ML19319D854

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Startup Rept, Suppl 9
ML19319D854
Person / Time
Site: Rancho Seco
Issue date: 08/16/1974
From:
SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
References
NUDOCS 8003260806
Download: ML19319D854 (18)
v • d • e


Text

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~...:. n p..,,;b SUPPLEMEUTAL REPORT NO. 9 TO THE 5,, ; ;, - " r,5

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RANCHO SECO UNIT 0%E STARTUP REPORT

-I.

TURBINE / REACTOR TRIP TEST FROM 100% FULL POhTR 1)-

-PURPOSE The purpose of the turbine / reactor trip test from 100% full power was to verify the plant could safely withstand'a trip of both the turbine and the reactor / turbine.

2)~

TEST METHOD The test was divided into 2 separate categories.

A)

The turbine trip from 100% power was a manual trip from the Control Room on March 11, 1977.

B)

The reactor trip frbm 100% power was accomplished dur_ng an inadvertant reactor trip.due to the loss of-both main feed pumps.

3)

TEST RESULTS A)

Turbinn trip from 100% power it.

E' The test was initiated with the manual trip push button of the turbine. This mode causes throttle-stop valves, governor valves,

)

reheat-stop and interceptor valves to trip closed.

The reactor immediately began its runback, tracking unit load i

demand at the rate of.20%/ minute.

RCS Tave began to increase due to the decrease in feedwater flow. ECS pressure increased rapidly l

in response to the increase in Tave.

The generator breakers remained closed, keeping the generator This synchronized to the electric power grid for 45 seconds.

reduces the acceleration transient'seen by the turbine / generator.

The steam relief capability of the plant includes 2 turbine bypass valves, 3 atmospheric exhaust valves, and 9 code safety valves on each of 2 steam headers.

During the transient, the. turbine header pressure reached only 1025 l

psig..The stean generator outlet pressures were 1065 psig and 1055 psig,,respectively. The steam was relieved by the four turbine bypass valves, the 6 atmospheric exhaust valves, and 8 of the 18 code safety valves..(The settings of the code safety valves are:

four;at 1050 psi, four at 1070 psi, six at 1090' psi and four at 1102.5 psi.)

Steam pressure gradually decreased to setpoint (890.psig) at time 220 seconds.

Feedwater flow tracked down from 6

6 lbm/hr at-time 240 seconds. Reactor 11.8 X 10 -1bm/hr to:2.7 X 10

coolant pressure and temperature were maintained within acceptable 3
operating limits during the transient.

'8 0 03 260]O $ [

C

-.2-Reactor power decreased to 20%. power and-leveled as the feedwater control transferred to low limit control.

Post transient reactor Leoolant ' average temperature l increased above -its 582*F setpoint to

'approximately 593*F caused by-a-grounded test lead. Transient'

' response'.vas not impaired nor were the test results invalidated.

The results of the Turbine Trip Test are shown in Figures I-1, 2 and 3.

B)

Reactor / Turbine trip.from 100% power LBoth steam generator outlet pressures were slightly greater than 900 psig before the transient.

Feedwater flow was cut off as both feedwater pumps tripped simultaneously.

Reactor coolant temperature increased as the heat sink of the reactor was removed causing reactor coolant pressure to rise rapidly.

.The reactor tripped on high reactor coolant pressure on all 4 RPS channels. The highest-recorded pressure was 2410 psig.

The auxiliary feedwater pumps started and feedwater was automatically injected through the auxiliary feedwater nozzles directly onto the steam generator tubes to cool the reactor coolant system. Pressurizer level increased to high of 205 inches, then decreased to 20 inches as the reactor coolant temperature decreased.

.As the reactor coolant system cooled, the pressure dropped to 1820 psig approximately 6 seconds following the initial reactor trip.

.s The maximum' indicated turbine header pressure was 1020 psi.

Pegging steam worked properly controlling feedwater at approximately 200*F.

-Safety features actuation was not required though high pressure injection was' initiated to prevent a loy level in the pressurizer.

4)

CONCLUSIONS A)'

The plant and all systems can safely withstand a' turbine trip from 100% power.

The' steam relief. system is of adequate capacity to prevent overpressurizing the steam generator. The maximum difference between feedwater temperature and OTSG shell temperature was not exceeded. - No Technical Specification limits were exceeded during this' transient.

The. plant.and all systems can safely withstand a reactor / turbine B);

g trip ~ from 100% power.

No Technical Specification limits were

exceeded during this trannient.

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Pl.AltT RESPOUSE TO TURBINE TRIP TEST AT 100% POWER 4

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I T.

BIOLOGICAL Sil! ELD SURVEY - 1002 FULL POWER The results of the final portion of startup test procedure TP 800-32, the biological shield survey at 100% power, show that in general aose rates are below design levels.

Tables II-1 to 5 give the results of the surveys nade in the reactor building, accompanied by Figures II-l to 5 which show the loca-tions where resdings were taken. Analysis of the survey data showed that at Javeral locations in the reactor building, the dose rates were greater than the zone limit for that 1ccation.

As Table II-6 shous, all dose rates were less than 100 mrom/hr at the locations in question, and in many cases the rate was only_slightly ciive the 15 tren/hr limit.

It uas determined that these values were acceptable an' actually expected in some areas, and that the cause of the high readings was inexact knowledge of radiation conditions during the shield design. The slightly elevated levels will not affect plant operation and vill not significantly increase personnel exposure.

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-TABLE II-1 Y

4 REACTOR BUILDIEC '+60 Elevation

SURVEY

.y DOSE RATE.

.n DOSE RATE TOTAL DOSE RATE 7.ONE LIMIT

_ arcin/hr)_

_((tarcra/b r) '

(mrce/br)

(mrem /hr)

(

LOCATION 1_

. 5 10 15, 15

- 15 25' 15

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11 15 4-

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REACTOR EUILDING'+40' ELENATION

. SURVEY-y-DOSE ~ RATE n' DOSE RATE TOTAL ' DOSE RATE ZONE LIMIT LOCATION.

6nree/hr) _

(mrem /hr)

(mrom/hr)

(crem/hr)

~ l' 0'

1 1

15 2

0.5 0.5 1-15' 3

0.6 0.5 1.1 15 4

0.4 0.5 0.9 15 5

1.2 1.0 2.2 15 6

8 35 43 15 7

3.5 25 28.5 15

~

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8 40 48 15 9

4 25 29 15 10 2

20 22 15 11 0.5

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15 12 0.5 0.5 1

15

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2 2

15 14~

0 0

0 15 15~

0' 1.0 1

15 16-5:

40 45 15 17 12.

60 72 15 184 il2

'60 72 15

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50-

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20

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15 322 0.5 5

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TAllLE II-3 REACTOR BUILDING +20' ELEVATION

. SURVEY Y DOSE RATE n DOSE RATE TOTAL DOSE RATE ZONE LIMIT LOCATION (nrem/hr)

(mrom/hr)

(nrem/hr)

(mrem /br) 1-0.4-1.0 1.4 15 2

0.4 1.0 1.4 15 3

0.2 0.5 0.7 15 4

15 4

0.2 0.5 0.7 5

0.4 0.5 0.'9 15 6

0.8 0.5 1.3 15 7

0.4 0.5 0.9 15 8'

O.2 1.0 1.2 15 9'

O.2 0

0.2 15 10 2.5 0.5 3.0 15 11 0.7 0.5 1.2 15 12.

0.4 0.5 0.9 15 l

13 0.2 0.5 0.7 15 14-1.0 0.5 1.5 15 15 1.1 1.0 2.1 15 16

~ 1.0.

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1.0' 2.0 15 18 cl.2 0.5 1.7 15 19' O.8 0.5 1.3 15 20

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REACTOR BUILDING GRADE ELEVATION SURVEY

-y DOSE RATE n DOSE RATE TOTAL DOSE RATE ZO::E-LIMIT

LOCATION (nren/hr)

(nrem/hr)_

(nren/hr)

(mrem /hr) 1

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15

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3 2.2 1.0 3.2 15

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2.4 1.0 3. <.-

15 6-2.4 1.0 3.*

15' 7

1.6 1.0 2.6 15 8

0.9 0.5 1.4 15 9

0.3 1.0 1.3 15 10-0.2 0.5 0.7 15 11 0

'O.5 0.5 15 12 1.0 0.5

- 1. 5 15 13?

.0.8 0.5

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0.8 0.5-1.3 15 15 0.6 0.5 1.1 15' i

16 1.8 0.5 2.3 15 17 1.4 1.0 2.4

-15 181 5-1.0.

6 15

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3 15

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' TABLE II-5

-REACTOR BUILDING -27'/-12' 8" ELEVATION

' SURVEY Y. DOSE RATE n DOSE RATE TOTAL DOSE I: ATE ZCNE LD!IT

- LOCATION.

(mrom/hr)'

(mrem /hr)

(mree/hr)

(mrem /hr) 1.

4 2

6 none'

'2 2000 1000-3000' none 2

3 180 1.

181 none 4-30 1

31 none 5

12 1

13 none 6

20 25 45 15 7'

1100 500 1600 none 8

2 3.5 5.5 none b

.9 4

1 5

none 10 3

1 4

'none 11 2

5 7

none 12-

'0.5?

1 1.5 none 13 1000 2000-3000 none

~14 1200

-1000 3000 none 4

r.'

x 05 l.

f e-

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r 1 W v.

N tt

.M. '~a -*n c

  • % f,t 5,

4

+ -ww N 5.*-.

u.

oy w

.,,,'**,'i

,e'

@C l

w

P--.L1.

g yQ

- (7.--

ww

.=

g 3

< 03 m

~

E

_ _/

m

/

.6

/

N

/

\\

\\

/ /

N N

j j

- f

\\.,

-(g.

l

\\.<,

e kc l

e rg r-7 t 3 x

Klh /

n

.X 1 l s

y t

n.

j

_t' h - @

i gk 1

i f-g,,

x 3

. le K '.y x

N'.-

,7

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s q

s a

f

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a h

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k'

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l f.

fM.

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e e.

4.

4 5

T

F-I'

!. i. ;.,

TABLE II-6 REACTOR RUILDING 60'~ ELEVATION

1. -
POINT NO.

LOCATION ACTUAL SURVEY ACCEPTANCE CRITEP,IA COORDINATES DOSE RATE DOSE RATE (mrem /hr)

(mrem /hr) 15 15 42'E, 25'S 15 11 25 45's, 20's 15 2'

19 42'E, 45'N 15 10' 17 22'E, 22'N 15 11-17 14 25'W, 48'N REACTOR BUILDING 40' ELEVATION _

2.

POINT NO.

LOCATION ACTUAL SURVEY ACCEPTANCE CRITERIA I

COGRDINATES DOSE RATE DOSE RATE (mrem /hr)

(mrem /hr) 15 48'U, 34'S 43 15 l

6 28 l

'7 42'W, 17'S 15 48 63'W, 12'S 15 8

29 9

58'W, 15'N 15 22 10 38'W, 35'N 15 16 25'E, 35'N 45 15 l l'-

17-22'E, S' N

-72 18 22'E, 5' S 72 15_

49 15 l

19 38tE, 2' N 15

[

62'E, 12'N 15 22 l

20 22 l

i 21 61'E, 12'S l

R_EACTOR T iLDING GRADE ELEVATION.

.3.

POINT NO.

LOCATION

' ACTUAL SURVEY ACCEPTANCE CRITERIA l

COORDINATES DOSE RATE DOSE RATE (mrem /hr)

(mrem /hr) 41 15 2 --

18'W, 48'N

-4.-

' REACTOR BUILDING -27' ELEVATION

, POINT NO.-

LOCATION' ACTUAL SURVEY ACCEPTANCE CRITERIA H

l

. COORDINATES DOSE RATE DOSE RATE (mren/hr)

(mrem /hr) l 6

30'E, 52'S 45 15 l

Le N.

.,

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