ML17249A372

Plant Transient Analysis, Revision 1

ML17249A372
Person / Time
Site:	Ginna
Issue date:	07/03/1979
From:	Kahn J, Leach C, Markowski F SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17249A368	List: ML17249A367 ML17249A369 ML17249A370 ML17249A372 ML17249A373 ML17249A374
References
XN-NF-77-040, XN-NF-77-040-R01, XN-NF-77-40-R1, NUDOCS 7912280222
Download: ML17249A372 (25)

Text

PLANT TRANSIENT ANALYSIS FOR THE R.

E.

GINNA UNIT 1

'UCLEAR POWER PLANT XN-NF-77-40 Revision 1

s e Date: 07/03/79 Prepared by F. J.

Markowski J.

D ~

Kahn Approved:

C.

E. Leach, Manager Thermal-Hydraulic Engineering Date

. Approved:

G. A.

So

, Many er Nuclea~uels En)ineering Date Approved:

W. S.

Nechodom, Manager Licensing and Compliance d'-27 Date Approved:

G. J.

8 sselman, Mana Contract Performance Date NONPROPRIETARY 9XQ880

C t\\

XN-NF-77-40 LIST OF TABLES Table 2.1 2.2 2.3 2 ~ 4

SUMMARY

OF RESULTS, Revision 1

PARAMETER VALUES USED IH PTSPMR2 ANALYSIS OF R.

E.

GINNA UNIT 1

R.

E.

GINNA UNIT 1 TRIP SETPOINTS R.

E.

GIHNA UNIT 1

FUEL DESIGN PARAMETERS EXXON NUCLEAR FUEL R.

E.

GIHNA UNIT 1

ENC KINETIC PARAMETERS

~Pa e

~

~

~

~

3

~

~

~

~

7

~

~

~

~

8

~

~

~

~

9 10

2. 5 MODERATOR AHD DOPPLER COEFFICIENTS............

11 4.1 4.2 COMPARISON OF TRANSIENT-SPECIFIC INPUT PARAMETERS COMPARISON OF OPERATING PARAMETERS FOR R.

E.

GINNA UNIT 1

~

~

~

~

~

~

~

~

\\

~

~

~

~

~

~

~

~

~

~

~

~

69

~

~

~

o 70 4.3 COMPARISON OF R.

E.

GINNA UNIT 1

KINETIC PARAMETERS.........,..........

72

-v-XN-NF-77-40 LIST OF FIGURES (Continued)

Fi<iure

3. 33

3. 34 3.35 3.36 3.37 3.38 3.39 3.40

3. 41 3.42 3.43 3.44

POWER, HEAT FLUX AND SYSTEM FLOWS-LARGE STEAM LINE BREAK, Revision 1

CORE TEMPERATURE RESPONSE-LARGE STEAM LINE BREAK, Revision 1

PRIMARY LOOP TEMPERATURE CHANGES LARGE STEAM LINE BREAK, Revision 1

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS LARGE STEAM LINE BREAK, Revision 1

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS-LARGE STEAM LINE BREAK, Revision 1

NUCLEAR REACTIVITY FEEDBACK EFFECTS-LARGE STEAM LINE BREAK, Revision 1

POWER, HEAT FLUX AND SYSTEAM FLOWS-SMALL STEAM LINE BREAK, Revision 1

CORE TEMPERATURE RESPONSE SMALL STEAM LINE BREAK, Revision 1

PRIMARY LOOP TEMPERATURE CHANGES-SMALL STEAM LINE BREAK, Revision 1

PRESSURE CHANGES IN PRESSURIZER AND STEAM GENERATORS SMALL STEAM LINE BREAK, Revision 1

LEVEL CHANGES IN PRESSURIZER AND STEAM GENERATORS-SMALL STEAM LINE BREAK, Revision 1

NUCLEAR REACTIVITY FEEDBACK EFFECTS-SMALL STEAM LINE BREAK, Revision 1

~

~

S

~Pa e

53 54 55 56 57 58 59 60 61 62 63 64

XN-NF-77-40 Revision 1

U dates for Revision 1

Revision 1 was issued in June 1979.

The revision was issued because the pressure feedback coefficient and the boron worth coefficient for Cycle 9 were slightly changed from their respective Cycle 8 values.

Using the same dimen-sions as in Table 2.4, the changes are:

Previous Value Updated Value EOC pressure'oefficient EOC boron worth coefficient

+4.0

-0.950

+3.5

-0.872 The transients affected by these changes were the large and the small steamline break.

Since the steamline break is a depressurization transient, the positive moderator feedback contributes some negative reactivity.

The boron worth coefficient is used for the borated water coming from the high pressur'e injection line; This injection terminates the power generation in case of the large steamline

break, and it preserves the shutdown margin in case of the small steamline break.

The analysis 'showed that the thermal margin values previously calculated were still protected for the updated reactivity data.

All changes have been incorporated into the report, all applicable pages have been'arked "Revision l."

The revised analysis is expected to cover the future reload cycles for the R.

E. Ginna Plant.

XN-NF-77-40 Revision 1

Table 1.1 Summar of Results Transient Cl.ass Initial condi tions for transients Uncontrolled rod withdrawal (II) 9 6.0 x 10 1/sec Uncontrolled rod withdrawal (II) 9 5.0 x 10 1/sec Loss of flow (III) 2-pump coastdown Loss of flow (IV)

Locked pump rotor Loss of load (II)

Large steam line break (IV)

Small steam line break (IV)

Maximum Power Level MWt 1550.4 1906 1748 1550.4 1550.4 1648 532 Maximum Core Average Heat Flux Btu hr-ft~

177,560 195,120 197,850 181,170 181,160 184,320 61,670 Maximum Pressurizer Pressure sia 2220.

2235 2283 2250 2273 2511 MDNBR

~W-3

2. 00 1.77 1.73 1.61 1.23 1.'.83 58***

Pressure decreases-from initial value.

The core does not go critical.

• • • MDNBR calculated using t1acbeth correlation.

10 XN-NF-77-40 Revision 1

Table 2.4 R.

E. Ginna Unit 1

ENC Kinetic Parameters

~Smbo 1 aD ap Parameter Moderator Coefficient (ap/F x 10

)

Doppler Coefficient (ap/F x 10

)

Pressure Coefficient (hp/psia x 10

)

Be innin -of-C cle 0.0

-1.25

+2.5 Value End-of-C cle

-3. 5

-2.00

+3. 5 CCy aB eff

'RC Moderator Density Coefficient

(%hp)/(g/cm

)

Boron Worth Coefficient (hp/ppm x 10

)

Delayed Neutron Fraction

(%)

Total Rod Worth

(%ap) 0.0

-0.875

0. 610 1 89**

+29.635

-0.872 0.510

-2.83*+

• • Minimum required (these are conservative values for analysis purposes only; the actual va 1ues are s ignificantly hi gher).

TABLE 2.5 MODERATOR AND DOPPLER COEFFICIENTS Transient

Fast, Rod Withdrawal Slow Rod Withdrawal Pump Coastdown Locked Pump Rotor Loss of Load Steam Line Break Desired Moderator Feedback Effect Minimum Minimum minimum tlinimum Minimum

tlaximum, Moderator Resulting Coefficient

~Hulti lier eu/'f x 10" 0.0 0.0 0.0 0.0 0.0 tlinimum tlinimum flaximum tlaximum Minimum minimum 0.8 0.8 1.2 1.2 0.8

-1. 0

-1. 0

-1.5

-1.0 Doppler Desired Doppler Resul ting Feedback Coefficient Effect

~Multi lier du/.'" x lee For minimal effect no moderator feedback allowed.

See Figure 3.31

~

.~**. See Figure 3.32

XN-NF-77-40 which could, under pessimistic circumstances, lead to criticality and core damage if unchecked.

As a worst case, the steam line break is assumed to occur at hot zero power conditions.

At this time, the steam. generator secondary side water inventory is at a maximum, prolonging the duration and increasing the magnitude of the primary loop cooldown.

For conservatism, the most reactive control rod is assumed to be stuck out of the core when evaluating the shut-down capability of the control 'rods'he reactivity as a function of core average temperature and the variation of reactivity as a function of core power used in this analysis are shown in Figures 3.31 and 3.32 'he moderator and Doppler feedback coefficients are valid for Cycle 8 fuel.

Minimum capability of the boron injection system was

assumed, which implies that only two of the three high-pressure safety injection pumps (HPSI) are available.

A low pressurizer pressure signal in combination with low pressurizer level initiates the safety injection system.

Borated water starts entering the injection lines after the pressurizer pressure has come down to the shutoff head (1400 psia) of the injection pumps.

The time required to sweep the lines of low concentration borated water prior to the introduction of 20,000 ppm borated water from the Boric Acid Tanks has been r

accounted for in the analysis.

No credit was taken for the effects of the resident low concentration borated water being swept into the primary loop from the safety injection lines.

The initial pressure was. set at 2280 psia (nominal

+30 psi) to delay the onset of safety injection.

Two steam line breaks were analyzed.

The large br'eak at the exit of the steam generator with outside power available was shown in the reference cycle analysis to give the greatest return to power and the highest core (6)

20 XH-NF-77-40 Revision 1

average heat flux.

This case was analyzed for an ETC-fueled core to ensure the core integrity is maintained during the transient.

A 1

~ 8 percent shut-down margin has been us'ed in this analysis.

A small steam line break assuming one loop operation was analyzed to ensure an adequate shutdown margin exists at the end of the cycle such that the core does not go critical during such an event (safety valve, relief valve, or bypass valve failed open).

The system responses for a large steam line break at the exit of one steam generator (initial flow - 518 percent of rated value) are shown in Figures 3.33 to 3.38.

The core returns to criticality at about 20 sec.

The power reaches a peak value of about 532 MW (35 percent of rated power) at 90 sec with a corresponding peak core average heat flux of 61;670 Btu/hr-ft

~

~

2 At this time, the borated water reaches the core, initiating a power decrease.

As the core parameters (pressure, flow, i~let enthalpy) at the time of peak heat flux are outside of the range of the W-3 correlation, the critical heat flux was determined using the modified Macbeth CHF correlation.

At the time of maximum core average heat flux, the margin to the critical heat flux is minimized. Using the core conditions for this time (22 percent of rated core averate heat flux, 386 psia, and 387'F inlet temperature) and applying a conservatively large local hot rod peaking (F

14.0), 'the minimum CHF T

1 ratio was calculated to be 1.58

~

The responses to a small steam line break (273 lb/sec) are shown in Figures 3.39 to 3.44.

The boron injection again is triggered by a low pressurizer pressure signal in combination with low pressurizer levels Borated water from the safety injection system reaches the core at 175 sec.

N The shutdown margin of 1.8 percent ensures that the core does not go critical following the small steam line break.

350 R E QhlvA + URZ SiBH.JhE BRBK + 15 OEC 78 +

PCVER L C.

iKATFUK 300 4.

5.

TOTAt PR a OTAt.

TOTfg. ST ART COOtAlST ATER FU)M

~IV 250

~200 LU C)

~150 4J CJK 4J Q.

50 513.5 0

30 40 50

TINE, SEC 70 80 380 SEOUL CCPOX41 LS/12/F8 ORrOF.33 Figure 3.33

Power, heat flux and system flows - large steam line break

R E G24% + (URGE GFGK3hE BREA' 15 KC 78 +

550 l.

2.

3.

CORE IIII.

AVE. COR CLAD TE TEIIP ERAT COOLANT T ERATORE 500 475 I-

-450 III CKK II.

3

$00 375 0 30 50 I IHE, SEC EEO.

CXPOXC7 16'2/78 O2. 07. 33.

Figure 3.34 Core tetaperature response

- large stea01 line break.

2.

R E G2+R + LIKE SiEA42hE BRAC + 15 gZ 78 +

RVE.

PRTNAR COOLANT RVK. PRZIIAR ~T LOOP I LOOP 2 30 IIHEr SEC 10 SO SEA.

CZPOX47 M/l2/7II

02. O7 ~ SS ~

Figure 3.35 Primary loop temperature changes

- large steam line break.

R E GZh&l + LFKE SIBLE BK-K + 15 KC 78 +

0 2

S.

STERN STEAN PRESSURE E PRESSURE E PRESSURE PRESSURE NOES lOOP 1 r

LOOP 2 CE

-200

%00 2

aH~00

-1<0 20

~.

30

$0 50 60 I lHEi SEC 3

70 80 SO OC m z CD IR VI I

0 100

~O SEO.

CXPOX47 1CC12C78 0?. ~ 07 ~ 33 ~

Figure 3.36'ressure changes in pressurizer and steam generators

- large steam line break

0 R E GPhA + L%GE.'HH+f '!ATE 883K + 15 CEC 78 +

CRACE QQuCE SsEfH CEN.

S'iEAH CEN.

TER LE'/EL.

fEi? LEVELS z-150 H

td C9K CZxo-200

-250 "300

-35$

40 50 TIVE.

SEC 70 80 OC RZ rp IR EA 1

0 380

~O SEQ. CIPOX4f 1c/12/f S O2. Of 33.

Figure 3.37 Level changes in pressurizer and steam generators

- large steam line break.

R E Qhl4 + URZ SRM3lK BRB-K + 15 CEt: 78 +

IIOPERAT POP PLER BOROII R TOTAL REACTZVZTV ZVZTY ZVZTY ZVZTY 23'0 I IHEi SEC 10 80 80 4

SEQ.

CZPOX47 18712/7II O?.07.33 Figure 3.38 Nuclear reactivity feedback effects - large steam line break.

R E GIN@ + SNJ SEAL%'RH-K + SE LKP CPERA1TN + 11 CK 78 +

140 1.

2.

3.

4.

5 POVER L EL HEATFLUX TOTAL PR TOTAL TOTAL ST lQRY COOLANT LOV OWATER FLOV ARLIHE FLOV

~80 h.o

~60 K

Lu CJ 40 0

40 1

2 80 120 2.

1 2

200 240 TlME'EC 280 380 OC Vl I

400

~O SEA.

CXRQWO l2/12/7$

01.2$

~ 27 ~

Figure 3.39 Power, heat flux and system flows - small steam line break.

R E G&M + SNJ-S1EALIhE BREA( + QE LOCP CPERA7lN + ll CEC 78 +

575 1.

2.

3.

CORE T TB4PERATLR AVE.

COOLANT T CLAD TE ERATURE 550 500

~ 475 450 425 I

400 80

~

120 TAHE. st 240 280 3GO EED-CXÃ4AVD 12,I12ITC 01.22

~ 27 ~

Figure 3.40 Core temperature response

- small steam line break.

I W

r r

R E Or+A + SHI-LL SiEPH 'BE BRBX + QE MCP CPERATEH + ll CEC 18 +

CHA~ I CHAHCE I AVE. PRINAR AVE. PRIHAR COOL%IT TE COOLfHT T

LOOP l.

LOOP Z.

80 TXMEr SEC HO 280 380

>C I

U)

I Q

400

~O SEA CIHHAVD 12/l2/I8 0l ES ~ 27 ~

Figure 3.41 Primary loop temperature changes small steam line break.

R E QM + 9'1EA42hE SR' QE UXP KERATIN + 1L CEC 18 +

STEAM DO SIEAH PRESSUR E PRESSURE E PRESSURE Ce ER PRESSURE MCEi LOOP L

MCE.

LOOP 2 RMCE 0

"K50

%00 CL "750 a.

40 80 120 YXHEi SEC 240 2.80 380

$00 SEO.

CXMMAVD l?/12/7e OX.2e.27.

Figure 3.42 Pressure changes in pressurizer and steam generators

- small steam line break.

R E GIM + BAL SEEKS% BREA( + 0% UXP CPERATKN + 1L OEC 78 +

80 2.

3.

CHANGE I CHANCE I CHANCE I STEAN CEH.

STEAN CEH.

PRESSURXZER TER LEVEL. L P

1 TER LEVEL, P 2 dATER LEVEL z-40 III K

CK o+0 IIj III "120

-180 40 80 180 TXHEi SEC 2.40 P.80 380 400 OC ID IX Ul I

0 I

~O SEO.

CIHHAVO 12/12/TB

01. 2B ~ 21 ~

Figure 3.43 Level changes in pressurizer and steam generators

- small steam line break.

R E GXhhA + PAL S1EA~" BREA( + QE UXP CPERAIIOM + 11 IXC 78 +

2 3.

4.

HOOERAT REACTZVZTY OOPPLER EACTZVZTY BORON' CTZVZTY TOTfL, R TZVZTY 1

2

'2 3

2 3

2 3

$0 80 TXHEr SEC 360 OC C R Vi I

400

~ Q, SEO.

CZNNAVO 12J 12I13

01. 2S. ~.

Figure 3.44 ttuclear reactivity feedback effects

- small steam line break.

Plant Transient Analysis for the R.

E.

Ginna Unit. 1 Nuclear Power Plant

'XN-NF-77-40 Revision 1

Issue Date: 07/03/79 Distribution F.

deWaegh K. P. Galbraith J.

S.

Holm S.

E. Jensen J.

D.

Kahn F. J.

Markwoski J.

N. Morgan G. F. Owsley F.

B. Skogen G. A. Sofer RG8E/L. J.

Federico (80)

Document Control (10)

L. J.

Federico

1 I