· 2017. 2. 28. · '0PACIFIC.. GAB A.KD ELECTRIC CGMPANT 77 8EALE STREET, 31ST FLOOR ~ SAN FRANCISCO, CALIFORNIA 94106 ~ (415) 781.i211 LIOHN C. MORRISSt Y VICC OSSSICSST AIIC 0CMSSAL

'0 PACIFIC.. GAB A.KD ELECTRIC CGMPANT77 8EALE STREET, 31ST FLOOR ~ SAN FRANCISCO, CALIFORNIA 94106 ~ (415) 781.i211

LIOHN C. MORRISSt YVICC OSSSICSST AIIC 0CMSSAL COIINSSL

MALCOLM

M H. IIUR6U6HASCOCIASC St%SEAL COUDSSL

CHARLES T. VAN CCV BC N

PHILIP A, C RAN Lr~ J R ~

ROB LsRT CHL BACHs . sess l'llgDANICL C ~ GIBBON

IIIIIII»ICI»Isis CCV»ICS

September 28, 1979

E

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COUSLAS A, Coltsst~ a»IOS

Mr. John F. Stolz, ChiefLight Water Reactors Branch No. 1Division of Pxoject ManagementU. S.. Nuclear Regulatory CommissionWashington,'. C. 20555

Re: Docket No. 50-275Docket No. 50-323Diablo Canyon Units 1 S 2

g p5 3+30 R'vB.R~QCt iG'ib05cig

Dear Mr. Stolz:

I The enclosed material is submitted in responseto several Staff questions.

s

Five advance copies of this material have beensent'directly to Mr. Bart Buckley.

Kindly acknowledge receipt of the above mate-rial on the enclosed copy of this letter and xeturn itto me in the enclosed addressed envelope.

Very truly yours,

Enclosures

NIIliI)p~pfm OgdenAl-,l~"

ss,

r(~ l4

l.4 J

~a% i

Response to Several

1. uestion 212.1 - LOCA Anal sis with Metal/Water Correction-NRC Letter Au ust 3 1979

The previous large break analysis submitted for the Diablo CanyonUnits on December 7 and 19, 1979 included a 0.6 DECLG break analysisfor Unit II, and a generic four-loop plant 17X17 three break spectrumof large breaks indicating a worst break size of 0.6 DECLG. Theconcept of the generic spectrum was to reduce the total number ofbreak sizes to be reanalyzed with the February 1978 Evaluation Modelfor al'1 our customers collectively. The 0.6 DECLG break submittedfor Diablo-Canyon Unit II bounded the lower power Unit I plant, sono new analysis for Unit I was done.

As a result of question 212.1 from the NRC letter dated August 3,1979, a new three break spectrum was performed for Unit II includingbreak sizes'. of 1.0, 0.8 and 0.6 DECLG breaks. Input changes fromthe previous analysis of Unit II were included that represent improve-ments to the analytical rep'resentation of the plant. The input valuesfor this analysis are in conformance with currently approved methodsof code input calculations, and are consistent with other plant AppendixK analyses. This spectrum resulted in the 0.8 DECLG being the limitingbreak size. A peak clad temperature of 2187 F at a peaking factor of2.32 resulted.

i

A single break analysis for Diablo Canyon Unit I was performed assuminga 0.8 DECLG break. A previous Salem Unit II analysis of a spectrum oflarge breaks provide the basis of the worst break being the 0.8 DECLGfor Diablo Canyon Unit I, since these plants are very similar in design.Results of the 0.8 DECLG break for Unit I yielded a peak clad tempera-

. ture of 1930 F at a, total peaking factor of 2.32, illustrating that0

Unit II large break analysis does represent a bounding calculation forboth units.

Attachment 1 is a summary of the reanalysis. It will be filed shortlyas an amendment in the FSAR.

2. uestion 231.1 - Fission Gas Release Model - Au ust 2 1979

'An analysis has been performed for Diablo Canyon Units I and II withthe NRC approved Westinghouse fuel performance model, PAD 3.3, (WCAP-8720). ,The results show that all applicable fuel rod design criteriaand, in particular, the current rod internal pressure criterion are met.

j%

The effects of the restrictions on the PAD 3.3 Code given in the letterdated February 9, 1979, from J. Stolz, NRC, to T. Anderson, Westinghouse,were assessed. It was determined that the restrictions on the use ofthe code curing normal operation had no effect on the safety analysespresented in the FSAR. There was also a restri,ction on the use of thecode to analyze transients. An analysis was conducted which conser-vatively bounded transient phenomena. Based on this analysis, it wasconcluded that there is no effect on the results of the safety analysesdiscussed in Chapter 4 of the FSAR.

3. uestion 231.2 - Internal Fuel Rod Gas Pressure - Au ust 2 1979

The fuel rod design criteria for Diablo Canyon Units I and II haveincorporated the current fuel rod internal pressure criterion (referencedin WCAP-8963 as mo'dified by the NRC) as documented in Section 4.2.1.3of the Diablo Canyon FSAR. The criterion is that the internal pressureof the lead rod in the reactor will be limited to a value below thatwhich could cause the diametral gap to increase due to outward creepduring steady-state operation and which could cause extensive DNB

propagation to occur. As shown in the report "Safety Analysis for theRevised Fuel Rod Internal Pressure Design Basis," transmitted by theWestinghouse (C. Eicheldinger) to NRC (D. F. Ross, ~ Jr.) letter, SerialNo. NS-CE-1290, dated November 24, 1976, no reanalysis of the DiabloCanyon 1 and 2 FSAR transients is required since the criterion has beenmet with margin. The cited analysis was resubmitted to the NRC asWestinghouse topical report WCAP-8963 by Westinghouse'etter, SerialNo. NS;CE-1391, dated March 31, 1979.

4. uestion 231.3 - Potential Grid Stra Dama e - Au ust 2 1979

Evidence from other 17X17 refuelings indicated that the potential fordamage can be minimized or eliminated by exercising care during the'handling operations. For practical, commercial, and safety reasons weintend to include in our fuel handling operations all precautions thatmay protect the fuels integrity. These will include proper training ofoperators, confirmation of proper functioning and alignment of the fuelhandling and transfer equipment and implementation of appropriate fuelhandling precautions and recommendations recently provided by Westing-house on August 15, 1979. These precautions and recommendations arebased upon their assessment of the Salem experience.

5. uestion 1 - Sin le Rod Dro - Au ust 2 1979

Plant Technical Specifications have been revised so that negative fluxrate trip setpoints are decreased from 5 percent to 3 percent and therate-lag time constant are decreased from 2 seconds to one second. Acopy of the Technical Specification page2-5 is included as Attachment 2.

AOl ~\

2

Informal uestion - Heat Transfer Anal sis'Lv',The Staff r'equested that we su'pply, for heat transfer analysis pur-poses, sectional diagrams of the Limitorque valve motor operatorand the Fan Cooler 'Case. Attached are the sectional diagrams of thevalve motor operator (Attachment 3) and the fan cooler 'case (Attach-ment 4) with the physical dimensions and properties as used in theequipment model. The analysis of the fan cooler case results in atemperature boundary condition which is then used to determine theresponse of the motor windings.

7. Informal uestion - Debris Screen Modification for Containment Pur e Valves

We included as Attachment B to our September 27, 1978 letter drawingsof the Debris Guard for the Containment Purge Valves. After the Staffrequested that the debris guard be strengthened, the screens were re-designed as shown in Attachment 5. Installation of the new design iscomplete.

8. Informal uestion - "Rise Time" - A arent Discre anc

0

The Staff questioned the apparent discrepancy between the "rise times"shown in two of our submittals for the environmental qualificationsof the Sostman RTD's. The intent of Figure 3 of our September 21, 1978proprietary submittal was to demonstrate long term qualification of theSostman RTD's. Therefore the "rise time" was not considered importantfor showing the long term (excess of 24 hours) test conditions and wasn'ot accurately represented. The rapid "rise time" is correctly repre-s'ented in Figure 4 of our June 14, 1979 proprietary submittal. A moreexact! plot of the "rise time" is contained in Figure 6-3 of WCAP-9157,"Environmental Qualification of Safety Related Class IE Process Instru»mentation" September 1977.

ATTACHMENT 1

LOCA REANALYSIS TO EVALUATE ECCS PERFORMANCE OF THE

DIABLO CANYON NUCLEAR PLANT - UNITS 1 AND 2

MAJOR'REACTOR COOLANT SYSTEM PIPE RUPTURES

(LOSS"'OF"„COOLANT ACCIDENT)It 4 lt 4

I

The analysis specified by 10 CFR Part 50.46(1) Acceptance Criteria forI,

Emergency Core Cooling Systems, for Light Mater Cooled Nuclear Power

Reactors is presented in this section. The results of the LOCA analyses

are shown in Tables 2 and 2a -and show compliance with the Acceptance

Criteria. The analytical'echniques used are in compl-iance with Appen-'t

dix K of 10 CFR'Part 50,„and 'are described 'i'n Referenc'e,(2) ~

I t

t'houlda major break occur, 'epressurization of ..the'eactor coolantsystem results in a pressure 'decrease in the pressurizer.'eactor tripsignal occurs. when the pressurizer low-pressure trip set point isreached. A safety injection system signal is actuated- when the appro-priate set point is reached. These countermeasures will limit the con-sequences of.,t:he accident in,'two ways:

I

Reactor trip and borated water injection complement void formationin causing rapid reduction of power to a residual level correspond-

ing to fission product decay heat.

i2. Injection of borated water provides heat transfer from the core and

prevents excessive clad temperatures.

At the beginning of .the blowdown phase, the entire reactor coolant sys-tem contains subcooled liquid which transfers heat from the core by'.

forced convection with some fully, developed nucleate boiling. After thebreak develops, the time to departure from nucleate boiling is calcu-lated, consistent with Appendix K of 10 CFR Part 50. Thereafter, thecore heat transfer is based on local conditions with transition boilingand forced convection to steam as the major heat transfer

mechanisms'uring

the refill period rod-to-rod radiation is the only heat transfermechanism.

L

r

When the reactor coolant system, pressure. falls below 600 psia, the accu"

mulators begin to inject borated water. The conservative assumption ismade that accumulator water injected bypasses the core and goes out

through the break until the termination of bypass. The conservatism isagain consistent with App'endix K of 10 CFR Part 50.

Thermal Anal sis

Westinghouse Performance Criteria for Emergency Core Cooling System

P

The reactor is designed to withstand thermal effects caused by a LOCA>

including the double-ended severance of the largest reactor coolantsystem pipe. The reactor core and internals, together with the emer-

gency core cooling system, are designed so that the reactor can be'p

safely shut down and the essential heat transfer geometry of the corepreserved following 'the accident..

The emergency core cooling system, even when operating during the injec-tion mode with the most severe single active failure, is designed tomeet the Acceptance Criteria.

I

Method of'Thermal 'Analysis

The description of the various aspects of the LOCA analysis is given inReference (2). This document describes the major phenomena modeled, theinterfaces among the compu'er codes, and features of the codes whichmaintain compliance with the Acceptance Criteria. The individual codes

are described in detail in References (3) through (6).

The analysis presented here was performed using the February, 1978 ver-sion of the Westinghouse Evaluation Model. This version includes the

modifications to the models, referenced above, as specified by t'e NRC

in Reference (7) and complies with Appendix K of 10 CFR Part 50. The. February, 1978 Westinghouse Evaluation Model is documented in References(8) through (10) an'd (14).

-2-

I

The analyses were performed using an upper head fluid temperature equalto the hot,leg temperature. The effect of using the hot leg temperaturein the reactor vessel upper head region is described in Reference (13) ~

A reference 3«break spectrum analysis was performed for Diablo Canyon

Unit 2, the unit with the higher power rating. Both units employ 17x17

fuel assemblies and have containments with similar internal steel and

concrete structural heatsinks'he

containment backpressure is calculated using the methods and assump-

tions described in Reference (2), Appendix A. Input parameters used forthe Diablo Canyon analyses are presented in Table.3.

The containment initial conditions of 90oF and 14.7 psia are represen-tatively low values anticipated during normal full-power operation. The

initial relative humidity was conservatively assumed to be 98.8 percent.

Results

Table 2 presents the peak clad temperatures, hot spot metal-water reac-tion, and other key results for a range of break sizes for Diablo CanyonUnit 2. The range of break sizes was determined to include the limitingcase for peak clad temperature (PCT) from sensitivity studies reportedin References (ll) and (12) ~ Results obtained show the limiting breakto be the DECLG, C

D ~ 0.8. Table 2a presents the samg parameters forthe limitin case DECLG C

D ~ 0.8) break for Diablo Canyon Unit 1.The low PCT value computed for Unit 1 makes it apparent that Unit 2

results are bounding. The peak linear power and core power used in theanalyses are given in Tables 2 and 2a. Since there is margin between

the value of the peak linear 'power density used in this analysis and thevalue expected in operation, a lower peak clad temperature would be

obtained by using the peak linear power density expected duringoperation.

For the results discussed below, the hot spot is defined to be the loca-tion of maximum peak clad temperature. This location is given in Tables2 and 2a for each break si:ze analyzed.

-3-

OC

C

Figures la through $ 6a and lb through 16b represent the transients for the prin-ci'pal parameters .for., the break sizes analyzed. The following items are noted:

~ ~ ~.f+"4 ~'", 30».The following quantities are presented at the clad burst.ib thw ~" .;location and at the hot spot (location of maximum clad14 thru 34 .,'temperature). both on the hottest fuel rod (hot rod):

1 i Pluid quality2. Mass velocity3. Heat transfer coefficient

The heat transfer coefficient shown; is calculated by the

LOCTA IV Code.

Z

F>&reS 48 thru 6a, The system pressure shown is the calculated pressure in the4b thru 6b, core. The flow rate out the break is plotted as the sum

44 thru 64 of both ends for the guillotine break cases. This core

pressure drop shown is from the lower plenum, near thecore, to the upper plenum at the core outlet.

I 4

~~gureS 4 thru 98,: These figures show the hot spot clad temperature transient

7c thru 9c,.'and the clad temperature transient at the burst location.

74 thru 94 'he fluid temperature shown is also for the hot spot and

burst location. The core flow (top and bottom) is alsoshown+

I

1

g r. I ~ These figures present the core ref lood transient.

10b thru Ilb,-'0c thru llc,,„~ 104 thru 114

F 'gure J2 th .13 These figures show the emergency core cooling sys «m flowfor all cases analyzed. As described earlier, the accumu-

12c thru 13c,134 lator delivery during blowdown is discarded until the end

of bypass is calculated. Accumulator flow, however, isestablished in refill-reflood calculations. The accumula-

tor flow assumed is the sum of that injected in the intactcold legs.

-4

Ej >x

0

y

Figures 14a These figures show the containment pressure transient.through 14d

Figures 15a These figures show the core power transient.through 15d

Figure 16 This figure shows the break energy released to the con-

tainment during blowdown for the limiting case.

Figure 17 This figure provides the containment wall condensing heat

transfer coefficient for the limiting case.

k

In addition to the above, Tables 4 and 5 present the ref lood mass and

energy releases to the containment and the broken loop accumulator mass

and energy flowrate to the containment, respectively for the DiabloCanyon Unit 2 limiting break. The clad temperature analysis is based on

a total peaking factor of 2.32. The hot spot metal water reactionreached is 7.5 percent, which is well below the embrittlement limit of17 percent, as required by 10 CFR Part 50.46. In addition, the totalcore metal water reaction is less than 0.3 percent, for all breaks as

compared with the 1 percent criterion of 10 CFR Part 50.46.

The results of several sensitivity studies are reported in Reference

(12)- These results are for conditions which are not limiting in naturel

and hence are reported on a generic basis.

Conclusions —. Ther'mal Analysis

For breaks up to and including the double-ended severance of a reactor„coolant pipe, the emergency core cooling system will meet the Acceptance

Criteria as presented in 10 CFR Part 50.46. That is :

1. The calculated peak fuel element clad temperature provides margin tothe requirement of 2200 F.

2. The amount of fuel element cladding that reacts chemically withwater or steam does not exceed 1 percent of the total amou". oF

zircaloy in the reactor.

3. The clad temperature transient is terminated at a time when the core

geometry is still amenable to cooling. The cladding oxidationlimits of 17 percent are not exceeded during or after quenching.

The core temperature is reduced and decay heat is removed for an

extended period of time, as required by the long-lived radioactivityremaining in the core.

The time sequence of events for all breaks analyzed is shown in Tables 1

and la.

References

l. "Acceptance Criteria for Emergency Core Cooling Systems for LightWater Cooled Nuclear Power Reactors," 10 CFR Part 50.46 and Appendix

K of 10 CFR Part 50. Federal Register, Volume 39, Number 3, January

4, 1974.

2. Bordelon, F. M., Massie, H. W. and Zordan T. A., "Westinghouse ECCS

Evaluation Model - Summary," WCAP-8339, July, 1974.

I

3. Bordelon, F. M., et al., "SATAN-VI Program: Comprehensive Space-

Time Dependent Analysis of Loss of Coolant," WCAP-8302, June, 1974

(Proprietary) and WCAP-8306, June, 1974 (Non-Proprietary).

4. Bordelon, F'. M., et al., "LOCTA-IV Program. Loss of Coolant Tran-

sient Analysis," WCAP-8301, June, 1974 (Proprietary) and WCAP-8305>

June, 1974 (Non-Proprietary) ~

5. Kelly, R. D., et al., "Calculational Model for Core Ref looding aftera Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170, June, 1974

(Proprietary) and WCAP-8171, June, 1974 (Non-Proprietary).

6. Bordelon, F. M. and Murphy, E. T., "Containment Pressure AnalysisCode (COCO)," WCAP&327, June, 1974 (Proprietary) and WCAP-8326,

June, 1974 (Non-Proprietary) ~

-6-

7. "Supplement to the Status Report by the Directorate of Licensing inthe matter of Westinghouse Electric Company ECCS Evaluation Model

Conformance to 10 CFR Part 50 Appendix K," Federal Register,November, 1974.

8; Bordelon, F. M., et al., "Westinghouse ECCS Evaluation Model — Sup-

plementary Information," WCAP-8471-P-A, June, 1975, (Proprietary)and WCAP-8472-A, June, 1975 (Non-Proprietary) .

9. "Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-

8522, November 1975 (Proprietary) and WCAP-8523, November 1975 (Non-Proprietary).

10. Letter from C. Eicheldinger of Westinghouse Electric Corporation toD. B. Vassallo of the Nuclear Regulatory Commission, Letter Number

NS-CE-924, Dated January 23, 1976.

ll. Johnson, W. J., Massie, H. W. and Thompson, C. M., "WestinghouseECCS - Four Loop Plant (17xl7) Sensitivity Studies," WCAP-8565-P,

July, 1975 (Proprietary) and" WCAP-8566, July, 1975 (Non-Proprietary) ~

12. Salvatori, R., "Westinghouse ECCS —Plant Sensitivity Studies,"WCAP-8340, July, 1974 (Proprietary) and WCAP-8356, July, 1974 (Non-Proprietary) ~

I13. Letter fran C. Eicheldinger of Westinghouse Electric Corporation to

V. Stello of the Nuclear Regulatory Commission, Letter No. NS-CE-

1163 dated August 13, 1976.

14. "Westinghouse ECCS Evaluation Model - February 1978 Version," WCAP-

9220-P-A, February, 1978 (Proprietary), and WCAP-9221-A, February,1978 (Non-Proprietary) ~

-7-

9

0TABIZ 1

LARGE BREAK

TIME SEQUENCE OF EVENTS

DIABLO CANYON UNIT 2

DECL C 1 0

(Sec) .

DECr. C .- 0.8(Sec)

DECL C ~ 0.6(Sec)

START 0.0 0.0 0.0

Rx Trip Signal 1 65 1. 65 1.66

S I Signal 0. 84 0 91 1. 03

Acc Injection- 13. 6 14.3 16.5

End of Blardown 28.38 28. 72 30. 63

Bottom of Core Recovery -39-19 38. 79 41.27

Acc. Empty 48.27 48. 79 50. 87

Pump Injection .25 84 25. 91 26. 03

End of Bypass 25. 62 25 46 - 27.63

a ~ 'l

TABLE la

LARGE BREAK

DIABI,O CANYON UNIT f

TIME SEQUENCE OF EVENTS

DECL CD 0 8

(Sec)

t

START 0.0

Rx Trip Signal 1.66

S. I. Signal 0.94

Acc. Injection 14.6

End of Blowdown 29.38

IBo t tom bf Core Re co very 39.52

Acc. Empty 49.15

Pump Injection 25. 94

End of Bypass'6.11

TABLE 2


DECL C - 1.0 DECL C ~ 0.8r

DECL C ~ 0.6

Results

Peak Clad Temp. F

Peak Clad Location Ft.Local Zr/H20 Rxn(max)X

Local Zr/820 Location Ft.Total Zr/H20 Rxn X

Hot Rod Burst Time sec

Hot Rod Burst Location Rt.

2006

7.5

4. 55,

6.0

<0. 3

28. 8

6.0

2187'.= 6.0

7;5—. '6.0

-<0. 3

26. 9

.-6.0

2003

7.5-v' ~ 23

—. 6.'0

. <0.3

29.4

6.0

Calculation

NSSS Power Mwt 102X ofPeak Linear- Power kw/ft 102X ofPeaking .Factor (At License Rating)Accumulator Mater Volume

3411

12.63

2-.32

- - 850 ft,/tank

7

TABLE 2a

LARGE BREAK DIABLO CANYON UNIT

DECL CD ~ 0

Results

Peak Clad Temp. oF

'eak Clad Location Ft.Local Zr/H20 Rxn(max)%Local Zr/H20 Location Ft.Total Zr/H20 R

Hot Rod Burst Time sec

Hot Rod Burst Location Ft.

1930

7.5

2. 87

6.0,<0.329.06.0

Calculation

0'SSS Power Mwt 102% ofPeak Linear Power kw/f't 102% ofPeaking Factor (At License Rating)Accumulator Water Volume

3338

12.36

2.32

850 ft3/tank

r-"e

TABLE 3

CONTAINMENT DATA (DRY CONTAINMENT)

NET FREE VOLUME 2,660,000 Ft3

INITIALCONDITIONS

Pressure

Temperature

RWST Temperature

Service Water Temperature

Outside'Temperature

14.7 psia90 oF

47 oF

52 oF

37 oF

SPRAY SYSTEM

Number. of Pumps OperatingRunout Flow Rate

1

Actuation Time

3460 gpm

12

SAFEGUARDS FAN COOLERS

Number of Fan Coolers- OperatingFastests Post Accident Initiation of Fan Coolers 20 secs.

STRUCTURAL HEAT SINKS

Thickness (D:j Area (ft2)

~ 0.375 Steel48.0 Concrete

12.0 Concrete

2.75 Steel0.75 Steel

'.5

Steel0.438 Steel0.375 Steel0.2496 Steel0.2 Steel

~ r

110,520

51,524 .

3,360

„25,200

31,440~ . 14,160

44,"400

11,400

5,190

~C7

TABLE 3 (Continued)

CONTAINMENT.DATA (DRY CONTAINMENT)

0.1872 Steel0.15 Steel0.1248 Steel0.0936 Steel0.0624 Steel0.02 Steel0.5 S.S.

0.1872 SoS.

0.438 S.S.

0.2 S.S.

12.0 Concrete

60,600

12,552

20,520

21,600

10,800

33,075

10,958

1,595

1,080

1~572

18,630

Cj.

(i

TABLE 4.

REPLOOD MASS AND ENERGY RELEASES

DIABLO CANYON UNIT 2 DECLG~ CD ,0

TIME

(sec)M(TOTAL)

(LBm/sec)

Mh(TOTAL)

(BTU/sec)

38.795

45.183

53.633

66. 333

81.633

98.533

116.733

156.833

202.433

'318.533

"404.433 „-

0.

38.51

157.39

322.09

375.99

387.39

393.63

404.17

414.49

438; 17

456.69

0.

4.99 + 4

1.62 + 5

2.09 + 5

2.18 + 5

2.14 + 5

2.10 + 5

1.96 + 5

1.79 + 5

1.44 + 5

1.24 + 5

i;7 1

TABLE 5

BROKEN LOOP INJECTION SPILL DURING BLOWDOWN

DIABLO CANYON UNIT 2 DECLG~ CD'

TIME MASS

0.000

1.010

2.010

3.010

4.010

5.0106.010

7.010

8.010

9.010

10.010

11.010

12. 010

13.010

14.010

15.010

16.010

17.010

18.010

19.010.

20 010

21.010

22. 010

23.010

24.010

25+569

25.670

2967.550

2769.743

2608.921

2474.110

. 2359.115

2259.195

2170.903

2091.643

2019.910

1954.487

1894.476.

1839.371'788.442

'1741.911

1698.918

1658.815

1621.331

1587.059

1555.692

1526.385

1498.960

1473.324

1449.107

1426.142

1404.317

1382.406

1380'.438

172117.918

160645.122

151317.426

$ 43498 ~ 354

136828.673

131033.290

125912.357

121315.292

117154 792

113360.240

109879.607

106683.528

103729.658

101030.839

98537.247

96211.252

94037. 173I

92049.431

90230;143

88530.319

86939.680

85452.766

84048.218

82716.252

81450.404

80179.532

80065.398

TABLE 5 (Continued)

BROKEN LOOP INJECTION SPILL DURING BLOWDOWN

DIABLO CANYON UNIT 2 DECLG, CD ~ 0 8

J25 ~ 870

26.061

26.461

26.861

27.261

27.661

28.061

28.462

28.670

29.67

1376.425

1372.630

1364.765

1357.035

1349.416

1341.929

1334.544

1327.254

1323.501

~ 1307 '

0

79832.621

79612.530

79156.372

,78708.053

78266.156

77831.859

77403.547

76980.741

'76763.069

75852.

~ '

T 1000

i.8503

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TABLE 2.2-1

REACTOR TRIP SYSTEM INSTRUMENTATION-TRIP SETPOINTS .:. ".

TRIP SETPOINT. 'LLOWABLEVALUES'I

Not Applicable Not Applicable

*Low Setpoint -. < 25K of RATED .... —.... Low Setpoint - < [email protected] RATED, ....THERMAL POWER .. -':;" '-.. THERMAL POWER .

High Setpoint - < 109K of RATED .;..-- 'igh Setpoint - < llOX of RATED~ '-. THERMAL POWER . THERMAL POWER.

< 5X of RATED THERMAL POWER with;- -:.. = < 5.5X of RATED THERMAL POWER

a time constant > 1 second '.=.::. 'ith a time constant > 1 second

FUNCTIONAL UNIT

1. Manual Reactor Trip :

2. Power Range, Neutron Flux

3. Power Range, Neutron Flux,High Positive Rate

4. Power Range, Neutron Flux, . < 3X of RATED THERMAL POWER with". ':., < 3.5X of RATED THERMALPDWER. High Negative Rate . a time constant > 1 second . : ', ':.:, with a time constant > 1 second

5. Intermediate Range, NeutronFlux

6. Source Range, Neutron Flux

7. Overtemperature bT

8. Overpower bT

10. Pr essurizer Pressure —High

11. Pressurizer Water Level —High

12. Loss of Flow - + 90K of design flowper loop~

"Design flow is 87,700 gpm per loop.

, < 25K of RATED THERMAL POWER -': < 30K of RATED THERMAL POWER .

~

'10 counts per second ,--,=..-- "- < 1.3 x 10 counts per second

( ~, ~-

See Note 1 " ' '"-. ~ '=:-„'-.:.~ See Note 3

'ee Note 2;-.', .".:..':-':;-::..: See Note 3:. - ~;

9. Pressurizer Pressure""Low .'.

> 1950 psig .: . :::;- . ,';: -.-.... '.. > 1940 psig -.

' 2385 psig ';..:'..':,'-::.'.=. ". '...'.- ' 2395 psi.g

< 92K of instrument span '':--'-:-.,':..:=-,".-. < 93X of instrument .span"-"."-,'=;-',".'

... ':-;:::-":. ':-'::. > 89K of design flow'..—,'.-: =:".'-" per loop

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