine Bypass V31ves 7-6

w MAIN FEEDWATER SYSTEM VALVES Redundant Feedwater System Isolation Valving shall be provided in both the economizer feedwater lines and the downcomer feedwater lines such that the following criteria are met when the effects of single failure criteria are imposed:

1 Complete termination of forward feedwater flow is assumed within 5 seconds after receipt of an MSIS.

l Abrupt complete termination of reverse feedwater flow with the existence of a reverse flow condition.

(Check valves are considered to be an accept-able means of achieving the above).

l The total reverse leak rate of feedwater check valves from each steam generator i

shall not exceed 1000 cc/hr.

l The Economizer and Downcomer Feedwater Line Isolation valves in each main feedwater line shall be remote-operated and be capable of maintaining tight shutoff under the main feedwater line pressure, temperature and flow re-sulting from the transient conditions associated with a pipe break in either direction of the valves.

There will be a maximum of 500 cubic feet in the piping between the feedwater isolation valves and the associated steam generator.

~

7-7

MAIN STEAM SYSTEM SAFETY VALVES Number of Valves

• 20 l

Set Pressure

• 4 valves 1270 psia 4 valves 1305 psia 12 valves 1333 psia Total Capacity (minimur.1, all valves) 19.0 X 106 lb/hr. at max, set -

Accumulation 3%

pressurG Blowdown 5%

Maximum Capacity of any one valve 1.9 X 106 lbo/hr. a t _1000 Di@

l

• C-E's recommendations.

i 7-8 4

MAIN STEAM ISOLATION VALVES (MSIV) l The MSIV leak flow shall not exceed 0.001 percent of nominal flow at 1270 psia in the forward direction and shall not exceed 0.1 percent of nominal flow at 1270 psia in the reverse direction.

The full open to close stroke time of the MSIV's shall be 5 seconds or less upon receipt of a Main Steam Isolation Signal (MSIS).

The MSIV shall be either a fail close valve or a valve guaranteed to close upon receipt of a Main Steam Isolation Signal.

The MSIV Bypass Valve shall be a fail close, power operated valve.

No single MSIV Bypass Valve or MSIV Byppss Line shall have a capacity 6

greater than 1.9 X 10 lb/hr of saturated steam at 1000 psia.

No automatically actuated valves shall be located upstream of the MSIV's except as required for steam driven emergency feedwater pumps. The max-imtsn allowable flow rate per line is 1.9 X 106 lb/hr at 1000 psia.

There will be a maximum of 2000 cubic feet between one steam generator and its associated MSIV's.

(This includes both steam lines).

7-9

MAIN STEAM ATMOSPHERIC DUMP VALVES Remote-operated Atmospheric Dump Valves shall be provided in each of the four main steam lines to allow cooldown of the steam generators when the i

Main Steam Line Isolation Valves are closed, or when the main condenser is not available as a heat sink.

Each valve shall be capable of holding the plant at hot standby dissipating core decay and Reactor Coolant Pump heat, and allowing controlled cooldown from hot standby to Shutdown Cooling System initiation conditions. Each valve shall be sized to allow a controlled plant cooldown in the event of d line break or tube rupture, which renders one steam generator unavailable for heat removal, concurrent with a loss of nonnal AC power and single active failure of one of the remaining two Atmospheric Dump Valves.

To accomplish the above, each Atmospheric Dump Valve shall have sufficient capacity to meet the saturated steam flow conditions shown in Figure 7.1. An Atmespheric Dtanp Valve with a satursted steam capacity of not less than 950,000 lb/hr at 1000 psia (critical flow assumed) will satisfy the steam flow requirements over the range of ir.let pressures shown in Figurg 7.1. Also no single valve shall have a maximum capacity greater than 1.9 X 100 lb/hr at 1000 psia.

The Atmospheric Dump Valves shall be failed closed and shall be capable of being remote manually positioned to control the plant cooldown rate.

7-10 l

TURBINE ADMISSION VALVES i

The Turbine Admission Valve shall be sized to limit turbine power to no more than 4256 Mwt.

The maximum delay from the initiation of a reactor trip signal until steam flow to the turbine and moisture separator reheater is terminated 'e 3.5 sec.

I A typical

• value for the pressure drop between the steam generator and the turbine admission valve is 34 psi at 17.18 X 106 lb/hr and 1070 psia.

The main steam lines shall be headered together prior to the turbine stop valves, but not upstream of the MSIV's and, a cross connect line shall-be provided which will maintain steam generator pressure differences within the stated limits for all normal and upset conditions:

1) 0-15% pcwer operation, pressure difference to bel psi.

2) 15%-100% power operation, pressure difference to be 3 psi.

A typical

• value for the diameter of the cross-connect line is 34".

There will be a maximum of 14,000 cubic feet between the MSIV's and the turbine stop valves. (Includes all steam lines).

• The values presented as typical are plant specific, and may vary.

l e

7-11

i TURBINE BYPASS VALVES The total capacity of the turbine bypass valves is 55% of full steam flow. (1070 psia.$

The maximum capacity of any one valve is 1.9 X 106 lbm/hr at 1000 psia.

The Turbine Bypass Valve operating speeds shall be as follows:

The valves shall stroke from the full closed position to the full open position and from full open position to full closed position in 15 to 20 seconds when a modulation signal is applied to the valve control systea.

The valves shall stroke from the full closed position to the full open position in less than i second when a quick opening signal is applied to the valve control system.

The valves shall stroke from the full open position to the full closed position within 5 seconds when the permissive gating signal is removed from the valve control system.

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~200 400 600 800 1000 1200 1400 INLET PRESSURE (PSIA)

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SECTION 8.0 FUIL F.C3 CI !" :

VALUE I.

Fuel Pellet Data A.

Composition Table 8.1 5.

Enrich =ent(s)

Table 8.1 C.. Cold state tem;:erature for fuel dimensions Table 8.1 D.

Density Table 8.1 E.

Fuel pellet heignt Table 8.1 F.

Diameter Table 8.1

'G.

Pellet dish spnerical radius Table 8.1 H.

Pellet dish depth Table 8.1 I.

Pellet dish diameter Table 8.1 J.

Burnup at end of each cycle Table 8.1 K.

Fuel sinteririg'le:perature Table 8.1 L.

0/M ratio Table 8.1 M.

Fuel surface roughness Table 8.1 N.

Radial pcwer distribution across pellet Figure 8.1 such that 2

2 N

P, (r 3 - r ) = 1 n

n

.E.i r2 f

where

~

radius to outside of fuel pellet

=

rf th r,

, inner radial ccordinate of n mesh spacing

=

outer radial ccordinate of n " mesh spacing

=

rpj power profile f acter fer n " mesh spacing t

P,

=

N

=,'n' umber of mesa spacings in fuel 8-1

II. Fuel ;;: Octa VALUE Table 8.2 5.

Fuel sta:t insulating pellets Table 8.2 1.

ce= position 2.

length a.

top pellet Table 8.2 b.

bottcm pellet Table 8.2 C.

Upper plenum volume including spring Table 8.2 D.

Plenum spring 1.

c:mposition Table 8.2 2.

number of coils Table 8.2 3.

unccmpressed height Table 8 a 4.

uncompressed outer diameter Table 8.2 5.

spring wire diameter Table 8.2 E.

Fill gas cc.positten Table 8.2 F.

Fill gas pressure at cold state Table 8.2 G.

Fill gas temperature at cold state Table 8.2 H.

Fuel rod cladding 1.

ccmpositien Table 8.2 2.

inside diameter Table 8.2 3.

outside diameter Table 8.2 4.

fuel r,cd length Th 82 5.

arithmetic.ean rougde.ess Table 8.2 1.

Axially iveraged and time averaged fast neutron fiux c1 adding exposed to during lifetime.

Fast neutron 1 ewer threshold is 1 MeV.

a e 8.2 J.

Axially averaged and ti=e averaged

. thermal neutren flux cladding exposed to during lifetime Table 8.2 K.

Time span of cladding neutr:n exposure Table 8.2 L.

Fuel red pit:n Table 8.2

,.. ~

8-2 1

val.UC III. Fuel Rod / Assembly Thema1 hydraulic Data A.

Hydraulic diametar, nc.:inal enanns1 Table 1.5 S.

Rod average linear nea: ra a Table 1.5 C.

Peat to average nea flux fact:rs as a function of axial elevation.

N/A D.

Hot, channel and het spo': p a'ra..e ers 1.

maximum heat flux Table 8.3 I.

maximum line.ar heat rate Table 8.3 3.

fuel maxi =um temperat::re '

Table 8.3 4.

cladding maximum tam:erature Table a-3 Table 8.3 5.

hot channel outlet ta=: erat::re Table 8.3 6.

hot channel outlet enthalpy

• /.

DNS ratio ('J-3 cerrelatien),

Steady state Table 8.3 9

O O

e e

e e

9 e

4 4

e d

e e

e 9

e a8' e

e

/*

8-3 e

e e

l Table 8.'

Fuel Pellet Data 3

Composition - Total Uranium Carbon Nitrogen Flourine Chlorine and Flourine Iron Nickel Thorium 1.92

• Enrichment (weight % of U235)

Cold State Temperature for Fuel Dimensions 68'F Density 92.13%

Fuel Pellet Height 0.390 inch Diameter 0.3250 inch 3

Pellet Dish Spherical Radius Pellet Dish Depth 0.017 inch Pellet Dish Diameter 0.125 inch Fuel Sintering Temperature 1700*C Minimum 0/M Ratio 2.00 to 2.02 Surface Roughness - u - in Pfis 2

Pellet

• This value for enrichmen't is the minimum value, which is conservative for fuel performance analyses.

(Other values are present in fuel loading).

8-4 J

Table 8.1 (continued)

Pin Burnups The maximum 1-pin rod averaged burnup at EOCl is 22721 MWD /T.

The maximum local burnup is 25610 MWD /T + 10%.

NOTE:

EOC (End of Cycle) refers to fuel cycle O

a 8-5

Table 8.2 Fuel Rod Data Fuel Stack Height 150.0 inches Fuel Stack Insulating Pellet A.

Composition Al 0 23 B.

Length a.

Top Pellet 0.250 inch b.

Bottom Pellet 0.250 inch Upper Plenum Volume Including Spring i

Plenum Spring A.

Canposition 302 Stainless Steel B.

Nunber of Coils

~3 C.

Uncompressed Height D.

Uncompressed Diameter E.

Spring Wire Diameter i

Fill Gas Composition a.t,.774 Nt/Mtu Fill Gu Temperature at Cold State

,_68'F Fill Gas PmssurUpsia) 4 t

a Fuel Rod Cladding A.

Composition Zircaloy-4 B.

Inside Diameter 0.332 inen C.

Outside Diameter 0.382 inch D.

Fuel Red Length 3

E.

RMS Average Roughness Axially averaged and time averaged fast neutron flux cladding exposed to during lifetime. Fast neutron lower thereshold is 1 MeV.

8-6 s

Table 8.2 (Continued)

Axially averaged and time averaged thennal neutron

~

3

~

flux cladding exposed to during lifetime

~

Time Span of Cladding Neutron Exposure 24,912 EFPH Fuel Rod Pitch 0.506 inch 0

e e

0 e

8-7

t Table 8.3 Steady State Hot Channel and Hot Spot Parameters Used in LOCA Analyses (1.2) 6 2

Maximum heat flux (10 BTU /hr-ft) 0.465 Maximum linear heat rate (KW/ft) 14.0 Fuel maximum temperature a) average 2175'F b) centerline 3425*F Hat channel outlet temperature 648*F Hot channel outlet enthalpy 692.9 BTU /lb DNB ratio Not applicable to LOCA Cladding maximum temperature 710.5'F

1) The information requested is not available for steam and feed line break analyses. However, this information can be derived from the data presented in Section 1.0.

2)

CESSAR-F Anclysis as performed by C-E.

8-8

Flaune 8,1 RADIAL POWER DISTRIBUTION ACROSS FUEL PELLET 4,5

~

e M

~

8-9

SECTION 9.0 Containment Data The following data has been included in this package in order to aid modeling of the containment during a Large Break LOCA*.

The C-E representation of containment is set up as a Heat Sink model. Each Heat Sink has an area, thickness and thermal properties. These parameters can be found on Table 9.1.

Additional information is inneluded on Table 9.2 which contains the mass and energy release data.

• CESSAR-F Analysis performed by C-E.

t 9-1

TABLE 9.1 CONTA.II"G:7 N!YSICAL PAPMETERS 6

3 Net Free Volume

• 3.7 x 10 fg Initiation Time for Spray Flow 0.0 see

- Containment Initial Conditions:

Temperature 50*F Pressure 14.7 psia Relative Humidity 100%

Enclosure Building Temperature 38'F' Containment Spray Water:

Temperature 60'F Flow Rate 11,000 gpm Heat Transfer Coefficient from Heat S

i 2

Sinks to Building Annulus 13.0 BTU /hr-ft,p l

Heat Sink Physical Data:

j n

0.375 in. Thickness Internal Steel 375.000ftj 1.65 in. Thickness Steel Liner 130,000fg

(

1.0 ft Thickness Internal Concrete 14,608 ft 2 1.5 ft Thickness Internal Concrete 180,933fj 3.6 ft Thickness Internal Concrete 36,589 ft f

9.5 ft Thickness Internal Concrete with 2

0.25 in. Steel Liner 18,219 ft Thermal Conductivity of:

Steel 26.0 BTU /hr-ft -F Concrete 1.0 BTU /hr-ft.-F Volumet'ric Heat Capacity of:

Steel 3

56.35 BTU /fg,7 Concrete 32.4 BTV/ft -F 9-2

-',IABLE 9.2 BLOWDOWN AND REFLOOD MASS AND ENERGY RELEASE FOR A LARGE; BREAK'LOCA* IN PUMP' DISCHARGE LEG-Integral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release (sec)

(1bn/sec)

(BTU /SEC)

(1bm)

(BTU) 0

0. 0 0.0 0.0 0.0 4

7 3

6

.05 7.097 x 10 3.9747 x 10 3.6574 x 10 2.0497 x 10 3

.10 8.2847 4.6555 7.3531 x 10 4.1238 4

.15 8.2234 4.6171 1.1464 x 10 6.4326 6

.20 8.0164 4.5023 1.5532 8.7170 x 10 7

.25 8.0621 4.5320 1.9547 1.0973 x 10

.35 7.8749 4.4326 2.7470 1.5430

.45 7.7544 4.3682 3.5251 1.9811

.60 7.6083 4.2885 4.6751 2.6291

.80 7.5353 4.2496 6.1934 3.4852 4

1.0 7.4282 4.1912 7.6878 x 10 4.3282 5

1.4 7.1519 4.0418 1.0608 x 10 5.9769 1.8 6.8132 3.8600 1.3394 7.5534 7

2.2 6.1414 3.4868 1.6014 9.0393 x 108 2.6 5.3776 3.0602 1.8299 1.0338 x 10 3.0 5.0588 2.8906 2.0381 1.1525 3.4 4.8416 2.7816 2.2356 1.2657 3.8 4.6863

~ 2.7169 2.4267 1.3759 4.4 4.2674 2.5319 2.6958 1.5335

5. 2 3.5744 2.2329 3.0080 1.7234 6.0 3.1232 2.0305 3.2743 1.8933 6.8 2.8498 1.8713 3.5123 2.0492

7. 6 2.6988 1.7547 3.7331 2.1937 8.4 2.6244 1.6879 3.9465 2.3315 9.2 2.5071 1.6090 4.1518 2.4633 10.0 2.3875 1.5354 4.3476 2.5891 11.0 2.2127 1.4380 4.5779 2.7378 12.0 1.9730 1.3389 4.7884 2.8767 13.0 1.4056 1.1824 4.9597 3.0029 4

7 5

8 14.0 1.0867 x 10 1.0224 x 10 5.0800 x 10 3.1132 x 10 h

9-3

TABLE 9.2 { Continued)

BLO'iD?iN AND REFLOOD MASS AND ENERGY RELEASE FOR A LARGE BREAK;LOCA*;IM PIM DISCHARGE' LEG-Jntegral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release (sec)

(ibm /sec)'

(BTU /SEC)

(1bm)

(BTU) 3 6

5 8

15.0 9.7275 x 10 9.1101 x 10 5.1854 x 10 3.2098 x 10 16.0 7.4384 7.8313 5.2711 3.2949 17.0 5.8326 5.7784 5.3367 3.3622 18.0 5.2488 5.0939 5.3897 3.4148 19.0 6.7388 5.4679 5.4501 3.4685 20.0 6.8805 4.8114 5.5198 3.5209 21.0 6.3304 3.9007 5.5860 3.5644 22.0 5.2770 3.0069 5.6442 3.5988 23.0 4.3468 2.3716 5.6922 3.6255 24.0 4.1383 2.1111 5.7324 3.6470 3.6651 25.0 2.4047 1.3428 5.7674 5

8 3

6 25.2 2.4476 x 10 1.3378 x 10 5.7738 x 10 3.6683 x 10 Time of Annulus Downflow Start of Reficed (values below are for 5tean only) 5 3.6633 x 108 31.9 0

0 5.7738 x 10 41.9 0

0 5.7738 3.6683 51.9 0

0 5.7738 3.6683 61.9 0

0 5.7739 3.6683 2

5 71.9 1.9177 x 10 2.5030 x 10 5.7831 3.6804 81.9 1.7378 2.2681 5.8012 3.7041 91.9 2.1002 2.7412 5.8204 3.7291 101.9 2.0898 2.7277 5.8412 3.7563 111.9 2.0705 2.7024 5.8620 3.7835 121.9 2.0652 2.6655 5.8827 3.8105 131.9 2.0652 2.6955 5.9033 3.8374 141.9 2.0665 2.6972 5.9240 3.8643 151.9 2.0675 2.6984 5.9446 3.8913 161.9 2.0728 2.7054 5.9654 3.9183 171.9 2.0858 2.7224 5.9861 3.9454 181.9 2.0935 2.7324 6.0070 3.9727 191.9 2.0972 2.7373 6.0280 4.0000 201.9 2.0999 2.7408 6.0490 4.0274 2

5 5

8 221.9 2.1168 x 10 2.7629 x 10 6.0911 x 10 4.0825 x 10 e

9-4

TTCLE 9,2 (Continued) 210WDOWN AND REFLOOD MASS AND ENERGY RELEASE F0R A LARGE BREAK-LOCA* IN PlMP DISCHARGE LEG-Int'egral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release (sec)

(1bm/sec)

(BTU /SEC)

(lbm)

(BTU) 2 5

5 0

~

241.9 2.1389 x 10 2.7917 x 10 6.1336 x 10 4.1379 x 10 261.~9 2.1375 2.7898 6.1764 4.1937 281.9 2.1595 2.8185 6.2194 4.2499 301.9 2.1622 2.8222 6.2626 4.3063 321.9 2.1823 2.8483 6.3061 4.3631 341.9 2.1904 2.8589 6.3497 4.4200 361.9 2.1895 2.8577 6.3935 4.4771 381.9 2.1984 2.8694 6.4374 4.5344 401.9 2.1942 2.8639 6.4814 4.5918 421.9 2.1976 2.8683 6.5254 4.6493 441.9 2.2052 2.8782 6.5695 4.7068 461.9 2.1994 2.8706 6.6136 4.7643 481.9 2.1994 2.8707 6.6576 4.8219 501.9 2.2050 2.8780 6.7017 4.8794 511.9 2.2075 2.8913 5.7557 4.9369 541.9 2.2052 2.8783 6.7898 4.9943 561.9 2.1971 2.8677 6.8338 5.0518 551.9 2.1944 2.8642 6.8777 5.1091 601.9 2.1947 2.8645 6.9216 5.1654 2

5 0

8 631.9 2.1852 x 10 2.8521 x 10 6.9872 x 10 5.2520 x 10 d

e 9-5 e

A

COMBUSTION ENGINEERING, INC.

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