Text

Plant Transient Analysis for Operation of Re Ginna Unit 1 Nuclear Power Plant at Reduced Pressure & Temp
	ML20062K429
Person / Time
Site:	Ginna
Issue date:	07/22/1982
From:	Cooke G, Morgan J, Nutt W; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML17256B178	List: ML17256B177; ML17256B179; ML17256B180; ML20062K429; ML20062K454;
References
	XN-NF-82-45, XN-NF-82-45-R01, XN-NF-82-45-R1, NUDOCS 8208170188
	Download: ML20062K429 (98)
	v • d • e

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l XN NF 82 45 g

REVISION 1

I i l l

il PLANT TRANSIENT ANALYSIS FOR OPERATION OF THE

!I R.E. GINNA UNIT 1 NUCLEAR POWER PLANT AT lg REDUCED PRESSURE AND TEMPERATURE li I

JULY 1982 l

l ERON NUCLEAR COMPANY,Inc.

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XN-NF-82-45 l

Revision 1 ISSUE DATE:

l 07/22/82 PLANT TRANSIENT ANALYSIS FOR OPERATION OF THE R.E. GINNA UNIT 1 NUCLEAR POWER PLANT AT REDUCED PRESSURE AND TEMPERATURE I

Prepared by : [//T W.T. 'NuM /

Prepared by :

[

G.C. Cooke, Manager Plant Tra sient Analysis Approve :

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'N'.' Bofgan, Mansger I

icensing & Safety Engineering Approve :

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g E^i P#de#"i"*3ETechnical Services e

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ERON NUCLEAR COMPANY,Inc.

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XN-NF-82-45 I

Revision 1 I

TABLE OF CONTENTS I

SECTION PAGE lI

1. 0 I NTRO DU CT IO N AN D SUMM AR Y.....................................

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l 2.0 CALCULATIONAL METHODS AND INPUT PAR AMETERS...................

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3.0 TRANSIENT ANALYSIS 3.1 Initial Conditions......................................

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3.2 Uncontrolled Rod Withdrawal.............................

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3.3 Loss of Coolant Flow....................................

20 3.4 Locked Rotor............................................

21 3.5 Loss of Electric Load...................................

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3.6 Steam Line 8reaks........................................

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4.0 CONCLUSION

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5.0 REFERENCES

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Revision 1 LIST OF TABLES Table Page 1.1 Summary of Results for TAVE Reduction of 150F...............

5 2.1 TAVE Schedules for Operation at 2000 psia...................

12 2.2 Thermal Parameters for Operation of R.E. Ginna I

Unit 1 at Reduced Average Temperature and Primary System Pressure......................................

13

2. 3 Re ac to r T r i p Se tpo i n t s......................................

14 2.4 Exxon Nuclear Reload for R.E. Ginna Unit 1 Fuel Design Parameters......................................

15 2.5 R.E. Ginna Unit 1 K inetic Parameters........................

16 2.6 Moderator and Doppler Coef f ic ient s..........................

17 3.1 Uncontrolled Rod Withdrawal (Fast)-- Event Table.............

26 3.2 Uncontrolled Rod Withdrawal (Slow) - Event Table............

27 3.3 Loss of Coolant Flow - Event Table..........................

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3.4 Locked Rotor - Event Table..................................

29 3.5 Loss of Electric Load - Event Table.........................

30 3.6 Large Steam Line Break (5470F) - Event Table................

31 3.7 Large Steam Line Break (5140F) - Event Table................

32 3.8 Small Steam Line Break - Event Table........................

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3.9 Pressure Control and Safety J3rameters.....................

34 4.1 PPS Trip Settings for Reduced T, P..........................

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LIST OF FIGURES I

Figure Page 2.1 PTSPWR2 System Model.....................................

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3.1 Uncontrolled Rod Withdrawal (Fast) -

Power, Heat Flux and System Flows........................

35 3.2 Uncontrolled Rod Withdrawal (Fast) -

5 Core Temperature Response................................

36 3.3 Uncontrolled Rod Withdrawal (Fast) -

Primary Loop Temperature Changes.........................

37 3.4 Uncontrolled Rod Withdrawal (Fast) - Pressure Changes....

38 3.5 Uncontrolled Rod Withdrawal (Fast) - Level Changes.......

39 3.6 Uncontrolled Rod Withdrawal (Fast) -

I Minimum DNB Ratio........................................

40 3.7 Scram Curve Used in R. E. Ginna Unit 1 1

Transient Analysis.......................................

41 3.8 Uncontrolled Rod Withdrawal (Slow) -

Power, Heat Flux and System Flows........................

42 3.9 Uncontrolled Rod Withdrawal (Slow) -

Core Temperature Response................................

43 3.10 Uncontrolled Rod Withdrawal (Slow) -

Primary Loop Temperature Changes.........................

44 3.11 Uncontrolled Rod Withdrawal (Slow) - Pressure Changes....

45 3.12 Uncontrolled Rod Withdrawal (Slow) - Level Changes.......

46 3.13 Uncontrolled Rod Withdrawal (Slow) -

Minimum DNB Ratio........................................

47 3.14 Loss of Coolant Flow - Power, Heat Flux and System Flows.............................................

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Revision 1 I

LIST OF FIGURES (Continued)

I Figure Page 3.15 Loss of Coolant Flow - Core Temperature Response........

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3.16 Loss of Coolant Flow - Primary Loop Temperature Changes.................................................

50 3.17 Loss of Coolant Flow - Pressure Changes.................

51 3.18 Loss of Coolant Flow - Level Changes....................

52 3.19 Loss of Coolant Flow - Minimum DNB Ratio................

53 3.20 Locked Rotor - Power, Heat Flux and System Flows............................................

54 3.21 Locked Rotor - Core Temperature Response................

55 3.22 Locked Rotor - Primary Loop Temperature Changes.........

56 3.23 Locked Rotor - Pressure Changes.........................

57 3.24 Locked Rotor - Level Changes............................

58 3.25 Locked Rotor - Minimum DNB Ratio........................

59 3.26 Loss of Electric Load -

Power, Heat Flux and System Flows.......................

60 3.27 Loss of Eiectric Load - Core Temperature Response.......

61 3.28 Loss of Electric Load - Primary Loop Temperature Changes.................................................

62 3.29 Loss of Electric Load - Pressure Changes................

63 3.30 Loss of Electric Load - Level Changes...................

64 3.31 Loss of Electric Load - Minimum DNB Ratio...............

65 3.32 Variation of Reactivity with Power at Constant I

Core Average Temperature................................

66 3.33 Variation of Reactivity with Core Average I

Temperature at the End of the Cycle.....................

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XN-NF-82-45 I

Revision 1 i

LIST OF FIGURES (Continued)

I Figure Page 3.34 Large Steam Line Break (5470F) - Power, Heat Flux and System Flows...............................

68 3.35 Large Steam Line Break (5470F) - Core Temperature Response.................................................

69 3.36 Large Steam Line Break (5470F) - Primary Loop Temperature Changes......................................

70 3.37 Large Steam Line Break (5470F) - Pressure Changes.......'

71 3.38 Large Steam Line Break (5470F) - Level Changes...........

77 3.39 Large Steam Line Break (5470F) - Reactivity..............

73 3.40 Large Steam Line Break (5140F) - Power, 5

Heat Flux and System Flows...............................

74 3.41 Large Steam Line Break (5140F) - Core Temperature Response.................................................

75 3.42 Large Steam Line Break (5140F) - Primary Loop Temperature Changes......................................

76 3.43 Large Steam Line Break (5140F) - Pressure Changes........

77 3.44 Large Steam Line Break (5140F) - Level Changes...........

78 3.45 Large Steam Line Break (5140F) - Reactivity..............

79 3.46 Small Steam Line Break (5470F) - Power, Heat Flux and System Flows...............................

80 3.47 Small Steam Line Break (5470F) - Core Temperature Response.................................................

81 3.48 Small Steam Line Break (5470F) - Primary Loop Temperature Changes......................................

82 3.49 Small Steam Line Break (5470F) - Pressure Changes........

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l LIST OF FIGURES (Continued)

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Figure Page 3.50 Small Steam Line Break (5470F) - Level Changes..........

84 3.51 Small Steam Line Break (5470F) - Reactivity.............

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XN-NF-82-45 8

Revision 1 l.0 INTRODUCTION AND

SUMMARY

This document presents plant transient analysis to support uperation of the R.E. Ginna Unit I nuclear power plant with reduced primary coolant temperature and pressure.

Specifically this document supports a full load I

(1520 MWt) TAVE reduction from 573.50F to 558.50F and a primary system pressure reduction from 2250 psia to 2000 psia. The results of the analysis show generally greater thermal margins during limiting transient events for the new conditions than for normal temperature and pressure operation. This improvement in margin is because the 150F reduced temperature condition

.I outweighs the slight adverse effect on DNBR of reduced primary system pressure.

In performing the analysis reductions in primary temperature of 150F to 500F were considered.

Thermal margin results for 150F reduced temperature were found to bound the results for larger reductions in primary coolant temperature.

The reference TAVE and pressure inputs into the calculated overtem-perature AT and overpower AT trip functions were changed to reflect reduced temperature and pressure operation in order to maintain the steady state margin to trip and so that improvements in the initial DNBR also applied during transients that were protected by these two trip functions.

Plant transient analyses for ENC relnad fuel at R.E. Ginna Unit 1 are documented in References 1 and 2.

These analyses covered normal temperature and pressure operation.

Substantial margin improvements for 500F reduced temperature and 2000 psia pressure operation were shown in Reference 3. The present analysis supports operation with 150F to 500F reduced temperature operation at 2000 psia and hence bounds the analysis of Reference 3.

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XN-NF-82-45 Revision 1 As in the case of prior analyses, the present analysis was performed using the Exxon Nuclear plant transient simulation code, PTSPWR2(4) with supporting subchannel analysis using the standard ENC methodology (5).

In order to support operation over a range of TAVE schedules, two bounding schedules representing the highest and lowest schedules of interest were considered. The higher schedule was found to be more limiting with respect to thermal margin because it has a much less favora'cle margin at power, and because moderator feedback is much stronger at higher temperatures. Conse-quently the higher schedule results in lower calculated MDNBRs for all events initiated from full power and, during the steam line break transients, the core has a greater tendency to return to power than any lower temperature schedule. Thus, this analysis supports operation with TAVE reduced from 150F to 500F below the current TAVE schedule.

The design basis events, listed below, as well as the input parameters I

used to simulate the reactor system, are reported herein:

Event Incident Class

1.

Uncontrolled Control Rod Withdrawal Fast Rod Withdrawal II Slow Rod Withdrawal II 2.

Loss of Coolant Flow III 3.

Locked Rotor IV 4.

Loss of External Electric Load 11 I

5.

Large Steam Line Break IV 6.

Small Steam Line Break IV

Consistent with current FSAR incident classification for PWRs.

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XN-NF-82-45 Revision 1 Events 1, 2, 3 and 4 were initiated f rom full power, while events 5 and 6 were initiated from hot zero power (HZP).

The criteria for Class II and III events are:

(1) Peak System pressure should not exceed 2750 psia (= 110% of Design); and (2) the minimum departure from nucleate boiling ratio (MDNBR) should be greater than the 95/95 value of 1.3 for the W-3 correlation (5),

I In the case of Class IV accidents, some fuel damage is acceptable provided it is confined to a limited number of fuel rods in the core.

The criterion for steam line breaks is that shutdown margin must be sufficient to limit the occurrence of boiling transition in the core so that the extent of potential core damage is small for a large steam line break and that the core does not go critical following a small steam line break.

The analyses are based on an equilibrium ENC fueled core using conservative neutronic parameters calculated for ENC fuel.

The results of the calculations are summarized in Table 1.1.

The lowest MDNBR for Class II and III events initiated at 1520 MWt was 1.70 for the slow rod withdrawal transient. Evaluation of the bounding pressure transient, loss of electric load, indicates that peak primary system pressures wiIl not exceed the 2750 psia vessel integrity limit. The locked rotor accident, a Class IV event, was analyzed and the MDNBR was found to be 1.26.

While this value did not meet the 1.3 criterion, based on the W-3 correlation, the result is acceptable in light of the low probability of the event and the extremely short time the DNBR was below 1.3.

The large steam line break resulted in a minimum critical heat flux ratio of 1.10 based on the Modified MacBeth Correlation (6,7). Less i I than 1% of the fuel rods undergo boiling transition at this DNBR. The small I

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XN-NF-82-45 Revision 1 steam line break did not result in the reactor going critical. The results of the analysis show that the transients initiated from full power at reduced TAVE have increased thermal margins relative to prior analysis for normal temperature and pressure operation.

Finally it is noted that both the present and prior analyses have significant conservatism relative to thermal margins that might be expected if one of the postulated transients actually occurred. This conservatism is built into the analysis by stacking plant operating condition uncertainties to minimize MDNBR, by using conservative reactor trip setpoints and trip delays, and by using bounding reactor kinetics parameters. In addition, the transients generally represent worst case scenarios (e.g. stuck control rod for steam line breaks, neglect of a direct trip for pump coastdown, etc.)

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M Table 1.1 Siimmary of Results for TAVE Reduction of 150F Maximum Maximum Core Average Pressurizer Transient Power Level Heat Flux Pressure MDNBP 2

Class (MWt)

(Btu /hr_ft )

(psia)

(W-3) i I

Initial Conditions for 1520.0 177,608 2000 2.07 Transients Uncontrolled Rod 1861.5 191,492 2360 1.84 4

Withdrawal (II) i

@ 6.0 x 10-4 Ap/sec Uncontrolled Rod 1747.3 198,802 2366 1.70 Withdrawal (II)

@ 5 x 10-5 Ap/sec un Loss of Flow (III) 1520.0 177,608 2355 1.74 2-Pump Coastdown Loss of Flow - (IV) 1520.0 177,608

~2384 1.26 Locked Rotor loss of Load * (II) 1550.7 181,160 2528 2.07 Large Steam Line Break (IV) 685.2 74,165 2000 1.10**

Small Steam Line Break (IV) 2000 I

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Initiated from 102% power

r. n Calculated with the Modified MacBeth Critical Heat Flux Correlation 82 i

"h Does not go critical 1

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XN-NF-82-45 Revision 1 2.0 CALCULATIONAL METHODS AND INPUT PARAMETERS The analysis of R.E. Ginna transient performance was performed using the Exxon Nuclear Company plant transient simulation model for pressurized water reactors, PTSPWR2, a digital computer program developed to model the behavior of pressurized water reactors under normal and abnormal operating conditions.

The model is based on the solution of the basic transient conservation equations for the primary and secondary coolant systems.

The transient conduction equation is solved for the fuel rods, and a point kinetics model is used to calculate the core neutronics.

The program calculates fluid conditions such as flow, pressure, mass inventory and steam quality, heat flux in the core, reactor power, and reactivity during the transient.

Various control and safety system components are included as necessary to analyze postulated events.

A hot channel model is included to trace the departure from nucleate boiling (DNB) during transients. The DNB evaluation is based on the hot rod heat flux in the high enthalpy rise subchannel and uses the W-3 correlation to calculat e the DNB heat flux for pressures greater than 1450 psia and, as an optio1, a modified MacBeth correlation for pressures less than 1450 psia.

The PTSPWR2 code models the reactor, two independent primary coolant loops (including all major components such as the pressurizer, both pumps, and the piping), two steam generators, and their steam lines (including all major valves such as turbine stop valves, isolation valves, and pressure relief valves).

Figure 2.1 is a system schematic representing the model elements in PTSPWR2 and their interaction. For a more thorough discussion of the model details of the PTSPWR2 code, see Reference 4.

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7 XN-NF-82-45 Revision 1 I

Several updates were included in the present onalysis to (1) improve the I

initial steady state plant balance; (2) correct pressurizer surge flow calculations; (3) to incorporate the effects of pressurizer heaters on pressure control; and (4) to terminate steam generator heat losses af ter the steam inventory was exhausted. The pressurizer surge flow calculations are conservative for the loss-of-electric-load event, where the quantity of I

interest is the maximum primary system pressure.

The maximum pressure attained is limited by surge flow or by safety relief valve capacity versus plant heatJp rate.

For steam line breaks, the pressurizer control tries to maintain the pressure by turning the backup pressurizer heaters full on.

The resulting expansion of the water delays safety injection, which occurs during the rapid pressure drop accompanying the emptying of the pressurizer of water.

The pressurizer control has a slightly conservative effect on the transients from power, except for the loss-of-electric-load event, because pressurizer spray tends to hold down the pressure at the time of MDNBR, thereby lowering the MDNBR.

The correction of heat loss to the steam generator has no impact, since it occurs after the boron from the safety injection has reached the reactor.

Conservative approximations are applied for predicting those system responses which contribute to minimum values of the DNB ratio.

These approximations are categorized as either: (1) generic approximations applied to the steady state DNBR to account for plant instrumentation errors; (2) approximations which conservatively bound R.E. Ginna Unit 1 neutronics I

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8 XN-NF-82-45 Revision 1 parameters; or (3) conservative operation of plant control or safety systems, in a transient-specific fashion.

The generic approximations (Category 1) are applied to all full power transients to account for steady state and instrumentation errors.

The initial DNBR conditions are obtained by adding the maximum steady state errors to rated values as follows:

I 1520 MWt + 2% (30.4 MWt) for Reactor Power

=

calorimetric error.

558.50F - Schedule A Average Coolant Temperature

=

523.50F - Schedule B

+ 40F for deadband and measurement error Primary Coolant System Pressure = 2000 - 30 psia for steady state fluctuation and measurement errors, where Schedule A and Schedule B refer to the TAVE schedules defined in Table 2.1.

The combination of the above parameters acts to minimize the initial minimum DNB ratio.

It should be noted that none of the above steady state errors are explicitly included in the plant transient modeling, except for the loss-of-electric load event, but they are used to conservatively bound the initial MDNBR. Table 2.2 shows a list of operating parameters used in the analysis.

The trip setpoints incorporated into the PTSPWR2 model for R.E. Ginna Unit 1 are based on the Technical Specification limits and have been revised for the changed system conditions. These limiting trip setpoints with their associated time delays for each trip function are listed in Table 2.3.

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XN-NF-82-45

.I Revision 1 The overtemperature and overpower AT trips are calculated from the parameters in Table 2.3 as

'I ATo [K1 - 1+255 K3 (TAVE - TSETPOINT) + K 2 (P r - PSETPOINT) p 1+55

- f (AI)]

and I

105 ATo [K4 - K6 IAVE - K5:(TAVE - TSETPOINT) - f (AI)],

1+10S I

respectively. The function, f(AI), depends on the integrated top to bottom power skew, AI, and exacts a 2% penalty for each 1% that al falls outside the range -18% to 8%.

In the present analysis, the values of the constants K1 to K6 include an allowance for a 4% error in the trip signal calculation. This error allowance was not included in the prior analysis.

The effect of this 4% allowance is to cause the slow rod withdrawal transient to have a calculated MDNBR that is 2.0% lower than the two-pump coastdown transient.

The reference setpoint values for TAVE and PPR were set to the new reduced temperature and pressure values in order that the margin to trip would not be increased for the reduced temperature and pressure conditions.

The ENC fuel design parameters for R.E. Ginna Unit 1 are summarized in Table 2.4.

Table 2.5 lists the neutronics parameter values which conserva-tively bound the R.E. Ginna Unit 1 core for both the beginning and end of cycle, and those used for transient analysis. A design axial power profile with a peaking factor FZ = 1.64 at X/L = 0.6 was used in the analysis.

The approximations in Category 2 refer to the reactivity feedback effects from moderator temperature changes and Doppler broadening of the I

1 10 XN-NF-82-45 Revision 1 I

4 absorption resononces. For full power transients, the moderator temperature drid pr essur e toelI at sents were set to zer o.

Ihis provides o c oriservot ive estimate of the moderator density feedback for all transients f rom f ull

I power. The conservative choice for the Doppler feedback coefficient depends on the transient being analyzed.

Table 2.6 summarizes the moderator and Doppler feedback coefficients applicable for full power transients.

Note that feedback for steam line breaks is treated differently.

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12 XN-NF-82-45 Revision 1 j

i Table 2.1 TAVE Schedules for Operation at 2000 psia i

I Schedule

TAVE @ HZP TAVE @ HFP

[

A 547 558.5 i

8 514**

523.5 lI l

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TAVE varies linearly with power 1

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- This value is conservative with respect to prior estimates l

of approximately 5000F(3) and represents the same decrease in primary cold leg temperature as in Schedule A.

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Table 2.2 Thermal Parameters for Operation of R.E. Ginna Unit 1 at Reduced Average Temperature and Primary System Pressure CORE I

Total Core Heat Output (MW) 1520 Heat Generated in Fuel (%)

97.4 Pressurizer Pressure (psia) 2000 I

HOT CHANNEL FACTORS TotalPeakingFactor,Fh 2.80 EnthalpyRiseFactor,F$H 1.66 SYSTEM PARAMETERS FOR TAVE SCHEDULE A

B TAVE at full power (OF) 558.5 523.5 at zero power 547 514 Total Primary Flow at full power (mlb/hr) 69.2 71.8 Active Core Flow at full power (mlb/hr) 66.0 68.5 Pressurizer level at full power (% of span) 49 49 at zero power 33 36 STEAM GENERATORS Total steam flow (mlb/hr) 6.60 Steam temperature (OF) 503.8 feedwater Temperature (OF) 432.3 I

Steam Dome Pressure at full power (psia) 700 Tube Plugging (%)

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Table 2.3 Reactor Trip Setpoints I

I

_ Function (sec) value A

B Delay Time Tech. Spec.

Schedule High neutron flux (%)

0.5 109 116 116 Low coolant flow (%)

0.6 90 87 87 High pressure (psia) 1.0 2400 2400 2400 Low pressure (psia) 1.0 1880 1700 1700 High Water Level (% of span) 1.0 88 100 100 Low-low S.G. water level

(% of N.R. spen) 1.0 16 0

0 Reactor AT Trips ATo (OF)

Full Power 57.4 58.4 I

Value TSETP0 INT (OF) 573.5 558.5 523.5 PSETPOINT (psia) 2250 2000 2000 K1 1.12 1.165 1.165 Overtemperature

/K 7.356x10-4 6.82 x 10-4 6.82 x 10-4 2

K3

.01577 0.0141 0.0141 K4 1.083 1.134 1.134 Overpower

/K5

.001

.001 K6

.0262

.0273

.0273 HPSI-Actuation (psia) 1738 1730 1730 I

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Table 2.4 Exxon Nuclear Reload for R.E. Ginna Unit 1 l

Fuel Design Parameters

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Fuel Radius 0.1782 inch Inner Clad Radius 0.1820 inch Outer Clad Radius 0.2120 inch tI Active Length 142.0 inch I

i Number of Fuel Rods in Core 21,659 i

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M M -m ' ~M M

m~m M

M Table 2.5 R.E. Ginna Unit 1 Kinetics Parameters Values Used In Transient Analyses Cycle 12 Rod Withdrawal Loss of Steam Line Break LTP Nominal and Loss of Load Primary Flow Transients from Par ameters Value

~ Transients, BOC Transients, B0C HZP, EOC

~

BOC E0C ModeratorTemperagure Coefficient (pcm/ F)

-5.7

-24.8 0

0 Moderator Pressure Coefficient (pcm/ psia)

+.09

+.35 0

0 DopplgrCoefficient (pcm/ F)

-1.4

-1.6

-1.0

-1.5 5

Boron Worth Coefficient

-7 (pcm/ ppm)

-7.9

-C.4 Scram Worth (pcm)

-3467 -4827

-1600

-1600 1600 Shutdown Margin (pcm) 1000 1900 Delayed Neutron Fraction

.0059

.0052

.0059

.0049 Not Applicable to this transient.

2x See Figure 3.32.

7

- - See Figure 3.33.

q sp

I 17 XN.7-32-6 Revisior. 1 I

Table 2.6 Moderator and Doppler Coefficients I

I Desired Moderator Resulting*

Desired Resulting

, I Feedback Coefficient Doppler Coefficient Effect op/0F x 10-5 Feedback op/0F x 10-5 Transient Effect Fast Rod Withdrawal Minimum 0.0 Minimum

-1.0 Slow Rod Withdrawal Minimum 0.0 Minimum

-1.0 Loss of Coolant Flow Minimum 0.0 Maximum

-1.5 Locked Rotor Minimum 0.0 Maximum

-1.5 Loss of Load Minimum 0.0 Minimum

-1.0 Steam Line Breaks Maximum Minimum I

For minimal effect no moderator feedback is allowed

- See Figure 3.32

- - See Figure 3.33 I

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II 18 XN-NF-82-45 Revision 1 3.0 TRANSIENT ANALYSIS 3.1 INITIAL CONDITIONS Evaluation of the effects of the proposed reduced temperature and a

pressure operation on steady state thermal margin was performed in accordance with standard ENC thermal hydraulic calculational methodology described in Refererce 5.

All events were initiated with a pressure of 2000 psia compared to the 2250 psia used in References 1 and 2.

A range of average temperatures

~

was considered from Schedule A to Schedule B of Table 2.1.

References 1 and 2 used TAVE = 573.50F. Since, in all cases, Schedule A was more limiting, due to either initial DNBR or nonlinearity of moderator feedback, Schedule A was analyzed for all transients.

With Schedule A, the initial MDNBR at full power operation,1520 MWt, was 2.07 compared to 2.0(1,2) while Schedule B provided an initial MDNBR of 2.63. The improvements in MDNBR are due to the decrease in TAVE which more than compensates for the 250 psi decrease in pressure.

The improvement in initial DNBR is reflected directly in the MONSR for those events (1,2,3 and

4) initiated from full power.

The overpower and overtemperature AT trip functions were adjusted for the reduced temperature and pressure conditions in order to maintain the margin to trip and thus transients terminated by the AT trip tend to preserve the improvements in margin asociated with the improved initial DNBR.

The analysis results presented in the following subsections include the most limiting rod withdrawals at full power condi-tions, the most limiting loss of flow accidents, which have previously been shown as the worst accidents relative to thermal margin, the loss of load transient, and the steam line break transients, which demonstrate the characteristics of reactor coolant cooldown incidents.

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19 XN-NF-82-45 Revision 1 I

The transient events discussed in this Section are sumarized in Tables 3.1 to 3.8.

3.2 UNCONTROLLED R0D WITHDRAWAL The withdrawal of a control rod bank adds reactivity to the reactar core, causing both the power level and the core heat flux to increase. Since the heat extraction from the steam generator remains relatively constant, there is an increase in primary coolant temperature.

Unless terminated by manual or automatic action, this power mismatch and the resultant coolant temperature rise could eventually result in an unacceptable loss of the thermal margin.

While the inadvertent withdrawal of a control rod bank is unlikely, the reactor protection system is designed to terminate such a transient while maintaining an adequate margin to DNB.

In the rod withdrawal incident the reactor may be tripped by the overtemperature AT function, by the nuclear overpower function, or by another reactor protective safety system setooint. Both a f ast rod withdrawal and a I

slow rod withdrawal were analyzed from an initial power level of 1520 MWt.

Beginning-of-cycle kinetics coefficients were used with a minimum value for Doppler feedback.

Figures 3.1 to 3.6 show plant responses for a fast rod withdrawal (6 10-4 Ap/sec) from 1520 MWt.

A nuclear overpower trip (116% setpoint) x I

occurred at 1.99 seconds. The DNB ratio dropped from an initial value of 2.07 to 1.84.

Pressure increased to a maximum of 2360 psia with core average temperature increasing by less than 2.50F.

Pressurizer control was active during the transient simulation, resulting in lower system pressures and lower calculated MDNBR. The parameters for the two linear control functions I

I

i 20 XN-NF-82-45 Revision 1 are listed in Table 3.9, along with those of the pressure relief valves.

Following the reactor trip, reactivity was inserted via a programmed curve, depicted in Figure 3.17.

The system responses to a slow rod withdrawal of 5 x 10-5 A vsec are depicted in Figures 3.8 to 3.13.

The overtemperature AT function initiated the reactor trip at 24.27 seconds, and the minimum DNB ratio during the transient was 1.7.

Sizing of the parameters in the overtemperature AT trip function is such that for a reactivity insertion of 5 x 10-5ap/sec it is nearly coincident with the nuclea overpower trip (116%).

For the rod withdrawal accidents at reduced coolant temperature condition, power peaking increases about the withdrawn rod are not expected to be more than a few percent, and these are more than offset by the significantly improved DNBR resulting from lower TAVE-3.3 LOSS OF COOLANT FLOW The loss of coolant flow transient is postulated to occur as a result of a loss of electric power to the primary coolant pumps.

The transient results in an increase in coolant temperature which, in combination with the decrease in flow, reduces the margin to DNB. Only the limiting case has been analyzed.

This case is the loss of power to both pumps when the reactor system is operating at 1520 MWt.

Beginning-of-cycle values of kinetics coefficients are assumed with a conservative choice for the Doppler coefficient.

The loss of power to all pumps would ordinarily result in a reactor trip due to either under-voltage or under-frequency at the bus. No credit was taken for these protective functions and the trip allowed to occur on a low flow signal. This delay resulted in a further flow reduction at full power, and a more conservative calculation of margin to DNB. The pressurizer pressure control was retained to provide a more conservative MDNBR.

21 XN-NF-82-45 Revision 1 Figures 3.14 to 3.19 depict plant responses after the loss of all pump power. A reactor trip occurred at 3.57 seconds. A minimum DNB ratio of 1.74 was reached 4.45 seconds af ter the beginning of coastdown.

System pressure peaked at 2355 psia.

3.4 LOCKED ROTOR In the unlikely event of a seizure of a primary coolant pump, flow through the core would be drastically reduced, resulting in a reactor trip on a low flow signal.

The coolant enthalpy would rise, thus decreasing the margin to DNB.

The locked rotor transient was analyzed assuming two loop operation with instantaneous seizure of one pump from 1520 MWt. Beginning-of-cycle kinetics coefficients were used as the B0C moderator coefficient is the most adverse, A conservative value for Doppler feedback was used and pressurizer pressure control was retained.

The responses for the locked rotor transient are shown in Figures 3.20 to 3.25.

The reactor trip occurred at 0.64 seconds on the low flow function. Core average temperature increased by 18.7cF with system pressure reaching 2384 psia. The DNB ratio in the analysis reached a minimum of 1.26 at 1.85 seconds, and recovered the initial DNBR of 2.07 by 3.75 seconds. The total exposure of the core to a DNBR of 1.3 or less was approximately 1 second.

3.5 LOSS OF ELECTRIC LOAD Loss of electric load involves plant behavior following a trip of the turbine-generator without a direct reactor trip.

The major consequence of the loss of heat sink is a rapid increase in TAVE and an associated rise I

in pressurizer level and pressure due to expansion of the primary coolant.

Conceivably DNBR could be a problem, since rising temperatures adversely I

22 XN-NF-82-45 Revision 1 I

affect thermal margins. However, for R.E. Ginna Unit 1, the pressurization transient is sufficiently strong even with pressurizer pressure control functioning to cause an increase in DNBR with increasing TAVE.

Thus the purpose for evaluating this transient is to assess peak pressure versus the vessel integrity limit of 2750 psia.

In calculating the peak pressurizer pressure, spray control was turned off, and the power operated relief valves (PORVs) disabled.

The transient was initiated from 1550.4 MWt (102% power) and 2030 psia, with a conservative Doppler feedback and no moderator feedback.

The reactor tripped on a high pressurizer pressure signal at 4.81 seconds and peak pressure was 2520 psia at 6.89 seconds.

DNBR never dropped below the initial value. Figures 3.26 to 3.31 show the system responses to a loss of I

electric load.

Control of maximum pressure was exercised by the safety relief valves whose capacity far exceeds the rate at which the coolant can expand.

After 10 seconds the pressurizer was only 68% full (up from 49%

operating level) and there was no danger of " packing" the pressurizer.

3.6 STEAM LINE BREAKS A break of a steam pipe (or safety valve failure) would result in a sharp reduction in steam inventory in a steam generator. This pressure decrease, which accompanies the loss of heat via ejected steam, would cause a heat loss from the primary coolant, reducing primary coolant temperature and pressure. With a negative moderator temperature coefficient, the reduced temperature would lead to a reactivity insertion into the core which could it>ad to criticality and core damage if unchecked.

Steam line break transients are simulated with the PTSPWR2 plant transient simulation code. As a worst case, the steam line break was assumed to occur at hot zero power conditions corresponding to a core average

1 I

23 XN-NF-82-45 Revision 1 I

temperature of 5470F when the steam generator secondary side water inventory was at a maximum, thus prolonging the duration and increasing the magnitude of the primary loop cooldown.

For conservatism, the most reactive control rod was assumed to be stuck out of the core when evaluating the shutdown margin of the control rods.

The 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 3.32 and 3.33.

A shutdown margin of 1.6% was used for

.onservatism.

Table 2.5 summarizes the kinetics parameters used in steam line break analysis. Minimum capability of the boron injection system, which is based on two of three high pressure safety injection (HPSI) pumps being available, was assumed.

A low pressurizer pressure signal initiated HPSI.

The entry of borated water at 20,000 ppm into the primary loop cold legs was delayed by the necessity of first sweeping the injection lines of low concentration borated water. The delay time is dependent on the difference between the pressure and the pump shut-off head,1400 psia.

Initial system pressure is not an important factor in this analysis, since depressurization of the primary loop to the low pressure trip setpoint occurs rapidly once the pressurizer empties and the time is essentially independent of initial system pressure.

The effects of initial temperature are discussed for the large steam line break.

Flow from steam line breaks was calculated based on a fixed break area and the Moody curve (8) for choke flow. The break area corresponded to a double-ended guillotine rupture of the steam line at the exit of the steam generator in the large break analysis. The small break analysis represented a failed safety relief valve.

I

24 Xf:-t:F-82 t5

,,rcisica _

3.6.1 Larg_e_ Steam Line Break For the large break the steam flow was calculated from the Moody curve for critical flow of saturated steam, based on the flow area for I

a break inside of containment.

Initially the intact steam generator also blew down until the main steam isolation valve closed.

This case, which retained pump power, was shown to give the greatest retura to power (9).

I Figures 3.34 to 3.39 show the transient response for a large steam line break initiated from 5470F (Schedule A). A conservative choice of shutdown margin (1.6%) was used although the Technical Specification limit is I

1.9%.

The initial steam flow (8615 lb/second) induced a rapid cooldown to

~4000F. Accompanying this cooldown was a rapid rise in moderator reactivity such that the reactor went critical at 16 seconds and reached a peak power of 685 MWt, or 45.1% of 1520 MWt, before the borated water shut the reactor down.

The HPSI signal was received at 6.6 seconds on the emptying of the pressurizer. By 16.6 seconds, HPSI had occurred. The core parameters at the time of peak power were outside the range of validity for the W-3 correlation and a Modified MacBeth Correlation with a conservative local hot rod peaking factor, Fh, of 14 was used. Ine minimum critical heat flux (CHF) ratio of 1.1 l

occurred at 43.8 seconds.

Less than 1% of the fuel rods underwent boiling i

transition.

l Figures 3.40-3.45 show the transient response for a large steam line break initiated from 5140F (Schedule B). The steam flow for this case was based on the Moody curve as with Schedule A.

Because of reduced steam pressure at saturation, the initial flow is significantly less (5848 lbs/sec).

A rapid cooldown to ~ 4000F occurred following the break.

1 Accompanying the cooldown was an increase in moderator reactivity which was

! I

I 25 XN-NF-82-45 I

Revision 1 less than the Schedule A insertion af ter a large break.

The moderator reactivity as a function of temperature is shown in Figure 3.33. Because of the curvature, a change of 500F from a TAVE of approximately 5500F produced about a 30% larger change in reactivity than a change of 500F from 5000F. In fact, it requires about a 700F change from 5000F to produce the same change in reactivity that 500F produces from 5500F. The reactor went critical at approximately 24 seconds and reached a peak power of 271.9 MWt, or 17.9% of 1520 MWt. HPSI occurred at 18.4 seconds with the borated water shutting the reactor down at 41.7 seconds. The MCHFR was 3.21 for this transient.

The transient return to power from HZP on Schedule B was less than an Schedule A.

Similarly, because of lower steam pressures and the shape of the moderator reactivity curve, the tendency to return to power on Schedule B will be less for all steam line breaks.

3.6.2 Small Steam Line Break The small steam line break summarized in Figures 3.46 to 3.51 corresponds to a failed safety relief valve with a capacity of 228.5 ibs/sec at 1100 psia, during single loop operation on Schedule A. The shutdown margin of 1.6%, compared to the end-of-cycle Technical Specification limit of 2.4%

for one loop operation, prevented the reactor from returning to power.

No HPSI signal was generated because the pressurizer never emptied. As with the large break, Schedule B results, not shown, displayed even less of a tendency to return to power.

I I

I

I 26 XN-NF-82-45 Revision 1 I

Table 3.1 Uncontrolled Rod Withdrawal (Fast) j Event Table Time (sec.)

Event Value 0

Initiate Reactivity Insertion 6 x 10-4 Ap /sec.

j 1.99 Nuclear Overpower Trip 116%

I 2.11 Peak Power Level 1861.5 MWt

'I 2.95 Minimum DNB Ratio 1.84 3.11 Peak Heat Flux (average) 191,492 Btu /hr-ft2 6.13 PORV Opened 2350 psia 6.46 Peak Pressurizer Pressure 2360 psia I

I l

I 1 I i I I

I

27 XN-NF-82-45 Revision 1

,I l

Table 3.2 Uncontrolled Rod Withdrawal (Slow)

Event Table Time l

(sec.)

Event Value O

Initiated Reactivity Insertion 5 x 10-5

/sec.

{

24.27 OvertemperatureJST trip 115%

!l 24.28 Peak Power Level 1747.3 MWt l

24.40 Minimum DNB Ratio 1.7 24.46 Peak Heat Flux (average) 198,802 Btu /HR-ft2 l

l 28.53 PORV Opened 2350 psia I

37.13 Peak Pressurizer Pressure 2366 psia lll i

1 1

1 28 XN -NF-82-45 Revision 1 Table 3.3 Loss of Coolant Flow Event Table t

Time (sec.)

Event Value 0

Loss of Pumping Power 3.57 Low Flow Trip 87%

i 4.45 Minimum DNB Ratio 1.74

)

4.96 Peak Core Temperature (average) 5630F 7.26 PORV Opened 2350 psia i

7.45 Peak Pressurizer Pressure 2355 psia 1

4 I

I

I i I 4

)I

!I 4

I 29 XN-NF-82-45 I

Revision 1 I

I Table 3.4 Locked Rotor - Event Table I

Time (sec.)

Event Value O

Pump Seizure in loop 1 0.64 Low Flow Trip 82%

1.85 Minimum DNB Ratio 1.26 3.08 PORV Opened 2350 psia 3.60 Peak Pressurizer Pressure 2384 psia lI

'l I

lI I

I I

l 30 XN-NF-82-45 Revision 1 Table 3.5 Loss of Electric Load i

Event Table Time l

(sec.)

Event Value O

Loss of all Electric Load i

r 3.81 High Pressure Trip Signal Generated 2400 psia 4.39 Pressurizer Safety Valve Opened 2500 psia 4.81 High Pressure Trip I

6.89 Peak Pressurizer Pressure 2526 psia

!!l iI ill 4

l 4

i i

'I iI i

ll

,!l ii i

I 31 XN-f4F-82-45 Revision 1 i

i Table 3.6 Large Steam Line Break (5470F)

Event Table i

Time j

(sec.)

Event Value i

0 Double-Ended Break in Loop 1 i

l 5.00 MSIV Closed on Loop 1 6.61 Low Pressurizer Pressure HPSI signal generated 1698 psia 15.77 Reactor went Critical 16.60 HPSI Pump reached Operating Head 1400 psia 42.70 Boron Entered the Loop l

42.72 Peak Power Level 685.2 MWt 43.75 Minimum DNB Ratio

1.10 43.83 Peak Heat Flux (average) 74,166 Btu /hr-ft2

Calculated with modified MacBeth correlation i

I I

I

!I 32 XN-NF-82-45

,;a Revision 1 g

1 Table 3.7 Large Steam Line Break (5140F)

Event Table l

I Time (sec.)

Event Value 0

Double-Ended Break in Loop 1 5.00 MSIV Closed on Loop 1 8.42 Low Pressurizer Pressure HPSI signal generated 1697 psia 18.41 HPSI Pump reached Operating Head 1400 psia i

j 24.30 Reactor went Critical 41.71 Boron Entered the Loop 41.82 Peak Power Level 271.9 MWt 45.00 Minimum DNB Ratio

3.21 lI 45.06 Peak Heat Flux (average) 27,937 8tu/hr-ft2 i

i l

Calculated with modified MacBeth correlation I

'I i

llI 1,I rI l

J A

i=

33 XN-NF-82-45 Revision 1

!I

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Table 3.8 Small Steam Line Break Event Table

!I Time j

(sec.)

Event Value 0

Steam Line Safety Valve Opened 168.75 Maximum Reactivity

-50.84 1I

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I

I 34 XN-NF-82-45 1

Revision 1 Table 3.9 Pressure Control and Safety Parameters I

4 Value Used in PTSPWR2 lI j

PRESSURIZER SPRAY 1 g On, psia 2048

{g i

I Off, psia 1998 4

Maximum, lbs/sec 44.98 HEATER l

Pressure full on, psia 1981 Pressure full off, psia 2011

)

i Maximum, kw 800 l

1 PORVs Setpoint, psia 2350 Total Capacity, lbs/sec 99.44 I

SAFETY RELIEF VALVES Setpoint, psia 2500

I i

Total Capacity, lbs/sec 160 STEAM GENERATOR SAFETY RELIEF VALVES Setpoint, psia 1100 A 1118 B 1136 C

,I 1154 D j

Valve Capacity, lbs/sec 228.45 i

RELIEF VALVE

]

Setpoint, psia 1065 (open) 777 (close)

Total Capacity, lbs/sec 228.45

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40 XN-NF-82-45 Revision 1 I

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I

-1600 i

i N_

3 3

Ty I

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e_..n

~b 20 40 FO 80 100 120 140 100 ISC 20:

"7 i

o cn wa TIME, GEC I

CEO RECr?CC 2 2.' O E / 3 ?

21..~.43

't Figure 3.37 Large Steam Line Break (547 ) -

Pressure Changes i

t

I l

t l

l l

1 f

I i

100 1.

CHANE IN 3 TEAM GEN. =ATER LEVEL, LC?P t z.

CHEE IN OTEAM LEN. WATER LEVEL. LGJP c 3<

CHCNOE IN PRESCLSIZER 'VATE!P ' F VEL i

0 i

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2 z

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20 40 60 80 100 120 14C lt:0 180 200 w a' b

i sm TIME. SEC i

i,E O - FEGr7Ge Figure 3.38 Large Steam Line Break (54/o) cz N < ?^

01.26.43 l

Level Changes

E E

M E

E E

N E

E E

9 1

MODERG"CR PEGCTILITY 2

DGPPLE1; PEACTIVIY

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9 0R0tJ ltCACTIVITY 4.

TOTAL liEACTIVITY b

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x \\

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=

r*

5' $1 6 0 20 40 EO 90 100 320 140 160 180 200 A

TIME, SEC cto gcr7cc ez/cu sa 21 1s.43-Figure 3.39 Large Steam Line Break (547 ) - Reactivity

700

2..

Pao i evEL 2

HEAT Fl. t;X 3

TCTPL 1:RIMPRY t'OrmANT i C4 t

L TOTAL I EECWATER FLCW TOTAL 1:TEDM LINC FLGV 600 4

500 Q

4 t-CL E400 tt O

i

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i t-A

/

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[2.

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100

{

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s

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0 "N

O P0 40 FC 80 100 120 140 1GO 180 200 A

TIME, SEC f,E C DECC'KR 22.'00/32 21.00.02 i

Figure 3.40 Large Steam Line Break (514 ) - Power, Heat Flux and System Flows

~... -. -...

. m.

._..___________-._._._...m._

mM M

M M

I i

SE 0 i 2

ovc. ti,ct acerca;Ttwc l

2 CORC IlJL ET TLMPCP9TbPC 3

C'. E. CCRE COCLAN" TEMP.

g 4.

CLOS TLMPERGTbRE c,..

swU-

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EC 40 60 EC 100 120 140 lt:0

90 200

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~n TIME. SEC SEC REGC/KP L2/Or/si 22.00.55 Figure 3.41 Large Steam Line Break (514 ) - Core Temperature Response 1

M M

M W

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M i

i 1

100 1.

caps:c :n avc. ni>1r# 3 c cc:.an: TEMP.

toce i 2.

Cr@NTE IN PVC. F'lFIP94R Y CCC;J,NT TDiP.

LOOP 2 3.

CtiANCE IN HGT LEG COLD Lt.C TEMP. JTF ERENCE. LOOP.

4.

Ct@NOC IN HGT LEG-C OLO LI.C TEMP. JTFERENCE.

LOOP -

-4 50 w

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40 160 18C 200 b$

TIME. SEC EEC RCC-c/kR 22/GC/32 22 J0.5S.

Figure 3.42 Large Steam Line Break (514 ) - Primary Loop Temperature Changes

_. _ _. _ - -.. - _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _. ~. _... _. _. _. _ _

m m

m M

M M

M 6N W

N l

300 j

1 STEAM CGME FPESSiJcE CHetCE, LOCP 1 2

STEPM DOME PRECDeRE CHONCE.

LGC? 2 3

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SC 100 120 140 160

]90 E00 i

TIME, C E t' CEG REGr7kP 22/or/32 21 00.5s.

0 Figure 3.43 Large Steam Line Break (514 ) - Pressure Changes

l 100 1

cHan;E IN ClLAt4 GEN. MATER LEVEL. LOCP 1.

2 CHANCE IN OTEPM EEN. WATER LEVEL, LOPP d 3.

CHPNCE IN PRESCLO!ZER ' MATED LEVEL 0

7 2

2 2

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m TIME, E.EI' CEg.REGr7KR 22.'cc/97 21.00.50.

Figure 3.44 Large Steam Line Break (514 ) - Level Changes

M M

M l

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3 MODERWjCRREACTIbITy e

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BCRCN REPCTIVITY 4.

TOTAL IFEACTIVITY E

1 i

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)[

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TIME, SEC SEC PEGr?kR 22/06/92 21.00.5s.

Figure 3.45 Large Steam Line Break (514 ) - Reactivity

W M

M M

m b

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Figure 3.46 Small Steam Line Break (S47 ) - Power, Heat Flux and System Flows

aus em as aus em an em amo ma em um aus me aus em em man em om l

C (r <

l

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10 88 Figure 3.47 Small Steam Line Break (547 ) - Core Temperature Response

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~-

w

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Cr4PN E tJ 4GT LE COLL LI:G TEMP..TFFERENCE.

LCCP c

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TIME. iEC IEC-REGCE.TG za/CE!P?

2C 40-20 Figure 3.48 Small Steam Line Break (547 ) - Primary Loop Temperature Changes

_ _ _ _ _ _. _ _ _.. ~. _ _ _ _ _ _...

j m

m m

m M

M M

.i i

i l.

i i

j E.00 1,

sTLar7tt t esEssi,PE CH?tGE LUCP 1 2

STEAM LCHE PRESSI.PE CH4tCE torP t t

3.

PPESSU:CZER FREsEb?E CHPrCI:

i.

n K'

2 Q

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3;1 x

x NP

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/

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1

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0.

Figure 3.49 Small Steam Line Break (547 ) - Pressure Changes i

1

I j

U E

E E

E M

f i

i I

i 1C3-ic cHatcc In sTEpM cca.pTes tevct. LccP 2 2

CHOtJCE IN ';T L AM GE:J. =PTED LEVEL, LO:' P t.

1 C HptCE IN PRESSUHIZERWATEyLEVEL O

w, e

~

?

\\

N, N

I L-

\\

N-L

-100 l

i I

l

,,-2Ca x.

\\

I w

1 N

'e Nl 2

H s

-300 4;

N

'N N

-4 0:;

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o x 1

1 1

_.2 w m.

I J

o cn a na n g,,.

40 SC 32G ISO 200 240 P9C 320 360 400 v

wu*

TIME, SEC rt PEcce ;

ca<ur'sc 20.4s._c.

Figure 3.50 Small Steam Line Break (54/e) - Level Changes 1

1 I

ii l

l

,I l

e aw x=a2,ec~Lw a

e xm<.w oo -

. m 0

C C

4 3

m 4

G 2

=

0 6

m 2

?

m e

C_

S

=_

c 2

. y 2.

i t

v i

m tc a

,li-r

. a e

I R

r G

E R

)

?

7 r

C 4 T

0 L 5

. m I 'Y 4

G

(

n '

2 IIY Y T VT T k

4 cIII C

a p T V V e

ECII E

r RPT T G

B M

' i i i M

i ECC R R PA 0

e G

t ET 0

n P

2 E i

GE R L N M

L EPCN I

T m

3PR 0G0 a

MJ9 e

a t

O S

E M

1 4

l 1 2 3 l

a am a

S z

0 M

. 2

+

'a 1

S.

3

/

e

- M a

ru g

1

?l l

V 1

0 i

3 F

. M 2

,/

0

/

4

- M 4

2

/e

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M 6l i1 i!I

. C 1

3.'

1 3

2 0

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- M nmcJJo _.

c c

. M il l]

1:

)

I 86 XN-NF-82-45 5

Revision 1

4.0 CONCLUSION

S The transient analyses for the R.E. Ginna Unit 1 Nuclear Power Plant for conditions of reduced primary coolant temperature, Schedule A, and pressure show adequate margin to safety limits.

The neutronics data used in these analyses are consistent with, or conservative with respect to the previous analysis (3).

For reduced primary coolant temperature and pressure the limiting transient analyses reported in Section 3 showed generally increased margins when compared to the previous analyses (1,2).

Several additional transients commonly analyzed were not treated in Section 3.

These included:

startup of an inactive loop loss of feedwater RCCA drop loss of A.C. power chemical and volume control system malfunction reduction in feedwater enthalpy accident.

I They were not limiting transients in prior analyses (see References 1 and 2) and should remain non-limiting for reduced temperature and pressure since their rate of reactivity insertion is enveloped by the limiting transients discussed in Section 3.

Further, the steady state MONBR increased with reduced temperature and pressure, contributing additional margin.

Two average temperature schedules, which bound the proposed operating range for R.E. Ginna, were treated in the analysis.

The lower schedule, B, was shown to be bounded by the higher schedule, A, because of improved initial DNBR for transients from power and because of the shape of the moderator reactivity curve for steam line breaks.

I I

I 87 XN-NF-82-45 Revision 1

!I The transient analysis supports operation of the R.E. Ginna Unit 1 l

nuclear power plant at reduced temperature and pressure using the trip i

setpoint parameters listed in Table 4.1.

I Operation of R.E. Ginna at reduced temperature and pressure is more e

favorable from a plant transient point of view. A reduction of 250 psi in pressure is more than compensated for in the DNBR by a decrease in the average temperature of 150F. Further reductions in average temperature will improve thermal margins in all transients from power and decrease the reactivity insertion in steam line breaks.

i!I I

I l

!I

!I I

I lI 1I lI

lI l

88 XN-NF-82-45 j

Revision 1 i.

J Table 4.1 PPS Trip Settings for Reduced T, P l

l Function Value Supported by Analysis i

Low Pressure Trip (psia) 1730 i

Reactor 4T Trips

{

aTo (OF)

Full power value i

T*setpoint (OF) 523.5-558.5 l

Psetpoint (psia) 2000 K1 1.12 6.56 x 10-4 K2 K3 0.0136 l

K4 1.09 i

K5 0.001 K6 0.0262 lI g

equal to plant average temperature I

I 1

I

1 I

89 XN-NF-82-45 Revision 1 I

5.0 REFERENCES

1)

" Plant Transient Analysis for the R.E. Ginna Unit 1 Nuclear Power I

Plant", XN-NF-77-40, Rev. O, Exxon Nuclear Company, Inc., Richland, Washington 99352, November 1979.

2)

" Plant Transient Analysis for the R.E. Ginna Unit 1 Nuclear Power I

Plant", XN-NF-77-40, Supplement 1, Exxon Nuclear Company, Inc.,

Richland, Washington 99352, March 1980.

3)

" Reduced Primary System Temperature and Pressure Analyses for R.E.

Ginna Nuclear Power Unit", XN-NF-80-44, Exxon Nuclear Company, Inc.,

Richland, Washington 99352, September 1980.

4)

" Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTSPWR)", XN-74-5, Revision 1, Exxon Nuclear Company, Inc., Richland, Washington 99352 5)

" Definition and Justification of Exxon Nuclear Company DNB Correlation for Pressurized Water Reactors", Xh-75-48, Exxon Nuclear Company, Inc., Richland, Washington 99352, October 1975.

6)

MacBeth, R.V., " Burnout Analysis Part 5.

Examination of Published World Data for Rod Bundles", AEEW-R-358, June 1964.

7)

MacBeth, R.V., "An Appraisal of Forced Convection Burn-out Data",

Proc. Inst. Mech. Engs., Vol. 18, Part 3E, 1965-1966.

8)

Moody, F.J., ASME Transactions, p. 134, February 1965.

9)

" Rochester Gas and Electric Corporation Proposed Change to Shutdown Margin Requirements", Ginna Nuclear Plant Technical Specification, Amendment Application dated September 22, 1975.

I I

il l

PLANT TRANSIENT ANALYSIS FOR OPERATION OF THE XN-NF-82-vf RE GINNA UNIT 1 NUCLEAR POWER PLANT AT REVISION 1 REDUCED PRESSURE AND TEMPERATURE DISTRIBUTION ISSUE DATE: 07/22/82 l

lI i

I F.T. Adams i

G.J. Busselman G.C. Cooke N.F. Fausz L.J. Federico T.J. Helbling il l5 S.E. Jensen Ilg W.V. Kayser il3 R.H. Kelley M.R. Killgore J.E. Krajice'K T.R. Lindquist j

J.N. Morgan W.T. Nutt G.F. Owsley 1

l F.B. Skogen l

G.A. Sofer i

C.H. Wu l

i RG&E/T.J. Helbling (80)

I Document Control (10) lI i

.

ML20062K429

Contents

Text

4.0 CONCLUSION

5.0 REFERENCES

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

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4.0 CONCLUSION

5.0 REFERENCES

SUMMARY

4.0 CONCLUSION

5.0 REFERENCES

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