ML20054M221

From kanterella
Jump to navigation Jump to search
Responds to Requesting High Pressure Injection Nozzle Info for Review of Util Response to Aslab 820415 Memorandum & Order
ML20054M221
Person / Time
Site: Rancho Seco
Issue date: 07/02/1982
From: Mattimoe J
SACRAMENTO MUNICIPAL UTILITY DISTRICT
To: Eisenhut D
Office of Nuclear Reactor Regulation
Shared Package
ML20054M220 List:
References
NUDOCS 8207120112
Download: ML20054M221 (4)
v • d • e


Text

- - - - - - -

v. ;.

.'C,SMUD agy?

SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacramento, California 9M13; (916) 452-3211

'82 J1 -9 E3 O~

July 2, 1982 CFFCI Cf MCPKI ccu p A uv h[\\h is b d DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION DARRELL G EISENHUT DIRECTOR DIVISION OF LICENSING U S NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT 1 HPI N0ZZLE INFORMATION Your letter of May 21, 1982, requested information for your review of our response to the ASLAB Memorandum and Order dated April 15, 1982.

The District's responses to your questions follow:

QUESTION I:

Provide the bases for your conclusion that "no plausible means have been identified for the thermal sleeve to cause thermal-hydraulic damage to the core from flow blockage".

Consider the possibility of a damaged thermal sleeve configuration (e.g.,

longitudinally split and flattened) that can enter the reactor vessel plenum between the flow distributor and the lower grid plate.

RESPONSE

The assessment of the thermal sleeve's potential to cause thermal-hydraulic damage to the core from flow blockage included considera-tion of both an intact and a damaged sleeve geometry. Thermal-hydraulic damage in this context is defined as a significant reduc-tion in DNBR margin.

The assessment of an intact sleeve concluded that the sleeve is too large to pass through the lower end fitting grillage.

Thus an intact sleeve could not reach the active fuel region of the core, but could possibly lodge below and at an angle to the lower end fitting grillage.

This condition could cause a slight flow maldistribution away from the fuel assembly below which the sleeve is lodged.

Thermal-hydraulic licensing analyses assume a non-mechanistic 5% flow maldistribution away from the hot fuel assembly as a normal procedure. Thus a slight flow maldistribution is already specifically accounted for in kancho Seco's licensing basis.

In addition, analyses performed with crossflow codes for similar inlet flow blockages have shown that inlet flow perturbations have a negligible effect on DNBR downstream.

8207120112 820708 PDR ADOCK 05000312 C

PDR

- ir m

DARRELL G EISENHUT July 2, 1982 The assessment of a damaged sleeve geometry considered the potential of pieces of various sizes to lodge in inconvenient places. A longitudinally split and flattened piece of a sleeve that lodges against the lower end fitting grillage could cause a similar but potentially more severe flow maldistribution than an intact sleeve.

The result, however, is the same:

inlet flow perturbations do not significantly affect downstream DNBR's. A piece that is small enough to pass through the lower end fitting grillage and lower end spacer grid, and thus gain access to the active fuel region of the core, is also too small to produce a significant flow block-age by itself should it lodge in an intermediate spacer grid.

It would take many small pieces all lodging in the same subchannel at the same intermediate spacer grid to produce any significant flow blockage.

Thermal-hydraulic studies have shown that the flow blockage /DNBR reduction effect is very localized and does not propagate downstream significantly.

Thus the probability of a flow block' age occurring in the core of sufficient magnitude and in the appropriate location to signigicantly reduce core thermal margin is considered negligibly small.

We conclude, therefore, that based upon the results of thermal-hydraulic studies on the effects of flow blockages, no plausible means has been identified for the thermal sleeve to cause thermal-hydraulic damage to the SMUD core from flow blockage.

QUESTION II:

If there is a plausible means of flow blockage from your considera-tions of Question I, then describe the bases for concluding that the Loose Parts Monitor installed at Rancho Seco will be capable of detecting movement of the loose thermal sleeve.

RESPONSE

As indicated in the response to Question I, a possible means of flow blockage has not been identified; however, our response dated April 21,1982, to the ASLAB's Question I discusses in detail the location of loose parts monitor sensors and system set points which show its capability of detecting movement of a loose thermal sleeve.

QUESTION III: The thermal sleeve in nozzle A has been missing for an undetermined amount of time (up to six years of plant operation).

In this situ-ation thermal fatigue cracking may have occurred on the inside sur-face of the RCS cold leg piping downstream of the make-up line connection.

An illustration of such cracking in feedwater lines at I

the Barseback unit are different than the conditions at Rancho Seco, we request that you provide us with your plans for augmenting your ISI program to include non-destructive examinations of these areas on a periodic basis.

l l

DARRELL G EISENHUT July 2, 1982

RESPONSE

Thermal fatigue cracking has not occurred on the inside surface of the RCS cold leg piping downstream of tt'e make-up line connection. This was determined during the repair of the thermal sleeve and safe end on nozzle A.

This area is, of course, inaccessible for visual examination in the fature without again cutting and removing the make-up line from the nozzle. The B&W Owners Group Task Force is investigating the feasibility of developing ultrasonic examination techniques for the inspection of this area; however, this is not felt to be an item of safety concern since reinstallation of a thermal sleeve in this nozzle should prevent this form of cracking.

QUESTION IV:

In your reply to the ASLAB's Question 4, you indicate that new thermal sleeves are installed in nozzles A and B, but no replace-ment of C&D thermal sleeves is necessary.

RESPONSE

Radiographic examination is required to determine if the thermal sleeves have become loose.

Since no change has been made to the sleeves in the C & D nozzles, the District does commit to a radiographic examination of these sleeves at the next refuelling outage.

Since the sleeves in nozzles A and B are now positively held in place, no additional examination of these sleeves is con-sidered necessary. We will, of course, continue our normal in-service inspection program using ultrasonic examination.

QUESTION V:

Provide an evaluation of the number of thermal cycles imposed upon the make-up nozzle from all sources, against the design number of allowable thermal cycles. The evaluation should include thermal cycles from surveillance testing and plant startup. We understand that during a normal plant startup, and during surveillance testing, valve SFV 23604 (see Enclosure 2) is cycled numerous times since valve SIM 034 leaks. This closing and opening of valve SFV 23604 could cause nozzle A to be repeatedly heated up by the RCS and then quickly cooled by make-up flow, causing additional thermal cycling of nozzle A.

k

DARRELL G EISENHUT July 2, 1982

RESPONSE

Nozzle usage at Rancho Seco Unit 1 will continue to be tabulated as it has been in the past. The effect of valve cycling on the make-up nozzle cannot be accurately determined since the amount of cycling is not known and it is not known when the sleeve was lost. This effect should not have any safety significance however, because the nozzle was examined during the recent repair, with no defects found, and future inspections will detect any degra-dation, should it occur, hLt.I s

John J. Mattimoe Assistant General Manager and Chief Engineer cc: 'T. A. Baxter D. Holt Sworn to me and subscribed before me this day of July, 1982.

................................e omclAL SEAL l

PATP! CIA il GEISLER nown:c coor::a tiotary Putf1ic 5

N"~5?$$YY My Commrssion Experes Nowmber 22.1983

.e...............................e 1

e

OSMUD y

^

SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s Street, Box 158%gto, California 95813; (916) 452-3211 UShRC July 2,1982 M -9 AiDG7 0FF.CE OF S n y 7....

GCCUc:a & ;g,;,yg.

BRAkn DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 U S NUCLEAR REGULATOR COMMISSION WASHINGTON DC 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT 1 HPI N0ZZLE ANALYSIS INFORMATION Your letter of February 3,1982, requested additional information relative to our response to the ASLAB on the issue of HPI Nozzles cycles.

In particular, you requested more detail than was presented in our submittals of the Babcock & Wilcox Company Field Change Package, "HPI Nozzle", Document No. 04-3370-00 FC-0174-00 and Babcock & Wilcox Calculation f:o.'s 32-1121811-00, "HPI Nozzle Usage Factor" and 32-1119809-01, "HPI Nozzle Usage Factor". Attached ::

..,./ of Babcock & Wilcox Document No. 32-1134218-00, "SMUD HPI Nozzle Usage Factor". This document supersedes Babcock & Wilcox Document 32-1121811-00 previously transmitted.

The following answers are provided as sunnary responses to each of your questions relative to the superseded document.

1.

Update pressure and thermal stresses based on 1750 psig and 760 gpm.

RESPONSE: The calculations in B&W Document 32-1134218-00 were performed utilizing a pressure of 1600 psig and a total HPI flow rate of 1340 gpm (335 gpm per nozzle) - see Section 4.0 of the attached.

For the transient of interest, manual actuation of HPI following reactor trip, the 1600 psig/1340 gpm pressure / temperature assumption is acceptable because it results in higher transient stresses relative to the 1750 psig/760 gpm condition specifically requested.

The combination of 1600 psig and 1340 gpm was used by B&W because j

it was determined to be generically applicable and bounding for this type of transient.

2.

Specify code equations used and provide sample calculations for each category of stress and stresses in combination.

RESPONSE

The current analysis to determine the acceptability of the additional fatigue usage factor resulting from seventy manual actuations of HPI following reactor trip utilizes the thermal / mechanical stresses calculated in the original 1

.x g 1 TC A% itECTRic $Y$TEV SE8VING MORE THA1 600 003 IN T*E mea 8

  • CF A i!FGe% ;

John F. Stolz July 2, 1982 Rancho Seco design stress report.

The original work was performed in accordance with USAS B31.7, " Nuclear Power Piping Code", Draft 1968.

In general, however, the original stress report presented summary results developed by versions of computer codes which are not now utilized or retrievable. Therefore, sample calculations for the original analysis are not provided.

Section 3.0 of the attached describes the methods used in the design base.

Sections 5.0 through 11.0 of the attached detail how the design base data was utilized and the additional fatigue usage factor for the new transient calcu-lated to determine a new total HPI nozzle usage factor.

3.

Provide detailed tabulations of resultant stresses under various HPI transients.

RESPONSE

See sections 8.0 and 9.0 of the attached, in particular pages 24 through 27, 31, and 36 through 40.

4.

Attach fatigue curves used.

RESPONSE: The fatigue curves used are from USAS B31.7 and are referenced in the attached but not reproduced therein.

Specifically Figure F-106(a) of B31.7 was used for all carbon steel sections of the HPI nozzle, and Figure F-106(b) was used for all stainless steel sections.

This information relates to techniques used in the original Rancho Seco Stress Analysis, and we feel is more than adequate to address the Board's concern over high pressure injection nozzle usage.

If you feel additional information is necessary, we would propose a technical meeting since the computer codes used in the original stress analysis are not retrievable.

b John. Mattimoe Assistant General Manager i

and Chief Engineer I

l t

Attachment cc: / Tom Baxter Dave Holt i

~

l

BWNP-20210-1 (1-78)

CALCULATION DATA / TRANSMITTAL SHEET 1134218 00 CALC.

32 DOCUME?TT IDENTIFIER TRANS. 86 TYPE:

_tzstanca a om _sArm musts men _m. sm. um _essten mm. letsta mir.

_mn TITLE SMUD - HPI Nozzle Usage Factor f R. W : u-J f%.e. % -i_o_+ m PREPARED BY R. R. Schaefer/Y. C. Hamilton REVIEWED BY D. E. Killian m.

TITLE Engineer IV/ Engineer I DATE 5/26/82 TITLE Supervisory Engineer '/

6/ZM 8 3 -

I DATE PURPOSEt The purpose of this analysis is to justify, by analysis, the operational events for the HPI nozzle. The operational events include 40 test transient cycles, 240 heatup and cooldown cycles, 40 rapid depressurization cycles, 650 OBE cycles, and 70 additional HPI manual actuation cycles following a reactor trip.

I

)

SUMMARY

OF RESULTS (INCLUDE DOC. ID'S OF PREVIOUS TRANSMITTALS & SOURCE j

(

PACKAGES FOR THIS TRANSMITTAL)

]

t The HPI nozzle can withstand the 40 cycles of test transient, 240 cycles of heatup and cooldown transient, 40 rapid depressurization cycles, 650 OBE cycles and 70 additional HPI manual actuation cycles following a reactor trip.

Total Usage Factor = U = 0.80 Source Calculations 1.

SMUD Primary Piping Stress Reoort, Revision 1.

2.

FCA No. 04-3370-00.

3.

CDT 86-1131765-00, "HPI Nozzle Maximum Flow Rate".

4.

CDT 86-1131770-00, " Revision to Functional Spec. -SMUD".

DISTRIBUTION See DRN.

of GE Page y

A Babcock &Wilcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS e

TABLE OF CONTENTS Page 1.0 Introduction 3

2.0 Conclusion and Results 4

3.0 Discussion and Method of Analysis 6

4.0 Des'cription of Transients 13 5.0 Thermal Parameters 16 6.0 Thermal Discontinuity 18 7.0 HPI Nozzle Loads and Stresses 19 8.0 Primary + Secondary Stress Intensities 23 8.1 40 Test, 240 Heat-up and Cool-down, and 40 Rapid Depressurization Transient Cycles 28 8.2 70 Cycles of HPI Actuation and OBE Cycles 30 9.0 Peak Stress Intensities 36 10.0 Fatigue Analysis 41

(

10.1 40 Test, 240 Heat-up and Cool-down, and 40 Rapid Depressurization Transient Cycles 42 10.2 70 Cycles of HPI Actuation and OBE Cycles 51 11.0 Total Usage Factor 62 12.0 References 64 Figure 1 - Nozzle Thermal Grid 10 Figure 2 - Themal Sleeve Geometry 11 l

Figure 3 - Nozzle Axisymmetric Model 12 l

l Figure 4 - Temperature Transient For Test and Rapid Depressurization 14 I

Figure 5 - Tempera;.ure Transient for HPI Manual Actuation 15 Figure 6 - Primary + Secondary Stress Intensities 27 Figure 7 - Peak Stress Intensities 40 Table 1 - Primary + Secondary Stress Intensities at Juncture #8 24 j

(L-R Intensity)

Table 2 - Peak Stress Intensities at Juncture #8 (L-R Intensity) 37 PREPAaED gy k, b.

Ah bI2b Ib2 O A TE DOC.NO E I

  • M,, )

T 2{a.

1

'2 O F G 5 sewEwto av PAGE NO DATE mm.# rg

'e Babcock &Wilcox 32-1134218-00 nuciear pow 7rEe7ation oivision GENERAL CALCULATIONS

1.0 INTRODUCTION

Following each reactor trip, additional make-up flow must be obtained to prevent loss of indicated ~ pressurizer level during the transient. All four of the high pressure injection (HPI) nozzles, located in the cold legs on the discharge side of the pumps, have been used for these occurrences. The three nozzles which do not have continuous make-up flow to the system receive a thermal shock from the cold BWST (borated water storage tank) water. The operational events for the HPI nozzle analysis included 40 test transient cycles, 40 rapid depressurization transient cycles, 240 heatup and cooldown cycles and 650 OBE cycles. An additional 70 cycles of HPI manual actuation following a reactor trip is being added, Ref. #1.

The purpose of this report is to justify, by analysis, the operational events for the HPI nozzle following a reactor trip.

The analysis method utilized will be the simplified

(

elastic-plastic discontinuity analysis in Ref. #2.

1

  • =eraneo sv 02. c_ h 0 h oari 5 / 2 c, / M ooc. ~o.

bb >lL oarn f/.2.4,h2-A a = G5 envaewto si raon no.

I

Babcock &Wilcox 32-1134218-00 Nuclear Pow r e tion Division GENERAL CALCULATIONS 2.0 Conclusion and Results Primary + Secondary Stress Intensity Range 40 Cycles of Rapid Depressurization Transient Primary + Secondary = 115.1 KS1 > 3Sm = 51.3 KS1 Range 40 Cycles of Test Transient Primary + Secondary = 89.84 KS1 > 35m = 51.3 KS1 Range 240 Cycles of Heat-up and Cool-down Transient Primary + Secondary = 11.04 KS1 < 3Sm = 51.3 KS1 Range 30 Cycles of HPI Actuation With Inclusion of : OBE Stresses Primary + Secondary

= 143.56 K31 >35m = 51.3 KS1 Range 40 Cycles of HPI Actuation Without Inclusion of e OBE Stresses

(

Primary + Secondary s

7

= 119.74 KS1 > 3Sm = $1.3 KS1 Range 650 Cycles of 1 0BE Stresses Primary + Secondary

= 23.82 G1 < 3h = 51.3KS1 Range Total number of cycles in which the primary + secondary stress intensity range exceeded 3Sm is 40+40+30+a0 = 150 < 250.

Therefore, an elastic plastic fatigue analysis was performed.

/

PeePAnto av O.J. 0_. h oAre j i 2 G / O2.,_

coc. wo.

[2I &, J oire I, h, M PAGe NO.

Ob b5 eeviewto av

4 1

Babcock &Wilcox 3 2 - 1 1 9 4 9.18 - 0 0 nucieer po 7,Te% tion oivision GENERAL CALCULATIONS USAGE FACTOR Usage Factor For 40 Rapid Depressurization U

=

1 Transient Cycles = 0.16 Usage Factor For 40 Test Transient Cycles = 0.04 U

=

2 Usage Factor For 240-40 = 200 Heat-Up and Cool-Down U

=

3 Transient Cycles = 0.0 Usage Factor For 30 HPI Manual Actuation Cycles U

=

4 (With Inclusion of + OBE Stresses) = 0.43 Usage Factor For 40 HPI Manual Actuation Cycles U

=

S (Without Inclusion of + OBE Stresses) = 0.17 Usage Factor For Remainder of + OBE Stress Cycles = 0.0 U

=

6 U = Total Usage Factor

(

U=

V) + U2+U3+U4+US+V6 U = 0.16 + 0.04 + 0.0 + 0.43 + 0.17 + 0.0 V = 0.80 < 1 In conclusion, the HPI nozzle can withstand the operational events of 40 rapid depressurization, 40 test, 240 heat-up and cool-down, 650 OBE and 70 HPI actuation transient cycles.

l l

/tt. t. Wh oire 5 I 2 C, / 9 2 occ. wo.

retraneo av d 8de M /82 -

5 OF GE oire.

Pace no.

neviewee av l

/

/

/

1

Babcock &Wilcox 32-1194218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 3.0 Discussion and Method of Analysis This analysis calculates the additional fatigue usage factor on the HPI/ Makeup Nozzles resulting from seventy additional HPI actuations following a reactor trip. The method of analysis utilizes the themal/ mechanical stresses calculated in the original stress report. The following discussion is provided as background infomation to the methods used in the original HPI analysis. The discussion will address three topics; 1) thermal analysis, (2) structural model and 3) :: tress analysis.

Thermal Analysis A two-dimensional heat transfer analysis utilizing B&W computer code P91167, reference #3, was perfomed to obtain the temperature distribution in the nozzle and local shell region. A model of the nozzle and local shell region is shown in Figure 1.

The nozzle

{

and shell components are represented by a system of blocks with a nodal point at the center of each block.

Program P91167 solves a heat balance equation between each block and the four adjacent blocks.

The following transients were selected for thermal analysis, because these transients either contribute significantly to the usage factor or are of short duration and have larger temperature differences than other transients.

(1) Heatup and cooldown (Transient 1A or IB)

(2) Power loading and unloading (Transients 2A, 28, 3 and 4)

(3) Rapid Depressurization (Transient 9)

(4) Test transient-HPI system (Transient 22)

The heat transfer boundary conditions consist of convective heat transfer at the inside surfaces of the nozzle and shell.

The HPI nozzle contains a themal sleeve (see Figure 2). There is an enclosed water gap between the thermal sleeve and the nozzle inside surface.

The cut-off surfaces of the nozzle and shell are assumed

/

insulated. Also, the outer surfaces of the nozzle and shell are considered u

PlePAteo BY F

M DATe DOC.NO 4L.t. h h,J G oF G 5 #26 f M nevieweo 81

,4ce uo.

oare

Babcock &Wilcox 32-1194218-00

' = " '

. m e..........

Nuclear Power Generation Division GENERAL CALCULATIONS 3.0 Discussion and Method of Analysis - Continued insulated.

The results of this thermal analysis consist of temperature at each nodal point of the grid thermal model, Figure 1.

Several nodes and pairs of nodes representing critical locations of the nozzle are selected to evaluate the radial and axial thermal gradients resulting from the application of each transient condition. These thermal gradients for each selected node are plotted.

From these plots, selection of critical transient times for subsequent stress evaluation was made.

Structural Model The thermal stress calculations were performed utilizing B&W

(

Computer Programs P91206 and P91032, References #4 & #5 respectively Program P91206 uses the virtual work method to solve axisymmetric shells of revolution (see Figure 3).

Program P91032 is a general thermal motion and stress program solving for various shapes using appropriate classical theory.

The portion used in this analysis is the opening in a cylinder using flat plate theory modified i

to account for curvature in the circumferential direction.

The stresses were generated by inputting appropriate temperatures and geometry into the programs assuming no reactions at the nozzle to shell intersection. A two element discontinuity analysis was then perfomed and forces and moments generated were then superimposed l

on the thermal stresses to give a total thermal stress picture.

Stress Analysis Several computer runs were made to obtain the stresses in the nozzle resulting from the application of selected transients.

The loads for each selected critical transient time consisted of the temperature distribution, operating pressure at that transient time and the nozzle to shell interaction loads. The pressure was b

b PtePAteD BY a #

DATe DCC. NC.

s i

j IIA C. Wh 9 or G5 5 / 2 G / 9 7_

, Ace uc.

newswee av oire

BabcockgWi co3 82-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS Stress Analysis - Continued applied at the inside surfaces of the nozzle and the shell.

The resulting stresses in the elements included stress concentration effects at structural discontinuities. Stress concentration factors were obtained by using the indices of USAS B31.7, " Nuclear Power Piping",1968 Draft, and other sources.

Program P91206 also output primary plus secondary stresses. The range of these linearized stresses was then compared to the associated 35m stress limit. The peak stresses and associated usage factors were determined for the selected critical locations in the nozzle and the shell.

In the calculations following, some of the inherent conser-vatism in the original stress calculations is removed. The tabulation for primary plus secondary stresses (Table 1), external load and peak stresses (Table 2) are taken directly from the original stress report.

u The required thermal analysis for the HPI transient following a reactor trip is performed using temperature distribution program P91232, Reference #11.

This program calculates the temperature distribution through the thickness of a cylinder as a function of, time by solving one dimensional heat transfer equations. This program also determines the linear and non-linear portions of the radial temperature gradient and the associated stresses.

This program was run for various types of transients to develop a simplified " temperature gradient / stress" ratio method to determine stresses for the added reactor trip HPI actuation transient.

This simplified stress ratio method utilizes the stresses in the original design analysis for the rapid depressurization transients to obtain stresses for the added reactor trip transients.

The stresses and the associated cycles for the reactor trip transients were used in conjunctio'n with the stresses and cycles in the original design analysis for the test and rapid depressurization transients to determine a total usage factor.

[k b[

le 2--

Peeraneo sv oAre occ. No.

/

/

b OE bb bb[

PAGl NO agviewto sv

. C.-

oATE

Babcockp,3,ilco3 3 2 - 1 1.9 4 218 - 0 0 ' ~ ""

~

Nuclear Power Generation Division GENERAL CALCULATIONS Stress Analysis - Continued To conservatively analyze earthquake stresses, the following was used. There are 650 OBE cycles to consider consisting of 30 separate earthquake events, with each event having approximately 22 cycles. The full range of OBE stress was added to the thermal, pressure and thermal expansion stresses for a HPI transient following a reactor trip. This stress range was evaluated for 30 cycles. The 620 (650-30) remaining cycles of full range of OBE stress (: OBE) were analyzed separately.

M PeePAaeo sv oATe oO C. NO.

/

ievieweo sv

'h. L O

/ '2

/Q2 Pace ~O 4 0F G5 o.re

Babcockp,Wilcox 32-1134218-00

' " ~ ' ' " ' ' " "

Nuclear Power Generation Division GENERAL CALCULATIONS sir",

20 0 I

a fl l

't7 l.

I hl F/ CURE i

j j 1, n

2 nosELE T hE5LM A L 3

6 R.1D ld

/, %'t 'g l

io l

i 17f b

e i

a N

N O

k f

'% {c 6

b;

'7 3

4 7

112.

10 HL 9

4 it 3S I

I I

I l

i i

l i

O S t f'<, '31 % '5'! l'6 'A

'r,l Ir6 l 19/

l /96 19 7 t

k r

,e o Ar, S

S re.paa,o av ooc. No.

,,,,,,o,,

ru. c. d 426oe io eF c5 o.1,

....~o.

Babcock &Wi!?S,3 32-1134218-00 e

Nuclear Power Generation Division GENERAL CALCULATIONS ld 1.0625"--= 2 A

s

.ssi"9 i

' o.

saction i

/

N y

~

RoY N\\

m

\\

1. 0" i

s:

E In s

d % 6ecTsont 2

. 875 -e -.

r s

.75 +

N.x Y J

~

% } SEcriou 4 y.

SEEc,T) o N 3 C NN'M NN'NNNNN l

(d Sr6C770d f FIGulE 2

T stRMAL S L:_E.v E.

G E.O ME.TR'f I

b.

PetPAnto

,7 DAtt M>

DO C. N O.

na e. d 0 1, s,2 o. 9 2 P.O. ~0 u a c-

Babcock &Wilcox 9 2-11 'o A o 1 o hLCULATIONS

^ ^

G$A$RAl C

^

^

Nuclear Pow r e ation Division

/.2 t/ZS "

4

,e-. 5 3 S "

S Sgt

' '.. -'g

  • .43 %

> u c.

,. \\

2*G"

/ 5/55 Il 8/mi

[ 664p 4.-

If J 3'75

, s*

a

-e _

6,&

$ $$$m*

//

/. f(6 876 tjsgG eI

. 6(.25 24, M

4 SE&

a M'

N, O.

\\

E so

\\

M s ser-

,il 2.1893 33 l

FIGUFE_

3 l

NO?' ELE A %l SY rnm E T R \\ C-modi L d INm

.f/M /J 2 retrasto sv o,rg occ.so

/

/

/

oarg p.og so.

12 C F G5

$C#

5 /2G / 91 elvetwto av

BabcockgW,ilco3 82-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 4.0 Description of Transients The temperature transient for the 40 cycles of test and 40 cycles of rapid depressurization is shown in Figure 4 The test' transient 0

starts at a temperature of 550 T and a pressure of 11200 psi, drops 0

down to 60 F and ends at a temperature of 550 F and a pressure of 2200 psi. This transient lasts for 10 seconds. Th e rapid depressuri-zation transient starts at a temperature of 550 F a1d a pressure of 2200 osi, drops down to 60 F for 45 seconds, drops down to 40 F and 0

ends at a temperature of 500 F and a pressure of 600 psi. This transient lasts for 15 minutes.

The temperature transient for the 70 cycles of HPI manual actuation is shown in Figure L. The transient sta rts at a temperature of 579 F and a pressure of 1600 psi, drops down to 60 F for 45 seconds, drops down to 40 F and ends at a temperature of 558.2 F and a pressure of 1100 psi, References #8 and #9.

This transient lasts for 15 minutes.

The maximum flow rate through each nozzle is 335 gpm.

The temperature transient for heat-up and cool-down consists U

0 of heat-up from 70 F to 550 and 2200 psi and cool-down to 70 F.

This transient occurs at a temperature change rate of 100 degrees per hour.

PetraatD BY k1

(.

h Daft bl M / b DOC. NO.

PAGE NO IS O E bb M

stviewtD Sy Daft

=.

Babcock &Wilcox

~ 8 2 - 1 1 3 4 2 18 - 0 0

'" ~" ' '""

Nuclear Power Generation Division GENERAL CALCULATIONS g,

m e

og

~

B d

@2'C 2 E

c 0

W c C w'

t-9 c e O

LU 2 o' C

c-c

[

F o

C J

v w

4 2

y W L

c h

6 C b R

W o w

+

E Lc e)

=.

-D Q

r 3

a U-l I

d i

H

$5 l

c c.

_n 5

T

~

a 1E w

c a

D l

a f

1 J

C I'

c)

H Q

a

<J

=

3 w

a m

U-Yo H

S

-c l

~

5 s

i 8

2 8&

m e

~

kdo) 3 M O M O'$ 3 d W 5 1 PREP ARED BY k, C. 9h(km o [2 fc / S k goc, go, og..

(M6 /2 14 OF 65 teviewto By cart Pace so

Babcock &Wilcox 32-1134218-00

' " ' " ~ ' ' " "

Nuclear Power Generation Division GENERAL CALCULATIONS I

I C

I 2

L _ _ _ _ _ _ _ _ _ _. _. __ __ _

_ g a

r as C

1 3

F l

- I c-d 0

N l

d 2

l -U c

5 I

-=

H l

r L

v1

~

-2 Lu v1 W

H b

O

'r W

l 3

2 t.u

~

j MI

{

t F

e W

ce

? 2 j

J i v

G L

w a

N; N

W t.

A H

=

c N

2l 6

2 e

=

g f

I H

hl

]!

E

-d a.

D Cl t-c I

e w

I

_ g c.

E I

w s

F I

y.'

J r-N 4

I I

I l

8 8

8 8

8 e

o G

x e

a o

(d o) 3 8 019 8 3d W 31 reeraero av

  1. 1d. C.

3,re 5 / 2 c,/ 9 '2-occ. wo.

A3

? //1 i 5 oF G5 eevieweo av o,1,

,,c, wo

/

l

Babcock 8Milcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 5.0 Thermal Parameters On Page 4-2 of Reference #6, the film coefficient values used in the analysis for the rapid depressurization transient were calculated.

This film coefficient calculation assumed a flow rate of 425 GPM through each nozzle. The actual flow is a maximum of 335 GPM per nozzle, References #8 and #9. The actual film coefficient that should have been used is calculated below:

In Branch (Per Page 4-2, Reference 6)

The film coefficients were calculated from equation (VIII-1), Ref. #10, Page 139:

h = 0.o'2.3 A M

,g

p. C. p o,9 D

p.

Je.

2 Where:

h = Film coefficient, BTU /HR-FT - F, N = Thermal Conductivity, BTU /HR-FT -

F, D = Diameter Of Pipe, FT,

[

G = Rate Of Flow / Unit Area, LBm/FT2 - HR, M = Dynamic Viscosity, LBm/FT - HR, O P = Specific Heat, BTU /LBm 0F, All at the fluid temperature. The properties will be evaluated 0

at 60 F since the highest stresses developed occurred over the first 0

45 seconds, at which time the temperature was 60 F.

FoR Go* F W ATE R,

C R E F. ^r* G, P A G E H

'2. ]

C. p = 1. 0 Gt w / L b m - F Je_ = o. 3 4 9 G-t u / H e - F 4 -

F p = G2.'S S Lb m / f+2 D = o. I ') 'l F+

p 2. 9 i Lb m / F1 - H e A5 0.o'2.9 G Ft G --C3'3 5 %Llrna /H')(Go m.n /hrh Co.iss'7 4'! o aL3CGQ.39 L bm 147) 0.02H G et*

G G,810,188. cl L6m / Pt -he

~

ener neo av OJ.C.. h b /2b Ib b oAfe OOC. No.

N.$

{/.2.6 P L o41:

p or uo.

iq o= C5 nevieweo av

m Babcock &Wilcox 32-1134218-00 l

nocie., po,7,1;~e% ion oivision GENERAL CALCULATIONS 5.0 Thennal Parameters-Continued h=o.o2s o.wH Co. i q 1 X G B i o, i s 9. ct T.s.( 2.ri n c i. o 3 7 o. :.,

c o.119 (2.911 (c.34 H) h = 3 3 G B. 3 B + u. / h e - M-F 2

U A film coefficient of 4100 BTU /HR - FT F for the rapid depressurization transient was used in Reference #6.

The film coefficient used in Reference #6 for the test transient is 1300 BTU /HR - FT2,op,

(~

This film coefficient was calculated using the correct flow rate.

On page 31 of this calculation package, it has been deter-mined that with a film coefficient value of 4100 BTU /HR - FT2,op, the same aT and stress ratios were computed as with a film co-l 2

0 efficient value of 1300 BTU /HR - FT F,

Therefore, it has been concluded that using a film coefficient value of 3368.3 BTU /HR - FT2,oF would give the same AT and stress ratios.

Therefore, no adjustment needs to be made during the ratio method of analysis.

l s

entraneo av 01 t. h o4re S/2.G / P2 coc. no.

'8' M'8 oire CMfM2-i9 G; G) nevieweo av emoe wo.

Babcock &Wilcox 32-1134218-00

' = " ' " " ' " "

Nuclear Power Generation Division GENERAL CALCULATIONS 6.0 Thermal Discontinuity The thermal discontinuity effects would also be lower if 2 U the lower film coefficient value of 3368.3 BTU /HR - FT

-F for the rapid depressurization transient was used. With a lower film coefficient value, the metal temperatures would change at a slower rate, thus decreasing the axial AT temperature at the time point being evaluated. Therefore, using a film coefficient value 2

F produces a desired conservatism in the of 4100 BTU /HR - FT analysis.

l l

PREPAGED SY Q. H DATE b/O C /SS DO C. NO.

I$ DE ob l

kEvifwfD BY DATE PAGE NO

Babcock &Wilcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 7.0 HPI Nozzle Loads and Stresses Thermal Exoansion Loads The thennal expansion loads for the HPI nozzle are tabulated below.

(Ref. #6, Page 1-5)

Fy (LBS)

F Il03)

Nx (FT-LB)

My (FT-LB)

M W-kB) z 7

200 200 500 500 950 The thermal expansion stresses are calculated below.

From Ref. #6, Page 1-2, R2 = 1.O G 2 5 av.

==5 D a = 2.12 5 i ^>.

Re = 1. 5 l u.

=> De = 3. o i n.

I = ("/d ( R2 - R ? )

I = ("/4 ) [ ( b E I N Y - (. l. 0 Co '2. b.i tJ[ ] = 2. 9 3 I M "

T = C. 2 b rn =,

2I W H E R E ". C_ q = l.2 C REF.7: 2,TAGLE D-7 c l, Fo R TAPERED TR ANsiTrou Jos uTS]

th : = [ rn*< + rn*y rn*3 Y

+ V L Grudi v G

=

t rno rn E M T F} ] "* = S H E A R u u T R\\ 6 v Ttofs To V=

[ F y*

t SE va t u G rn o rn E M T L = D i s T A N C. E T o J u n c.Tv R E 9 8 = 0. 08'33 i:7.

[REF.PI'2.]

s Peeraneo sv Ill C.. h oAfe b/15 / b OOC. NO.

b M M.2 t9 0F G5 oate p.ce uO.

nevieweo av l

BabcockpWilcox 32-1134218-00

' " ' " ~ ' ' " "

Nuclear Power Generation Division GENERAL CALCULATIONS sus i o E

  • 0 =-C1.2)(2.t25)fC(5 coy + C5o03 2 (qgg)2 3v2(;g) s 2^
  • E(toot C too)2]rz (o.oB323 C l 2 3 }

2 c 2..'l B )

7

  • 6.'2 KSl

~

o u T 51 b E '.

7 = Cl.2) C 3.o') { C C Soo)2

( 3 c g $

, c c) g g ) 2 3 '/ 2 C i 2 )

2 2

+ t (20o1 e ( 2 0 0')23 v2 ( o,0 g 33 S c,g g,3 y 2 ( t. 9 B ')

r F = 8.15 K s l Peeraneo av

%.t. h 5 /'2 G / 9 2_

oire ooc so.

Od[

f/2/o 7,2 o re

,.oe so 20 OF G5 inviewee av

Babcock &Wilcox 32-11s4218-00 Nuclear Pow r e tion Division GENERAL CALCULATIONS OBE Loads The OBE loads for the HPI nozzle are tabulated below.

(Ref. #6, Page 1-5) y (FT-LB)

Mz( -LB)

F7 (LBS)

Mx (FT-LB)

M Fy (LBS) 520 300 1100 920 1760 The OBE stresses are tabulated below.

G'= C_2 B m:

2

US: BE

( L.'2d (2. it 5)([ C lio of+ ( 9 2 0Y + (i q Go9 3 D ( l 2 )

V =

- E C 52o9 - C 3 00')23 vz Co. OB33d C 12N 2(2.98)

T= II SI KSI outs i D E ~

7 - C t. 2') C3.o ') [ E ( i 10 09 -r ( 9 t o')2 + C i 1 G o ')2 3 ' 2 C12)

' E C 52 0') * + (.So oY 3 Y2 C o. o B T3d C. \\ 2) ~5 2 C 2. 9 B )

F = \\ G. 8 2 ks1 iteranto av

'D t. #

oat 5 / 2Gf_E_?_

occ. ~o MN3 5~/J/o /81 2i CD G ~c oire

..oe no neviewro av

Babcockp,Wilca 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 2x OGE (55 E) ST R E s s E S ARE C.ALCVL4 reb B E L O W '.

I tJ S I D E.

7 2 C \\ l.91 K S L ) = 2 3.B2 K S )

OU T 51 D E E = 2 ( I G. 9 2 K s, t 3 = 3 3. Co '1 K S I r

g 4

N b bb /h b PREPARED iY b.C.

DATE DOC. NO.

M 8.$

4/o//2 22 O P GE Dare Pace ~O.

Reviewed iv i

BabcockgWilco3 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 8.0 Primary + Secondary Stress Intensities Tabulated on page 24 are the primary + secondary stress intensities at juncture #8 for the HFI nozzle. Juncture #8 is the most critical location in the HPI nozzle.

Reference #2, paragraph F-104.4, gives the prir.ary plus secondary stress intensity range limit as 3 Sm.

If the 3 Sm limit is exceeded, then an elastic-plastic fatigue analysis must be performed in accordance with Ref. #2, paragraph F-105.2.7.

(The 3 Sm value at the critical location, Juncture #8, is 51.3 ksi, Ref. #6).

This fatigue method is valid only if the number of cycles that exceed 3 Sm are less than 250. The number of cycles that exceed 3 Sm are determined on page 29.

N O

l l

l s

k_C, DATe b /2S /kb Pt PAttD BY DO C. NO.

/[ 8 $

o re f/J [/.2 2.3 OF G3 Reviewco av paos so.

(

b

\\

5l Table 1 L - Longitudinal 2

(D 7

Primary + Secondary Stress Intensities R - Radial 5

03 At Juncture #8 (L-R Intensity)#

{ {

~

O

  1. Table references and explanations are included on the following page.

o o

d O'#

c

! 9' Stress Report a

k (ksi)I Stress (ksi)

(ksi)3 Intensityjksi)7 @ i ll6 k

i Iteration Pressure Pressure Stress Thermal Expa sion Thermal Stress Total Stress p

  • =

kO h

Inside Outside Inside Outside Inside Outside Inside Outside

.2x 3

2200 4.84 1.32 6.2 8.75 63.7

-58.2

'Pi Pi

-48.13 S.

6 2200 4.84 1.32 6.2 8.75 77.5

-70.8 83.54

-60.73 o

8 2200 4.84 1.32 6.2 8.75 78.8

-71.9 89.84

-61.83 15 2200 4.84 1.32 6.2 8.75 69.3

-63.0 80.34

-52.93 E

30 2200 4.84 1.32 6.2 8.75 7.7

- 6.9 18.74 3.17 E O 1211 2200 4.84 1.32 6.2 8.75

-0.4 0.3 10.64 10.37 (C

1491 2200 4.84 1.32 6.2 8.75 0.0

- 0.0 11.04 10.07 s

42021!

2200 4.84 1.32 6.2 8.75 0.0 0.0 11.04 10.07 g

?

?

27976 2200 4.84 1.32 6.2 8.75

-0.7 0.6 10.34 10.67 5072 2200 4.84 1.32 6.2 8.75 86.0

-78.3 97.04

-68.23 g) H h ?

5060 2200 4.84 1.32 6.2 8.75 73.1

-66.8 84.14

-56.73 mQ sh ]

5063 2200 4.84 1.32 6.2 8.75 97.9

-89.3 108.94

-79.23 2A 5124 2100 4.62 1.26 6.2 8.75 33.0

-29.7 43.82

-19.69 e

g N

5237 1800 3.96 1.08 6.2 8.75 11.5

-10.4 21.66

- 0.57 g0 1 y

s N

n 6059 1000 2.2 0.6 6.2 8.75 0.4

- 0.4 8.8 8.95 yH D p 6307 800 1.76 0.48 0.0*

0.0*

0.4

- 0.4 2.16 0.08 gg 6338 700 1.54 0.42 0.0*

0.0*

-7.7 7.0

-6.16 7.42 c

o 6

600 1.32 0.36 0.0*

0.0*

-3.5 3.1

-2.18 3.46 O

[ 8 A

1600 3.52 0.96 6.2 8.75 103.77

-94.66 113.49

-84.95 )o l

1 B

1100 2.42 0.66 0.0*

0.0*

-8.67 7.89

-6.25 8.55 r

OBE(0 E(5) 1 2x 0.0 0.0 0.0 0.0 0.0 0.0 0.0 23.82 33.64 OO l

o 6

0.0 0.0 0.0 0.0 0.0 0.0 0.0 111.91 t16.82 C

e r

C D

E O

d g

y O

E O

2 9 Un t.n 2

\\

l Babcock &Wilcox 82-1134218-00'"'*'""

Nuclear Power Gen'eration Division GENERAL CALCULATIONS Table References and Exolanations 1.

In the Stress Report, Ref. #6, Iteration 1 was run of pressure only on the nozzle at a pressure of 1000 psi. Therefore, pressure stresses are obtained by multiplying the pressure stresses from Ref. #6, page 5-17 times the ratio of actual pressure /1000 psi.

2.

Themal expansion stresses are calculated on page 20 of this analysis.

3.

Stress Report Thermal Stresses are from Ref. #6, Page 5-17.

These thermal stresses are conservative since they were calculated with a flow rate of 425 gpm instead of the new flow rate of 335 gpm (Ref. #8 and 9).

Therefore,- the film coefficient were higher than necessary, thus increasing the themal stresses.

4 Transient A is the start of the 70 cycles of HPI manual actuation.

Transient B is the end of the 70 cycles of HPI manual actuation.

This transient occurs following a reactor trip transient. The pressure stresses are calculated according to Note 1 above. The thermal expansion and themal stresses are calculated using iterations 5060 and 6338, and adjusting the stresses using a AT ratio. These calculations are presented in Section 8.2 of this analysis, l

5.

Full range OBE (2 x OBE) stresses are calculated on page 22 of the analysis for l

juncture #8. This is the stress range for a cold earthquake.

6.

Hot earthquake stresses and cycles will be applied and analyzed for in the primary + secondary and peak stress intensity range sheets in this calculation package. OBE stresses are calculated on page 21 of this analysis.

p l

Pairaneo sv

  1. h. C_

h o4re 5/2C / R 2 oce. so

'[d [d 3 oire [/2 //.2 2 % G.: C E.

neviewso av p.ce so.

/

Babcock &Wilcox 32-1134218-00'"*""'"

Nuclear Power Generation Division GENERAL CALCULATIONS 7.

Total stress intensity is obtained by adding pressure stress + themal expansion stress + stress report thermal stress.

  • Note - A value of 0.0 ksi is used to approximate the range of thermal expansion stress at the HPI nozzle end due to the change in temperature of the HPI line.

- The cross-section position that experiences positive themal expansion stresses is used to maximize the inside intensity which is the critical intensity.

- A graph of the L-R inside intensities is shown on page 27 of this calculation package to aid in determining ranges for maximum and minimum intensity values.

w M

PetPAatD BY 03 b-D Af t bIM /b9-DCC.NO RiviewtD BY DATE PAGE NO Nr O b

Babcock &Wilcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS Ci G U REG P R t m h R Y t S ECO ND A R Y E F. E 5 5 f NTENSITIE S A

190 12 0

\\O0 C

BC U1 M

wc 60 3x 2

- 3

!O 90 20 o

oL G50 car C.'/ C L E $

wo TEST h' Nee,[.$2D*C MPro g?aMrssunitaTioNy

-20 I

I I

I I

I I

I I

I I

I I

- - G o m ~

c-

.~. a3 o w

'm w cc w o =

f 7 7 j j E :oJ c1 0 F v1 o M v1 c

~

a e 2 25gg$o

- 9gm ITER ATION N Um6ER reeraneo av du. t. 'G 0 h oAre s / c 6 / o '2.

occ. ~o.

REvitwtD BY DATE PAGE No.

b

/

/

Babcock &Wilcox 32-1134218-00 nociear po 7r"J!e'Tation oivision GENERAL CALCULATIONS 8.1 40 Test, 240 Heatuo and Cooldown, and 40 Rapid Deoressurization Transient Cycles The primary + secondary stress intensity range will now be calculated for the 40 cycles of test transient, 240 cycles of heatup and cooldown transient, and 40 cycles of rapid depressurization tra* sient. The temperature transient for these cycles is shown in Fig :re #1, Ref. #6.

The stre<

intensities are tabulated on page 24 and shown in graphic fom on page 27 of this analysis.

Ptf Pit (O $Y ggyg hI

/h OOC. NO.

8/f 5 oire { /2 6,/ # A

, ce ~C 2%

oF c5 uviewto av

Babcock &Wilcox 82-1134218-00

~

Nuclear Power Generation Division GENERAL CALCULATIONS Maximum Primary + Secondary Stress Intensity Range This range is comprised of }(Iteration 5063) - (Iteration 6338) ).

It occurs during rapid depressurization and can occur for 40 cycles.

Qrfg{sec 108.94 - (-6.16)

= 115.1 ksi > 3 Sm = 51.3 ksi

=

2nd Maximum Primary + Secondary Stress Intensity Range This range is comprised of (Iteration 8) - (Zero Stress State)

It occurs during testing and can occur for 40 cycles.

TP" 89.84 - 0.0

= 89.84 ksi > 3 Sm = 51.3 ksi

=

ag 3rd Maximum Primary + Secondary Stress Intensity Range This range is comprised of } (Iteration 42021) -(ZeroStressState)l.

It occurs during heatup and can occur for 240-40 = 200 cycles.

7Prfg{sec 11.04 - 0.0

= 11.04 ksie-3 Sm = 51.3 ksi

=

The number of cycles in which the primary plus secondary stress intensity range exceeds 3 Sm is 80.

I e

a-Ptf P Att0 Y

, b.

D A TE IbS/b DO C. NO.

/885 fd F1 29 OF G5 Dire Pion No.

nevieweD ev

BabcockgW,i1cox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 8.2 70 Cycles of HPI Actuation and OBE Cycles The 70 additional cycles of HPI actuation following a reactor trip undergo a different temperature transient.

(Ref. #8). The transient 0

starts at a temperature of 579 F and a pressure of 1600 psig, droos to 0

0 60 F during the next 45 seconds and then drops to 40 F.

The transient 0

ends at a temperature of 558.2 F and a pressure of 1100 psig. The maximum flow rate is 335 gpm per nozzle.

(References 8 and 9) These additional 70 cycles will be justified by adjusting the stress intensity ranges calculated on the preceding pages for the rapid depressurization transient.

To justify the analysis of the 70 additional cycles for the higher 0

0 starting temperature (579 F versus 550 F), three (3) analogous themal stress runs were made using B&W computer program P91232. Using this slab temperature program, the HPI test and rapid depressurization transien'es were analyzed as to their effect on the nozzle end thennal stress without discontinuity effects. The results of this analysis s

are contained in Reference #7, Microfiche AC3IMJU.

w t.C_. O o.ri 5/ 2.6 / 9 2 reer.eco av occ. ~o SO OF G5

/f5 2u/ L

    • ce so.

neweweo av o re

Babcock &Wilcox 32-1134218-00

=>'a.'~"

Nuclear Power Generation Division GENERAL CALCULATIONS The pertinent results are tabulated below:

AT Film Linear Case Initial (T-60)

Coeff. 2 Ratio (I)

Stress (PSI)

AT Thennal Stress (2)

Temp.(OF) oF (BTU /HR-FT.oF)

Ratio 1

550 490 4100 1.00 73106 1.00 2

570 510 4100 1.04 76023 1.04 3

585 525 4100 1.07 78204 1.07 4

550 490 1300 1.00 57131 1.00 5

570 510 1300 1.04 59350 1.04 6

585 525 1300 1.07 61005 1.07 Notes - (1)6 T Ratio = 6T/oT) = aT/490

{

(2) Stress Ratio

=S 2 or 3/51 for Cases 1, 2, 3 i

5 U" 6/3 for Cases 4, 5, 6 Stress Ratio

=S 4

To justify the analysis of the 70 additional cycles for a higher 0

return temperature (558.2 F versus 5000F), the following results from B&W computer program P91232 are tabulated below.

These results are from Reference # 7, Microfiche AC31MJU.

I AT Film Linear Case Return (T-40)

Coeff. 2 Ratio (I)

Stress (PSI)

AT Thermal Stress (2)

Temp.(OF) oF (BTU /HR-FT OF)

Ratio I

1 550 510 35 1.00

-5658 1.00 2

570 530 35 1.04

-5876 1.04 3

585 545 35 1.07

-6039 1.07 Notes - (1) AT Ratio = 4T/oT1 = b.T/510 (2) Stress Ratio = S or /3 2

3 1 s

retraero av % t. h oire 5 l'2.6 / R 2 ooc. so.

M3 o4re(M4)P2 race wo.

'E ! OF G5 eeviewee av

Babcock &Wilcox 32-1134218-00

> > = > ~ > >

.m............,

Nuc3 ear Power Generation Division GENERAL CALCULATIONS 1

From the tabulations on the preceding pages, one can see that the AT ratio is the same as the stress ratio for each case. Therefore, there is sufficient justification for using a AT ratio method to justify the different temperature transient from Reference #8 versus the test and rapid depressurization transient from Ref. #5, Page 4-4.

In order to calculate a stress intensity range for the reactor trip followed by HPI Manual actuation, the following method will be utilized:

Thermal gradient stresses from the rapid depressurization transient will be multiplied by the AT ratio to obtain the thermal gradient stresses for the HPI manual actuation transient.

Pressure stresses for the HPI manual actuation transient will be similarly adjusted. Thermal expansion stresses will not be raticed because these stresses are due to the temperature in the HPI Line, and the only transient that changed was the cold leg temperature.

The following calculations apply to the 70 additional cycles of I

HPI actuation following a reactor trip. Two cases will be analyzed during these 70 additional cycles; one with the inclusion of 30 : OBE stresses and the other without the inclusion of + OBE stresses. The number of : OBE cycles analyzed is 650 - 30 = 620.

1 l

l l

s emeraneo av H d h

can 5 / 2 C / R 7_

ooc. no 8.5 oan { 16 71 32 OI Eb nevieweo av

    • oe No.

Babcock &Wilcox 32-11s4218-00 Nuclear Pow r b e ation Division GENERAL CALCULATIONS

~

or, c v rr J m F R i m A R Y + 5 E t o u c A R Y sv R E S S I N T : u e. i t y (losioEl

~ = t s c is A PRESSURE ST REis = CH.8M KSI) IG00 Ps i

'S. 5 2 K S I

=

2200 PSI T H E R rn A L E K P A tJ a l o tJ STRESS G.2 K S I T H E R yd A L ST R E S S = ( 9 7. 9 k s i)[5 ?9*F - 60 F

= los.? ? k 5 i G50 F-60 F TOTAL ST R E S S : 3. 5 2 + 6. 2 + 10 3. 7 '7 = I I 3. M 9 K s i

.{

ti E R Ai t o rs G PR E S SU R E ST RE S S :-(I 5 4 KSI) i 10 0 P S I 2.H '2.

K S i

'7 0 0 Ps i T H E R rn A L E A P A N S t o tJ ST R E S S = 0. 0 KS1 TH E R rn6 L ST RE S S

-9.9 XS I 5 5 9 '2 *

" O* F

= - 8. G ")

K Sl so o

  • F - v o F iOTAL ST R E S S = '2. H 2 + o. O - 8. G ') 5 - G. '2. 5 K S I 1

l l

entraneo av of C. h care S ['2.G / B'2.

occ wo.

[8$

fM6//2 o.n

    • ce so M OF Cc b r

nevieweo my 1

Babcock &Wilcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS rr 4 rc o m P P t rw A P Y + E E tor; n A c y c -- o r % c i >>- c r; c i -v (oursin:-)

l'* E R A T f oM A FRESSURE STRE S$ = (1.32 KS d 16 00 PS1

= 0.96 KSI

~

22co PS I T H E R m A L E RP A rd s t o /> STR E S S = B. 7 5 K s i "e amn L sra e s s = (- 89.3 ks i)(/sli F - G O F

=

C)4.GG KS]

55o F-Go F tot A L STRESS: 0. 96 + 8.7 5 - 9 4. 6 G = - 8 4. 9 5 k 5 i

\\ ~r E R A *T* I O M 3 r

PRE SSURE STRESS : Co S?. cli) _Ilo o P s i

0. G G k5I

=

900Psl TH ER m A L E xP A USlou ST RE SS = 0. 0 k 5 i T H E R rb A L ST A E s 5 = ( 7.0 r s i) 559.2 F - H o F

= 7. 8 9 k S I Soo F - H o F TOTA L ST RE s s = o. G G + o. o - 7. 9 '1 = 8. 5 5 K s i

  • =ecameo av Al C.. h care A,/ 9_ G / 8'2-occ. wo

[NS oate EM M2

'3 " O P CO Q

eevieweo av

    • oe ~o

Babcock &Wilcox 82-1134218-00

'="'a"""

Nuclear Pow r e ation Division GENERAL CALCULATIONS Case 1 - Primary + Secondary Stress Intensity Range w/ Inclusion of

OBE Stresses The maximum primary + secondary stress intensity range is comprised of (Iteration A + OBE) - (Iteration B - OBE) and can occur for 30 cycles.

QP" (inside) = f113.49 + 11.91) - (-6.25-11.91) 8

= 143.56 ksi ra g

>3Sm = 51.3 ksi pri

{ rang +e sec (outside) = 'I(-84.95 - 16.82) - (8.55 + 16.82)

= 127.14 ksi >

3 Sm = 51.3 ksi Case 2 - Primary + Secondary Stress Intensity Range w/o : OBE Stresses 7P" (inside) =

113.49 - (-6.25)

= 119.74 ksi > 3 Sm = 51.3 ksi ra g

(

P 7 ra g (outside) = (-84.95) - 8.55 { = 93.50 ksi > 3 Sm = 51.3 ksi This range of stresses can occur for 70 - 30 = 40 cycles OBE Primary + Secondary Stress Intensity Range (inside)

This range is comprised of OBE and can occur for the remainder of the earthquake cycles = 650 620.

7Prfg{sec.(inside): 11.91 + 11.91 = 23.82 ksi4 3 Sm = 51.3 ksi The number of cycles in which the primary plus secondary stress ut.9sity range exceeds 3 Sm is 70.

The total number of cycles during the operational events in which the primary plus secondary stress intensity range exceeds 3 Sm is 80 + 70 = 150 4 250. Therefore, the elastic-plastic fatigue analysis is valid.

%. C.. # O A oAfe 5 /'2 A / 9 R ener*seo sv occ. ~o 8AE oir,6%,

35OF GE

,,ce so eeviewco av

BabcockgWilco!

32-1134218-00 Nuclear Pcwer Generation Division GENERAL CALCULATIONS 9.0 Peak Stress Intensities Following is a tabulation of tre. peak stress intensities at juncture

  1. 8 of the HPI nozzle. Only the stresses at the inside of segment #8 will be tabulated as they are the most critical.

t.C.. b 'O h care 5 /26 / 9'2.

eneraero av occ. so 8/f$

oire {/h /82 3G C# G3 ervieweo av

,4ce ~o

Babcockswitcox 32-1134218-00

>="'a'""

Nuclear Power Generation Division GENERAL CALCULATIONS L - Longitudinal Table 2 2 - Radial Peak Stress Intensities At Juncture

  • 8, L-R Inside"
  1. Table references and explanations are included on following page.

Pressure Thermal Stress Report Total Iteration Pressure Stress Expansion Thermal (psi)

(ksi)1 Stress (ksi)2 Stress (ksi)3 Stresg (ksi) 3 2200 5.5 7.94 109.0 122.44 6

2200 E. 5 7.94 115.0 128.44 8

2200 5.5 7.94 113.3 126.74 15 2200 5.5 7.94 95.0 108.44 30 2200 5.5 7.94.

8.7 22.14 1211 2200 5.5 7.94

- 0.4 13.04 1491 2200 5.5 7.94 0.0 13.44 42021 2200 5.5 7.94 0.0 13.44 27976 2200 5.5 7.94

-0.8 12.64 5060 2200 5.5 7.94 147.4 160.84 5063 2200 5.5 7.94 143.3 156.74 5072 2200 5.5 7.94 117.3 130.74 5124 2l00 5.25 7.94 39.9 53.09 5237 1800 4.5 7.94 14.2 26.64 6059 1000 2.5 7.94 0.5 10.94 6307 800 2.0 0.0*

0.5 2.5 6338 700 1.75 0.0*

-9.6

-7.85 6850 600 1.5 0.0

-1.3 0.2 A(4) 1600 4.0 7.94 156.24 168.18 B(#)

1100 2.75 0.0*

-10.81

-8.06 2 x OBE(5). 0.0 0.0 0.0 0.0 30.49 OBE 0.0 0.0 0.0 0.0

15.24 PetPAnt0 BY kl. b.

DATE b b/h DO C. NO.

'M O F Cc 5 stVIEWED BY DATE PAGE NO

Babcock &Wilcox 32-1134218-00

' " ' * ~ ' " "

Nuclear Pow r G e tion Division GENERAL CALCULATIONS Table References and Explanations 1.

Pressure Stresses are obtained by multiplying the pea pressure stress from Ref. #6, Page 5-17 (Iteration #1) times the ratio of actual

\\

pressure /1000 psi.

s 4

2.

Themal expansion stresses are from page 20 of this calculation package for juncture #8, multiplied by the bending stress concentration factor from Ref. #6, Page 1-4 (KB = 1.28).

3.

Stress Report themal stresses are from Ref. #6, page 5-17.

4.

Transient A is the start of the 70 cycles of HPI manual actuation.

Transient B is the end of the 70 cycles of HPI manual actuation.

This transient occurs following a reactor trip.

The pressure stresses are calculated according to note (1).

The thermal expansion and themal stresses are calculated using iterations 5060 and 63,38, and adjusting the stresses using a AT ratio. These calculations are presented in Section 10.2 of this analysis.

5.

Full range OBE (2x0BE) stresses are from page22 of this calculation package multiplied times the bending stress concentration factor from Ref. #6, page 1-4 (KB=1.28). This is the rance for a cold earthquake.

6.

Total stress intensity is obtained by adding pressure stress + thermal expansion stress + stress report thermal stress.

  • - Note - A value of 0.0 ksi is used to approximate the range of thermal expansion stress at the HPI nozzle end due to the change in temperature of the HPI line.

PetPAtt0 gy k C.

b b/b D Aff DOC. NO.

bb W RtylEWED SY DATE PAGE NO.

BabcockgWilcox 82-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS

- The cross sectional position that experiences posi11ve thermal expansion stresses is used to maximize the critical inside intensity.

- A graph of the L-R inside peak stress intensities is shown on the following page to aid in determining ranges for maximum and minimum intensity values.

b / kh /hh PREPattp gy dl,(,

gayg DOC. NO -

[N$

( lk O

09 OF G5 esce NO.

enviewto av o4rg l

Babcock &Wilcox 32-1134218-00 Nuclear Pow r e ation Division GENERAL CALCULATIONS P t_ G U R E 7 PEAK E "" R E S E IMTEMEITIES

/\\

i G O._

g uo u j

\\20 too -

Ca 90 r

v C

Go 4e i

.J 90 20 o

G50 oGE c_y c.t e s l

1 l

l t

i l

I l

l l

l I

I I

I I

I I

N h - k N

hh$

S3$sG$3o

Em e

ITE R ATio^/ NumSER PetP Anto sv k

C.

DATE 5 / 2 G / 9 2-occ. wo.

b PAGE NO.

O O *~- GE REviewto SY DATE

/

/

5

Babcock &Wilcox 0 2 - 1 13 4 218 - 0 0

=->'a

> i">>

l' Nuclear Power Generation Division GENERAL CALCULATIONS 10.0 Fatigue Analysis The fatigue analysis is performed in accordance with paragraph F-105.2./ of Ref. #2, " Simplified Elastic-Plastic Discontinuity Analysis." The peak stress intensities and cycles are presented in graphic form on page 40. These are used to determine maximum peak stress intensity ranges and cycles used in the usage factor calculations. Actual peak stress intensity values can be obtained from the tabulation on page 37 Primary + Secondary stress intensities are also used in the fatigue analysis; these values are obtained from the tabulation on page 24.

r(

l l

s m

reer neo av

% - C.. h oare 5 /GC / 9 '2-occ. no.

883 s,2(hJ2 4I oF Gi oire

,4ce so _

neviewco av

3.

Babcock &Wilcox 3 2 - 113 4 218 - 0 0

' "~ ""

.~o....n.....,

Nuclear Power Generation Division GENERAL CALCULATIONS

,0.I "O ' TEST 240 HEAT-VP AND Co c L-D o w tJ AND Ho RAFID B E P' ~ S 5 U R i ? A v i o u - R A N S. ENT CytLES m A W rn u m P E A K STRESS i MTE M SIT Y RA M G E THis RANG E IS Com F R \\ S E D OF C ITE R A Tiou So G o)

- ( IT ER ATioN G339 CT ottu R 1 Du RiN G RAPtD DEPR E SS URi ? ATioM AMD CA M octu R FoR 40 C_y t L E S.

Tu.P Kss e = ( 1Go. B O - C-7. % 53 PG 1GB G '1 Ksi

=

T HE Prim A RY r 5,E C o N D A R Y STRESS I M T E N S, l T Y RANGE f

A SSoCI A T6 0 WITH THESE TR A N C I E /V T s = 11 5. )

ksi

'3 5 m -=. 51. 3 k s i tiTERATioNs So G 3 AND GTE6]

THE RE FORE, A N E L A S T I C. - P L A S T I C.

AN A LY S I S rnuST GE

'to CY C LE 5. F o L L ow I M 6-THE P E R F OR rn E b FoR THESE PRO CE tuRE OF R E F. M ?_, P A R A G R A P H F - i o s 2. 7, T H E fin A L PE A K A LT E R N A T I N G STR E S S IN TE N S TY, 78tr, I S ".

i cm

'/ 2 k p K e S e q,

Tat-A CK

-l)=

oVE RA LL F A TI G U E

=

K

?

WHERE Kp g

3 ETRE N GTH REbutilov FACTOR (P

b r)a,

fEAK 5 " A E S S, iNTEN5l'Y RA46E Kg

~-

t 57.O FRim ARY t S EttN B A RY STRE SS I NT E NSITY R4NGE

~

Ptf P ARED BY DATI G!

DO C, NO.

N#_$

4, /A 42 oF GS e.re Pace ~O

.evieweD iv l

l

Babcock &Wilcox 32-1134218-00 Nuclear Pow r e ation Division GENERAL CALCULATIONS

= oV E R A L L E F F E CT i V E STR E 15 Cc u CE N TF.A-Ti o N FOR THE C.Y c. L i C.

Logoi N G A5 0. 9 FoR STArNLESS STEE L A-39G, TP 316 F Rom F

G D-20i C c.), R E F. ** 2 S n rs F L I F y n M G,

1.47 IG B. G 9 K S i K

=

=

A 11 5. I KSI

.. Kr =

1. 9 '? + o. 7 ( 1. 9 9 - 1. o ) =
1. 9 o Ns

\\S DE TE RM INE D F Rom PIG.F~105(1).

Fo L L.o k/ I N G 15 THE C A L tV L A Ti o N OF THE P A R A rn E T E R S

/*(,

NECESSARY To 06TA I N Kg Sn P Rim 4RN + SEtoNB ARy STRESS inm e nsiv y RANGE l

35m SSm h

-11 5. 1 kst -

'2. 2 %

=

=

35~

5l.3 Ks\\

{ QH

[lGmlt {Qej]

t5 DETERMINED N E xT, wHERE*.

lQmj 15 THE rn A G Ni T V b E OF Prim A R Y t SEto N D A RY M Em 8RA ME STREIS IMTEN S I TY R A NG E AVERAGED TH ROUG H THE TuttkNESS OF THE S E C.T l o A/.

IGb l 1S THE Th A G N I T V D E OF PRtrnARyt S E co u b ARY

~

GENblN G STREES lA) 7 E N Il 7 Y R A /v G E 5MG/92 PREPARED BY k.C.

Daft Doc. ~o.

OAC5 REVIEWED BY DATE PAGE No

l Babcock &Wilcox 32-1134218-00 Nuclear Pow r G e ation Division GENERAL CALCULATIONS AVER A G E D TH50vGH T' H E T M 1 0. K tJ E E I.

CP THE S E C.T i o M.

CCoNtiTrou lo-

- ITE R ATicu so coM

'oE.99 rli

'9.e55 kg i 93.ess asi

,,m s a suofnte a

s

[

oursioe svemate

-qa.23 nsi sw.gss gs i 4 3, o g g g 3, IM 215 ksi

Qm O g -- = ico.eq s ksi s k C ON blTIO N Ib% IT6 8S TlON 733 b

,osine sogeats

-G.'G K s s o.G1nti

-G. 9 9 " 8 N

+

\\

ouTSIDE suR FA tE 1.92 K s i o.63 ns i G.q 9 xs i O-

=

19. 2 2 5 k s i

= 0.I'2.

Os l 'i. 2 2. 5 x s i e i c o. S ') 5 tes i Om v

F RO rA F i G. F - 10 5 C o-), R E F # 2,

i. ")

Ke c ce

'/2 k [ Kg b C YT %

AL

'/2 C i. 8 0)(L 9) lll 5. l ) = 11 Cc I o k $ i 74 t-

=

PetP AttD SY OJ b DAfg 5 / 2_ c / A '2_

occ so 885 f/26 //2 4 4 o F C, ~n oAn

,.ce so neviewso sv

Babcock &Wilcox 32-1134218-00

'o"' a ' ""

Nuclear Power Generation Division GENERAL CALCULATIONS O ~~ A II E

7J ~ O A CCO U td T THE E F F E c.T OF E L A S T r C.

m o o v L v 5, IfJLTIPLY U s.T GY THE R A TI O OP THE rnoDvLv!

0F E L h S T I C. l ? Y G I V E A/ ou THE DESIGu F A TI Gv E tuRVE To TH E.v A LU E OF THE rn o b u u u s of El As Ti ct T Y USEb IN ThE A N A LYS I S. DU RIN G THE RA Pl b DE PREsSu R \\ 2 A TICM TRA NSI E N T, TH E TE rnP E RA TV R E VARtES FRom 55o F

To Mo F.

AT Av AVER AG E TE mFE R ATU R E OP 295 F) 55cF +Wo F o

=

~

2 G

E = 2 1.l'S w to Psi C R E i:. 7* 2. ')

G

. ~. S o. : 11G 10 26 wio

=

1 G 8. 7 ')

ksI 6

'P_ 1. I'3 x 1 0

[THE D E SIGN FATIGUE tuRVE VsED iS F i G. ? - l O G CJ),

RE F. F 2 Ecur<vE

  • 2 G w 10 P S I.)

FRom F I G. F - I O G (. b), R E F. t* 2, NS A LLoW A SLE CY CLE s = 2 5 0 U,=

USAGE F A CTo R =

No CF R E Q' D tyc L5 5 Mo. OF e LLo wA 8 L E ty C LE S

'll. t.. b'A oire 5 / Z G / A '2_

raieanio av occ. ~o.

/785 oreEMhd2 G OF G-niva wee av

    • ci ~o

Babcock &Wilcox 3 2 - 113 4 218 - 0 0

'="'#"'

Nuclear Pow r e tion Division GENERAL CALCULATIONS U,=

HO

=

0. I Ce 25o 2e& m a x im o rn PEAK STRESS J NTE N S I T Y R A tJ G E THis RANGE i s co m PR \\ s E D oF

(, l T E R A T I o N fc) -

-(.?ERO STR E S S S T A rE)

. CT octu R S Du R t N G TE ST t u 6 AMD CAN oct u R FOR 90 ty C L E S.

7sIIse (12.8.'i4 ) - (0,O) 12 8. H H K S. I

=

THE f* R \\ m A R Y t S E co tu D A R Y STR E SS IN TEN S I T y RANGE A S Soc.l A TE D

\\^> l T H THESE TR AN SIEN TS i1 6 CI. B M K S I >

l 3 5m = St. 3 KSt C. l T E ft AT r o A> l 9 AN b I-E R o STREIS ST ATE 3 THE RE FOR E, A N ELASTIC # P L A S T l C.

FATIGVE l

ANA LY S IS musT SE PE R FORME D FoR THESE

'f o C. X C L E.S.

FoLLowl N G THE PROCE D U R E OF R E F. M '2., P A R A GR A P H l

F - 10 5. 7.. "), THE FluR L PEAN A LTE R N A T I ^) G STP LS itJ T 6 tJ S I T Y, 7 A uf iS~

1 1

$(n )

\\l AL'~~ f2 Y S W q,

rag, A(Kx-1.o)

WHERE K4 - Kg

+

s PetPAteD 87 k, b-DATe b / 9 $ / b '2-DOC. NO.

885 oire (M@ 82-HF OF GE Pace No.

o neviewco av

BabcockpWilco!

32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS U

S c;),

i'?_ 9 H H KSt 1.43 kg 5

-=.

SE" 89,8S M Si a

45 0. 9 o F, STAINLE55 STEEL A - S 1 G, T P 31 G,

FRom F i G. D

'2 o I C C.T, R E F. t* '2.

K f = i. 9 3 + o. 7 ( l. H 3 - 1. o l =

1.73 Ke is oE vc Rmi ME D F R o rY\\ F I G. F - i c s C o.3, F c L L e w s N G tS TrE CA L LVL.A Tion) cP THE P A R R t> E T E R L NECESSARy To O S TA I N Ks Sn _

9 *l. B H KSI

=

1. 7 5 S

'3 5 m s 1. 3 Ksa Qm\\[E)Qm +lOs 7 is D E T E R ro l N E b G E L o tu ;

( C.c tJ b tT I O N t a = I TE g a r l O A/ b) tusmE SURFACE 6064 Kg, i9,ee s xs :

9 5. B 3 5 t i l s

A

~

/

/

/

J r

~

visine su R F A e.E

-Gl.22 K s i M.oos ESI

-9 5'. 215 K i l

( C.O N D I T I O N

'2. b =

BERo S T R E 5.S S TA TE )

i

. '. Om=

19. 0 0 5 +
0. o =

I 4. c o 5 k s /

Qb5 9b.835t O.O%

) b. 8 3 5 K $~ l l

i b

b/b PREPARED sY k

(_

Daft 00C NO b

8 82 M oF G S neviewto av oart PAGE No i

BabcockgWilcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS Om IS.005 K11

= 0. lG

=

Oc

19. o o 5 v s. i + 15. 9 3 5 rs i Om T F R o rn FtG F - t o 5 C o.), R E F. p 2, Ke % i. Y v2 k ; K e. S<. >

L.e

=

cc d

h- = Vz (.1. #10 (i.9 ) ( 8 ct. 8 H ) = t o 8. 8 o e s. I bu R I N C-THE TEE TIM G TR A NS l E M T, TH E TE m F ei 8 6 TU R E VA R IE S i= R o rn 550 F To Go F.

A-T AM AVE R A GE

(

TE rnPE R A Tv A E 0 1:

550 Fr GO F 305 F,

=

2 E = 2 9.09 *lo 6 P5i C. 8 E F. 7* 2. ]

6 SA 109.80

'2 6 x l0 loH. H G KSI

=

+

29 o8 s (oG F R o rn 1: 1 G.1: - t o (o C b), R E F. W '2.,

N=

A LLo wA SL E CYCLES =

1100 Cr C LE J l

U2= USAGE FA t. tor =

40 C. c 4

-=

Il o O 5/26/A2 b

oAft PREPAtto BY k,C.

h occ. so.

885

/.2 82 Pace so.

  • S OF G *.

neviewso By o re

BabcockgWijcox 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS

'l e-a m A v er u m PE A K ITRESS t v T E n> c i T v RAoGE Tn tS RA N G E is Comf RIS E D OP l CITERATicu M 2o'2.13 -

STATE))

(EERc STR E S S IT octu R S DURIN G H E AT-V P AND C.oc L-D O W N AND C. A u ottvR FoR

'2. H o - H o =

'200 c.y c. L E S.

N [$ce = / ( l 'S. H S ') - ( o. o')

13. 4 H K S., i THE F R l /h 4 R 'Y t S E CO N b A R 'y S T R E 5.5 1,v TE N SI T y RANGE L S S O C.I A T E b woTH THESE TR ANS I ENTC sG ll. 0 4 ns!

?

4. '3 S m = 5 1. 3 K S 1 F_ i r E R A T r o u l 9 2 o 2. I AMb ZE Ro S T R E.S S STATE 3 TH E RE FO R E, THE PIMAL P E h r:

A LT E R N A TI N C-STRECS I N T E N S l T 'Y R h M C-E, Tei.r, l S '.

b r =-

'/ 2. M I O fre = Y2 C i'S. 4 4 K I i ) = G. 1 '2.

k s i B v F, i N C-THE NJ4T-vP ANb C.c o L - b e w u T R A M I l c /\\/ 7, THE TE mPE 86 Tu RE

\\> A R I E S pro rvi 55o F To ')o F.

AT AN AVE RA GE TE mf E R ATuR E cF 5 50 Fe 9o F 2

C '3 10 F,

Et 2 9. 0 5 x 10 G PSl EREF.#'2]

'b t.. Mh o re 51 e 6 / 9'2.

ene**aeo n occ. ~o 885 SYG/fA 4Cf 0;- G S neviewee n otte pace wo

Babcock &Wilcox 32-1134218-00 Nuclear Pow r e ation Division GENERAL CALCULATIONS 6

Ex= G. 7 2.

_2 G x 10

=.G.%G K51 6

29.05 %lO F R o rn F 1 G. F - 10 G (. b ), R E F. ty 7,

N A LLo W A B LE C.y c_ L E S = CO Us

  • USAc-E FACTOR

=

200

= o. O Co r(

l Peeraneo sv t, t. Mh oar, 5/ 2.6 / R R oc c. ~o.

((d $

oate ( 20,/82 EO OF G5 erweweo av race so.

Babcock &Wilcox 32-1134218-00

'="w='""

~

Nuclear Pow r ne ation Division GENERAL CALCULATIONS

.O'2.

10 CN C LE S OF H P I A CTU AT i c N AND OBE CY C LE S THE PE A K ST RE S S ES ARE Now C A L C.v L A T E D COR THE 90 CY C LE S OF HPI

/h A N V A L ACT V A T l o N Co LLowt M G A R E A CTo R TR \\ P. TH E S E lb CYC.LES UNDERGO A

DlFFERENT T E rn P E R A TU R E TR ANSI EN T TH A M FCR T HE 40 CY C L E S OF TEST T R AN SIE N T AND HO ty C L E S OF RAl lb D E P R E S S U R i l A T I o /v TRA N S I EN T. TMuS,THE FEAK STRE S S E 5 wiLL IS E F

s A D3 0 STE D USING S

67 RATIO Th E T H O D.

l

/b PttP Att0 By k, C.,

b DATE DO C. N O.

/885 Dare{/A6//.2 5! nF G5 RevaweD av

,,c, s o.

Babcock &Wilcox 32-1124218-00 Nuclear Pow r e ation Division GENERAL CALCULATIONS rr. A x i rr. u m cEAK s-REsS I N - E /u 1 17 Y (lMSib:D iT E P AT iou A 0 F. E 5 5 u R E. ST RE s 5 =(5.5 Ks) 1 G00 P S i

= H. O KSI 2200 PSI T H E R m A L E XPA M sio u ST R E S S = '7. 9 % KS)

THERMAL ST R E S S = (19 9. 9 K S il _59 9 F - G o F

= ) 5 G. 7_9 K S 1 550 F-Go F TOTAL STRESSt M.o + 'l.34 + 15G.2 % * \\G 8.18 k Sl ITGR ATiou 6 r

PRES $v&E ST R ES S Cl.15 k sil lioo PS I

2. '1 5 k S i

=

'7 o o PSI THERMAL ENP A N Sio M ST R E S S 5 O.O kSi TH E R rn A L si RE 5 5 =( C).G K S \\) 5 58 2 F - H o Fl = - 10. S I K5I boo

  • P - H o
  • F )

TOTAL ST R E S S = 2.15 t o.O - 10. 8 l

'8 OG kSI w0h o1,5/26/c9-eieriero av

91. t. e oc c. ~o.

885

{/X/#2 52 c: C5 o.re

    • oe no.

neviewio av l

Babcock &Wilcox 82-1134218-00 Nuclear Pow r G e tion Division GENERAL CALCULATIONS THE

~0 v.c wira G C A t. t u L A T I o N E APPL 7 Tc THE IO C.> t L E 5 OF HP_

rr\\ A N U A L A tv u A T lo N. T w o cases WILL GE AN ALY? E D DU RI N G THE 70 A D D I T i o N A L.

C.Y t L E S ; ONE wtTH THE

/NCLUEloM OF OSE S T F. E S S E S Aub THE o7HER WiTHouT vsE INCLULtou OF T-O G E STR E S S E 5.

THE N u m G. E R OF IOGE STRESS tyt LE E A M A L y ? E. D is G 5 0 - 30 = G 2 0.

C. A S E 'l - P E A k STRESS imTENsiTy R A N C-E w f rH

('

I M c_ L v S t o u OP Z oG E STRE$555 THE m A M m u rY\\

PEAK S T R E S S.

I N T E N S L T 'Y RAMGE r

IS Co m P R t S E D OF (LTE RATto M At O G E) -

1

( I T E Ps A T I O tJ 6 - O G E) \\ A N D CAN ottuR F o ft 30 C_Y t. L E $

TA!$[c=

( \\ 6 9.1 B - 15. '2.4 3 - C - B. o G - t 5. 2 W 3 = 20G.72. k s I l

THE FRtmA RY t S E C.O N D A RY ST R E S S IN TE N St T Y RAMGE ASSoci ATE D W IT H THIS T R A N S I E AJ T S

I S S. S G KS i oGE ]

5 3 5 m.

5 \\. 'S K S \\

E iT e R A i lo N S A

AND G 1

=-

l l

G. c.. % 2 0 h o 1,5l'2.6 / a ?

f r.er. so iv occ so.

885

/M /P.2 SS OF G5 oire

, ci uo.

.eviewio av

Babcock &Wilcox 32-1134218-00 om>ou.' i~n Nuclear Power Generation Division GENERAL CALCULATIONS THEF.EFCRE,Au E L A S lC-P L A L - i t F A TI G U E ANALN EI5 rrcs?

GE P E' R F O R rh E D F o Fs THESE 30 C_y t L E 5.

FoLLowirvG THE P Rote b u R E OF R E F, it 2, P A R A G R A P H F - 10 5. '2.. '), T H E FI M A L PE A K A LT E R VA"T I M G STRESS I M T E /u t l T Y, U'A LT, 1 5 '.

cm

~

V A LT 5 Y '2. K f N g b ra j

wsERE kg= kx -- A C. K x - 1. o 3

~

m S c 2_, _

2 c G '7 2 K S l

1. 4 %

~

K,3

=

5"c ',a' l M 's. 5 G k S 1 A = 0.1 FoR ST A iM L E S S S T E E L. A - 3 'l G, T P 'S i G,

F Ro m F t G. D - 2 O l ( tT, R E F.

  • 2 Kp
t. H H -

O. ') ( ). H M - 1. 0 ) = l.1 5 l

Kc is B E T E R m' NE b FRom FIG F - l O 5 C o.)

F O LLOWIN G 1

is Tsf

c. A t t u t s - r o u o= THE PARAMETERS

/v etE s s n t% y

~~ o O C T A I N K E.

_S-280 19 3.5 G ksl_

=

3 5m 51.3 KSi iGri /C-Q~}e } Q 'o ' ']

is DETERmi NE D

/v E N T '

ni..io.,

'M, t. h 'O h o.r 5 / 2 A / A 2 occ. ~o 285 I/%/f.2 Su OF G S o 1.

c o. ~o

.ev, o.,

6 Babcock &Wilcox 32-11949.18-00 Nuclear Pow r e ation Division GENERAL CALCULATIONS

( C ON D t T ! O /V A% IT E R $ T I O rd M T

C6E I N C i ts F sveFAr_E its.* 9 + t i. 4 i s i%E.w ki t 28.Gs 5, a

9 G,0G 5 -: t

=

/

s

.f_

ov?L ib E Su R F A C. E

-8S 9 5

  • I C.82 s-cs.es w t i 29.G35 x s

-?G.7 65 r S I Qm VI.96 KSI 4

Ob a l o t.1 l KS! w G - O B E)

(COM D IT IOM ib ITE R ATr ou m t i rs e tu o*Atg

-G.2 5, il.a l : - 19. l G n s i

-i3 3, s xs

% 5 est 4

r a(----

~

W u!--

OUTLtbE SURFACE' O.5 5 -ib.82

- 8.27 vs r

-ed.2r 5 ki l H.9MS K5i Q~

=_

91. 8 5 ks i o.'2 9 Qb
41. 9 5 K S I t 101.9 I K Sl Gm t FRorh F t G. P - 10 5 ( c '), R E F, t$ 2,

K e ~~ z.'2. 5 cm U'A c r = Vz. K r k e 5 r G

T4 c-- = V2 ( 1. 9 5 )(.2. 2 5 ) ( i n. 5 G ') = '2 8 2. G '3 KsI Du Rt M G

HE nFC rn A N u A L T-C.T U A T I o es ~ R A N 5 i E N T, THE reir nio v T.C

%dh ore 5/2C/9'2_

occ. ~o.

"885 SY6 Nf2 CR OF C 5 o.ri

. ci ~o neviewie.,

l

Babcock &Wi!c 3 32-1134218-00 Nuclear Power Generation Division GENERAL CALCULATIONS TE rr F E R6Tu F :

vA RIE S Rom 5 9 9 *.:

70 40'F. A T AN AvERP G E T E rn PE R A T u n E of 519*F-uo*P

= So t. 5* F c

2 E = 2 9. 0 5 x l oG PSI E R E F. # 2 ]

G

~

~

Sa = 282 G3 2G w to

= f_7(, G G k5l G

2 9.o 5 s IO F Ro rn F t G. F - \\ o G ( b), R E F. i:* 2,

A/ = A L Lo w A G L E tyC LE 5 = 70 C_y c_ LE 5 Uw = USAGE F A tr o tR = 30

=

0.43 r

no C ASE 2 - PE A K STRESS IM rE Ns rTY PAMGE w erH ou T LOSE S T R E S S E.5 l

THE mAximurY\\ FEAK STR ES S IN TE N & l TY RAN G E is ccmPRisE o CP C iTE R ATiou A) - (IT E R A Ti c A>

CE) 1 AtJD CAM ottuR FOR 40 Cy t L E S.

7Nu'os (16 8,18) - C-8.oG) l9G ZH K51

=

TME P R l rw A R Y t S EC o A> D A R Y STRESS I N T F A/ I I T Y R A NGE A S E C C.i A T E D W ITH THl5 T R A A/ S / E A> 7 15 l 13. 9 4 KSl

'> 3 E m :- 5 l. 3 K S I

[. I T E R A T I c N.S A AND G 1 C6E ].

Peeraneo sv dl.. C. #

oate 5/QC /C O oc c. wo _

[M$

(M b 3G OF C5 eevieweo av omre race so.

Babcock &Wilcox 3 2 - 113 4 218 - 0 0 om>= > -

Nuclear Pow r b e tion Division GENERAL CALCULATIONS TrEREFORE,AM E L A ST IC - P L A S T i C_

P ATIGU E AN ALY 5li cr u s. T SE PE R Po R rn E D i:o R T H E S E 40 C.y t.L E S.

P O L. L o w t M G THE P A O C. E D U R E OF R E F. M 2., P A R A C-R A F H

- \\ 0 5. 2. 9, T H E FINAL PEAN A LT E R M A T r u &

ST RE SS

~

itv7E M tiv y hr, 15~

%-= Y2 Kp Ks S$ag, A (. k x - I. 0 )

WHERE' k J K A-S.P) 196.2.4 Ks1 c

1. 4 ')

k

=

=

c2.

t 5"c'/

I i 'I. 'l V k S I 6

9 =

0. '7 FoR STAIM LE S S STEEL A-3'lG, T P S I G, F R O rw F t G. D - 2 O I C c.), R E F. ti '2_

Kp 5 l.91 + 0.7(.l.4')-f.O) =

1.80 Kg 15 bE T E R rn i N E D FRom F t G. F - 10 5 C a.). F o L Lo wi N G 15 THE C. A L C.v L A T l o N OF THE PA R A rn E T E R E NECEs;FR>

TO 0 6TA l N ks

= _ l l '). ') 4 kSI

=

2. 3 3 l

'3 5 m

51. 3 k 5 l l

Cm b

O rn l 6 'o \\

i$

D E -~ E 8 /n l N E O A/EX T T

s c_. ww

..,, s / u n 1

.cc ~c.

"M$

Edh.//,2

/

M op C,5 o.r.

, o. ~o

. vi.w.o,,

\\

Babcock &Wilcox 32-1134218-00

="'

Nuclear Pow r e tion Division GENERAL CALCULATIONS

( C0 ts D I T I C A) 2 0, = 1 'E Ps b T I O N O

,,,n; voept-til.*9 K S I

'919 "11 9 9 2 7 ril

+

/

/

=

+

/

/

ouTilb E SURF ACE

- 8 4.9 5 K S I l 3. af) gi l

- T i, 7.2 K S f

11. 31 K LI 4 Qm Qb H 10G.62KSl e

( C.O N b l T 101J 2. b : IT E PsA Tlo y G) r; sor suReatE

-G%S KE8 l 8 5 ki'

' O' 4 "I 8

~

N z

r e

M ouTSibE SVRI: ate e.5 5 x :#

i.i S K a 9.9 xs,

~

O-1 s.12 x s i c

o.11

=

Om Gb l's.12. K s i t t o G.67 KS I

+

F R o rrs F i G. P - \\ 0 5 ( od, R E F. 2 2,

Ke 2 I, 7 I

/2 K p K q b(.n )

~

N A LT 5 r0 l

IAL-V2 ( 1. 8 0) (1. 7) ( 119. 9 9 ) = l S 3 2 Ksi

'l S u Ps i lv C-THE HPI M A A/ U A L A c. - v A T 1 o A>

I R A A/ O I E ts T i

l l

Pee,anto sy Q_t hh S/ZG/91 oarg occ, so_

  1. 5 cA6/a s"8 OF G o.1,

...~o_

arviewto er I

r Babcock &Wilcox 32-1134218-00 Pos 2'ou > ' >>

a ucce*=ot' comosa, Nuclear Power Generation Division GENERAL CALCULATIONS Tr E TE n PE P h ivP E u AR\\ E 5 Prom 5 1 'l F

To Ho*F.

AT RN AVERAGE TE rn PE R A'r u R E OP 519*P+ 40*P

= 3 0 9. 5 ' F, 2

E-

2. 9. 0 5 x l o b PSI CREF.M2_]

5x= 183.2 ksi 2 G x io G esi I "1 G. 1 Ksi

-=

2 9.o ss. 10 6 Ps i P Rom fig F - t o G C b '), R E F. W 9_,

N

  • A L'4 w A IS L E tyc.LES = 230 c_y c L E s, U s = U S A C-E F A CT o R 5 40 0.17

=

230 l

l

{

l l

l

~.

PREPAREo BY h -C.

DAtt b/br /

DO C. NO.

o

(( I&

SY M

8 OF G R oire Pace NO.

neviewco av

/

O Babcock &Wilcox 3 2 - 11 R 4 2.18 - 0 0

'="'au'i Nuclear Power Generation Division GENERAL CALCULATIONS CGE FEA K 1-PESS iUTENSITv R A /V G E THIS FsA N G E IS C.o m P R l S E D OF L OG E AND C A IV cttv R FOR THE RE MAIN DE R OF TME E ARTH ou A K E ty c LE S, T H A T IS, G 50 - 3 0 = G 2 O C y c_ L E S.

7 N 7cre :

( 1 5. 2 9 ) - (- l 5. '2 4 )

3 0. H 8 K S I

=

THE prima RY + S Eco N D A RY stress lurenslTy RANGE A Slott ATG b W TH THIS T R A A> I I E /V T l5

23. 8 2 K 1 I 4 3 5 m. = E 1. '3 K S 1.

TH E R E F O R E, TH E PIMAL PEAN A LTE R M ATI N G S TR E S.5 I N TE N S I T )'

RANGE, % r, (5:

t l

hr V2.7Nu[c

)/2 (30. 4 8 K S l) =

15. 2 H ksI 5

s THE F Em AINDE R OF THE EARTHGUAKE C.Y C L G S o ttu R DuRIN G S TE A b y - S TA T E CoNDITiouS.

i THE RE FOR E, AT A T E rh P E R A T U R E OF 5 So c F, E = 2 5. 9 5 = 10 6 PSI C. R e p. it. 2 ]

reer. co.,

'll.t. h o.re S /2C / 6'2.

oc c. ~o.

88$

o.re f/26,NP.2.

GC OF G5

.evieweo

.1

,.ce uo.

Babcock &Wilcox 32-1134218-00

' o s a" ' <""

Nuclear Power G e ation Division GENERAL CALCULATIONS is 24 xs,

2cx io sps,'

sm i s.s 3 ns,

=

=

2 5.1 S wlo s eg, P P, o rn F I G. F - 10 6 ( b ), R E F. H G,

NO A LLo w A 6LE C.'y C L E S = CO VG = U S A C-E F A C.To R = G 2 O_

= 0,O co e

PetPatto ey k,(.

oATE 5/26 /92 oc c. ~o.

888 fM6 EI oF GS neviewie,

o ri Pice ~o

Babcock &Wilcox 32-1134218-00 Nuclear Pow r b e ation Division GENERAL CALCULATIONS il.O TOTAL USAC-E i: A C_T O R V e USAGE F A CTo R FoR 40 RAF)D DE P R E 5 50 R11ATioN i

TRANElEUT t y t L E E = 0. ) G U2 USAGE F A CT O R FOR Ho TEST T8 ANs t ENT CYCLES =

0.04 Us s USAGE F A CT o R FoR 24 0- 9 0 = 2 0 0 H E AT-U P AND too L-Do w M TR A N E l EN T C Y C. L E C = 0. 0 U+-

UEAGE FACTOR FOR So HP m ANv AL A CTU A T l o u CYC LE S C WIT H INC L U S t o v OF I o8E ST R E S S E S 3

=

0. 9 '3 Us=

USAGE FACTOR FoR 90 HP2 MANUAL ACTU A Tr o M CYCLE S

( WITH o u T

/N C LU S t o u OF IoeE STRESSEL) =

0.17 U G =-

USAGE FACTOR FOR R E rn A i N D E R OF 10 BE STRESS CY C L EI =

0. 0 b/2b/ 2 PetPAstD

.7 k,b-D ATE DOC.NO 888 5-4 M2-G2 OF GE DA,,

, A O, ~O

.eviewiD.,

0 Babcock &Wilcox 3 2 - 1 l '3 L 0.18 - 0 0

...c....,.......

Nuclear Power Generation Division GENERAL CALCULATIONS b

eoTAL USAC-E F A C.T o R V=

V,t U z'm Us+

Us*Us+ Us V=

0. l G + o, o M +
o. o

+ o,iq '3, r o, g7 + o, o

o. 8 0 c 1 U

=

B l

b[b PetPAago my $, C, DATE coc. No.

8 83 fMN//.2 C S OF C o

1,

,*oe so.

aiviewio av l

Babcock &Wilcox 32-1194218-00 Nuclear Power Generation Division GENERAL CALCULATIONS 12.0 References 1.

FCA No. 04-3370-00 for Reactor Coolant Piping, B&W Contract No. 620-0011-50.

2.

Nuclear Power Piping Code, USAS B31.7 - Draft 1968.

3.

B&W Computer Code P91167 4.

B&W Computer Code P91206

~

5.

B&W Computer Code P91032 6.

B&W Stress Report, "Therinal Mechanical Analysis of 2 Sch 160 Make-up and High Pressure Injection Nozzle", Design Report #5, for Sacramento Utility District, B&W Contract No. 620-0011-50, Rev. #1, 1/15/73.

7.

Microfiche AC3IMJU, " Reactor Trip w/HPI Nozzle".

(attached) 8.

B&W Document No. 86-1131770-00, " Revision to Funct. Spec.-SMUD",

March 3, 1982.

9.

B&W Document No. 86-1131765-00, "HPI Nozzle Maximum Flow Rate",

March 3, 1982.

10. JAKOB + Hawkins, " Elements of heat transfer", 3rd Edition, Wiley and Sons, Inc.,1957.
11. B&W Computer Code P91232.
12. B&W Drawing No. 143509E, Rev. 8 " Assembly and Detail for 2-h"

" Pressure Injection Nozzle" l

l s

P4f PAGED BY k,C-5/2C / A 2 oAft occ. so.

N N _3

.2/e, J

GH OF G 5 noviewco av

,,oe so oare

Babcock &Wijco3 32-1194218-00'"'"'""

Nuclear Power Generation Division GENERAL CALCULATIONS R E F E R E N C.E mitRO F1CH E A C_'ST.m 3 U l

l l

4

% g PtePAR:0 eY bd.C.

h#b b/b DAfe DOC.No kMS o fe 6N2% /1

/

G C. OF GS neviewro er P*oe ~o

.

Retrieved from "https://kanterella.com/w/index.php?title=ML20054M221&oldid=4490953"