Text

Plant Transient Analysis
	ML19276D708
Person / Time
Site:	Prairie Island
Issue date:	11/30/1978
From:	Markowski F; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML19276D707	List: ML19276D706; ML19276D708; ML19276D709; ML19276D711;
References
	XN-NF-78-035, XN-NF-78-35, NUDOCS 7901090219
	Download: ML19276D708 (76)
	v • d • e

I XN NF-78 35 I

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I PLANT TRANSIENT ANALYSIS FOR THE PRAIRIE g

ISLAND NUCLEAR POWER PLANT UNITS 1 & 2 I

I NOVEMBER 1978 g

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RICHUND, WA 99352 I

E(ON NUCLEAR COMPANY,Inc.

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wase

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ISSUE DATE:

12/01/78 XN-NF-78-35 PLANT TRANSIENT ANALYSIS FOR THE PRAIRIE ISLAND NUCLEAR POWER PLANT UNITS 1 & 2 By F. J. MARK 0WSKI IIll?l10 Approved: ((N< tmh a-29 -18 K. P Calbraith, Manager Nucle Safety Engineering Approved:

[N

-M-N I

W Mgdager G. A. S@ Fuels Er.gineering Nuclear Approved:,#/

///ido 2

/G' J. BtSselman, Manager Contract Performance Approved:

[

// 39 b'

E S. Nechodom, Manalgbr

/

I Licensing and Compli%nce I

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NON-PROPRIETARY ERON NUCLEAR COMPANY,Inc.

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I IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and develop ent proarams st onsored by Exxon Nuclear Company, Inc. It is being subrrtted by Exxon Nuclear to the USNEC as part of a technical contribution to facilitate safety analyses by licensees ofr the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and con e'.t to the best of Exxon kuclear's knowledge, information, and belief. Tne 3

information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.

Without derogating f rom the foregoing, neither Exxon Nuclear nor any person acting on its behalf:

A.

Makes any warranty, express or implied, with respect 3

to the accurocy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, nothod, or process disclosed in this document will not infringe privately owned rights; or B.

Assumes any liabilities with respect to the use of, or for damages resalting from the use of, any information apparatu<. raethod, or process disclosed in tt document.

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XN-NF-78-35 TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 CALCULATION METHODS AND INPUT PARAMETERS..........

4 3.0 TRANSIENT ANALYSIS....................

13 3.1 FAST CONTROL R0D WITHDRAWAL.............. 13 3.2 SLOW CONTROL R0D WITHDRAWAL.............. 14 3.3 LOSS OF REACTOR C0ulANT FLOW 14 3.4 LOCKED PUMP ROTOR...

15 I

3.5 LOSS OF EXTERNAL ELECTRIC LOAD 16 3.6 LARGE STREAMLINE BREAK 17 3.7 SMALL STREAMLINE BREAK 19 4.0 DISCUSSION OF RESULTS 64 73 5.0 SIMULATION CODE CHANGES 76

6.0 REFERENCES

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-ii-XN-NF-78-35 LIST OF TABLES Table Page 1.1

SUMMARY

OF RESULTS........................

3 2.1 PARAMETER VALUES USED IN PTSPWR2 ANALYSIS OF PRAIRIE ISLAND UNIT 1 & 2...................

9 2.2 PRAIRIE ISLAND UNIT 1 & 2 TRIP SETPOINTS.............

10 2.3 PRAIRIE ISLAND UNIT 1 & 2 FUEL DESIGN PARAMETERS FOR EXXON NUCLEAR FUE' 11 2.4 PRAIRIE ISLAND UNIT 1 & 2 ENG KINETIC PARAMETERS........

12 4.1 COMPARISON OF TRAN5!ENT-SPECIFIC INPUT P A RAE TE R S............................ 69 4.2 COMPARIS0N OF OPERATING PARAMETERS FOR PRAIRIE ISLAND UNIT 1 & 2....................

70 I

4.3 COMPARIS0N OF PRAIRIE ISLAND KINETIC PARAMETERS........................

72 5.1 C6MPARIS0N OF RESULTS FOR THE R. E. GINNA PUMP I

SEIZURE TRANSIENT 74 5.2 LIST OF CODE VARIABLES REMOVED FR01 INPUT OR REDEFINED...... 75 I

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LIST OF FIGURES I

Figure Page 2.1 PTSPWR2 System Model 7

2.2 Reactor Shutdown Reactivity Curve used in the PTSPWR2 Analysis 8

3.1 Power, Heatflux and System Flows for Fast Control Rod Withdrawal 20 3.2 Core Temperature Response for Fast Control Rod Withdrawal 21

3. 3 Primary Loop Temperature Response for Fast Control Rod Withdrawal 22 I

3.4 Pressure Changes in Pressurizer and Steam Generators for Fast Control Rod Withdrawal 23 I

3.5 Level Changes in Pressurizer and Steam Generators for Fast Control Rod Withdrawal 24 3.6 Minimum DNB Flux Ratio for Fast Control Rod I

Withdrawal 25 3.7 Power, Heatflux and System Flows for Slow Control Rod Withdrawal 26 3.8 Core Temperature Response for Slow Control Rod Withdrawal 27 3.9 Pri. nary Loop Temperature Response for Slow Control Rod Withdrawal 28 3.1 ) Pressure Changes in Pressurizer and Steam Generators for Slow Control Rod Withdrawal 29 3.11 Level Changes in Pressurizer and Steam Generators for Slow Control Rod Withdrawal 30 3.12 Minimum DNB Flux Ratio for Slow Control Rod Withdrawal 31 I

3.13 Power, Heatflux and System Flows for Coolant Pump Trip 32 I

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-iv-XN-NF-78-35 3.14 Core Temperature Response for Cool.nt Pump Trip 33 I

3.15 Primary Loop Temperature Response for Coolant Pump Trip 34 3.16 Pressure Changes in Pressurizer and Steam Gene.ators for Coolant Pump Trip 35 3.17 Level Changes in Pressurizer and Steam Generators for Coolant Pump Trip 36 3.18 Minimum DNS Flux Ratio for Coolant Pump Trip 37 3.19 Power, Heatflux and System Flows for Coolant Pump Seizure 38 3.20 Core Temperature Response for Coolant Purp Seizure 39 I

3.21 Primary Loop Temperature Response for Coolant Pump Seizure 40 3.22 Pressure Changes in Pressurizer and Steam Generators for Coolant Pump Seizure..........

41 3.23 Level Changes in Pressurizer and Steam Generators for Coolant Pump Seizure................

42 3.24 Minimum DNB Flux Ratio for Coolant Pump Seizure 43 3.25 Power, Heatflux and System Flows for Turbine Trip 44 3.26 Core Temperature Response for Turbine Trip 45 3.27 Primary Loop Temperature Response for Turbine Trip 46 3.28 Pressure Changes in Pressurizer and Steam Generators for Turbine Trip 47 3.29 Level Changes in Pressurizer and Stean Generators 5

for Turbine Trip 48 5

3.30 Minimum DNB Flux Ratio for Turbine Trip 49 3.31 Variation of Reactivity with Power at Constant Core Average Temperature 50 3.32 Variation of Reactivity with Core Average Temperature at the End of the Cycle 51 I

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-v-XN-NF-78-35 3.33 Power, Heatflux and System Flows for Large Steamline Break 52 3.34 Core Temperature Response for Large Steamline Break 53 3.35 Primary Loop Temperature Response for Large Steamline Break 54 I

3.36 Pressure Changes in Pressurizer and Steam Generators for Large Steamline Break 55 3.37 Level Changes in Pressurizer and Steam Generators I

for Large Steamline Break 56 3.38 Nuclear Reactivity Fzedback Effects for Large I

Steamline Break 57 3.39 Power, Heatflux and System Flows for Small Steamline Break 58 3.40 Core Temperature Response for Small Steamline Break 59 3.41 Primary Loop Temperature Response for Small Steamline Break 60 3.42 Pressure Changes in Pressurizer and Steam Generators for Small Steamline Break 61 3.43 Level Changes in Pressurizer and Steam Generators for Small Steamline Break 62 3.44 Nuclear Reactivity Feedback Effects for Small Steamlim Break 63 I

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I XH-NF-78-35

1.0 INTRODUCTION

AND SUMMAR,Y The Cycle 5 extended burnup reload of the P airie Island Nuclear Power Plant with Exxon Nuclear fuel results in core parameter values only slightly different from previous cycle values. The only significant difference is a slightly positive moderator temperature feedback coefficient at low power operation during the initial part of the cycle. The positive moderator coefficient results from the extended burnup requested for Prairie Island Cycie 5.

The coefficient is positive at low power, approaches

-6 zero at about 70 percent power and is calculated to be -36.9 + 20 x 10 fp at full power for beginning of Cycle 5.

The reload fuel design has been shown to be both neutronically and hydraulically compatible with the existing fuel, and thus, the system response during plant transients would not be expected to be particularly sensitive to the fuel type. To demonstrate that the reload fuel meets plant regulatory requirements during design basis events, the most limiting transients identified for the existing fuel were reanalyzed with Eyr'n Nuclear fuel using the Exxon Muclear plant transient simulation code M SPWR2.0 ) This report presents the results of the analysis of the following design basis events, as well as the input parameters used to simulate the reactor system.

Event Incident Clas

1.

Fast Control Rod Withdrawal II 2.

Slow Control Rod Withdrawal II 3.

Loss of Power to Both Reactor Coolant Pumps III 4.

Locked Rotor in One Reactor Coolant Pump IV 5.

Loss of Electric Load II 6.

Large Steam Line Break IV 7.

Small Steam Line Break IV Consistent with current FSAR incident classification for PWR's.

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Events 1 through 5 are initiated from full power, while events 6 and 7 are initiated from hot standby conditions. The criteria to be satisfied in the Class II and III full power events are a peak system pressure of < 2750 psia and a Minimum Departure from Nucleate Boiling Ratio (MDNBR) of 1 1.30 based on the W-3 correlation.(2)

The criterion for the steam line break is that the end-of-cycle shutdown margin be adequate to ensure (1) the design thermal margin, MDNBR 1 1.30 for the large break, and (2) that the core does not become critical from hot standby following a small break.

The analysis is based on an equilibrium ENC fueled core using conserva-tive neutronic parameters calculated for ENC fuel.

The results of the analysis are sumarized in Table 1.1.

The lowest MDNBR for Class II and III events was 1.87, which is above the acceptable minimum of 1.30.

The locked rotor incident, a Class IV event, was analyzed and the MDNBR was found to be 1.09.

This result I is acceptable for this low probability incident. The peak pressure criterion for the reactor coolant system was met in all cases.

The small steam line break analysis showed that the smallest expected shutdown margin at the end of Cycle 5 is adequate to prevent return to criticality during such an event.

T The analysis is valid for a maximum power pecking factor of F = 2.32 and an axial power peaking factor of F = 1.45, with the axial peak located at X/L < 0.60.

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TABLE 1.1

SUMMARY

OF RESULTS Maximum Maximum Maximum Core Average Pressurizer Transient Power Level Heat Flux Pressure MDNBR 2

(Class')

(Percent)

(Btu /hr-ft )

(psia)

(W-3)

Initial Conditions For Transients 102 194,790 2220 2.32 Fast Control

-3 Rod Withdrawal (II)(10 /sec) 134 213,870 2229 1.97 Slow Control

-6 Rod Withdrawal (II)(25x10 /sec) 112 210,022 2279 2.03 Loss of Flow - (III) 2 Pump Coastdown 103 194,840 2246 1.87 Loss of Flow - (IV)

Locked Pump Rotor 105 194,840 2277 1.09 Loss of Load (II) 105 194,840 2511 2.16 Large Steam Line Break (IV) 52 38,600 1.35 Small Steam Line Break (IV) **

5a

Pressure decreases from initial value.

[

The core does not go critical.

m

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2.0 CALCULATION METHODS AND INPUT PARAMETERS I

The tra,nsie,t analysis for the Prairie Island plant was performed using the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTSPWR2).(I) The PTSPWR2 code is an Exxon Nuclear 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 the point kinetics equation is used to calculate tha core neutronic behavior. The program calculates fluid conditions such as flow, pressure, mass inventory and steam quality, heat flux in the core, reactor pcwer, and reactivity during the transient.

Various control and safety system omponents 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 evalua-tion is based on the hot rod heat flux in the high enthalpy rise subchannel and uses the W-3 correlation (2) to calculate the DNB heat flux. Model features of the PTSPWR2 code are described in detail in Reference 1.

A diagram of the system model used by PTSPWR2 is shown in Figure 2.1.

As illustrated, the PTSPWR2 code models the reactor, two independent primary coolant loops including all major components (pressurizer, pumps),

two steam generators, and the feedwater lines and steam lines, including all major valves (turbine stop valves, isolation valves, pressure relief valves; etc.).

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I XN-NF-78-35 To ensure conservative predictions of system responses with resulting minimum values for the DNB flux ratios, as well as maximum values for the system peak pressure, conservative assumptions are applied to the input data. These assumptions can be grouped into three general cate'gories:

1.

Generic assumptions, applicable to all transients, based on steady-state offsets.

2.

Assumptions which conservatively encompass ENC neutronic parameters.

3.

Transient specific assumptions yielding the most adverse system responses.

The generic assumptions (Category 1) are applied to all full power transients to account for steady-state and instrumentation errors. The initial core conditions are obtained by adding the maximum steady-state errors to the rated values as follows:

Reactor Power

= 1650 MWt + 2% (33 MWt) for calorimetric error.

I Reactor Inlet Temperature

= 530.5 + 4 F for deadband and measurement error.

Primary Coolant System Pressure = 2250 - 30 psia for steady-state fluctuation and measurement errors.

The combination of the above parameters acts to minimize the initial I

minimum DNB flux ratio. These values are consistent with those in the Plant Technical Specifications.

Table 2.1 shows a list of operating parameters used in the analysis.

The trip setpoints incorporated into the PTSPWR2 model for the Prairie Island Plant are based on the Technical Specification limits and the I

assumptions used are consistent with those used in the reference cycle E

I XN-NF-78-35 I

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analysis (Ref. 3).

These limiting trip setpoints with their associated time delays for each trip function are listed in Table 2.2.

The design parameter values for the Cycle 5 ENC fuel are summarized in Table 2.3.

Table 2.4 lists the neutronic parameter values which conservatively bound the ENC fuel for both the beginning and the end of N

Cycle 5.

A symmetric axial power profile with a peaking factor F, = 1.45 was used. The scram reactivity curve used in the analysis is shown in Figure 2.2.

The assumptions in category 2 refer to the reactivity feedback effects from moderator temperature changes and Doppler broadening.

For all BOC transients, a positive moderator temperature feedback has been used. To the Doppler feedback coefficient, an attenuation factor of 0.8 or a magnification factor of 1.2 has been applied, depending on which factor results in the worst case.

The assumptions in category 3 apply to plant control and protection systems.

As an example, pressurizer spray and pressurizer relief valve action are ignored in the locked pump rotor transient.

Since these assumptions are considered separately for each transient, they are de-tailed in Section 3 where each transient is described. The conservatisms applied to each transient analyzed are usually identical to those used I

in the reference cycle analysis.I ) The assumptions are quite standard, as given by any PWR FSAR cr other ENC safety. analysis reports.

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XN-NF-78-35 I

T URB 1:4f THRCTTt[ V ALV[

SliAM I

H[ ADE R N N U'.7 RELIEF ED 5AF(TY AI"l05Pe[R IC ATHOSPel R lC Rltl[F ED 5METY t AL VE5 DLMP UUP'E WAlIIS I

,, - =-n.ns.

150t Afl0, 1%l 811C/*

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ic z srt An cm suu em o

n Sn Amcas

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AND ORTIR5 W EO LRYtR$

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3 REACTOR COOLANT l

REACTOR CGGLANT.

PlfMP #2 PUMP #1 o m w:.-

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..,,. " " m,,,a a,1 c,

,me a,,s tt,,,,no s

m<1

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/EllTI% E 5thALP' 4:*D ik>Rt#3 CC'!CEhl F Ai!*' T I'l Di L# '

w Fuw R;stsiac u I

FIGURE 2.1 PTSPWR2 SYSTEM MODEL I

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80 I

3 3

70

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60 I

t3 5

50 I

t' 8

.5 3

40

.R t

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30 20 I

10 I

0 m

0 0.5 1.0 1.5 2.0 2.5 Time Since Start of Rod Movement (seconds)

I FIGURE 2.2 SCRAM REACTIVITY CURVE USED IN TRANSIENT ANALYSIS OF THE PRAIRIE ISLAND NUCLEAR PLANT I

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TABLE 2.1 PARAMETER VALUES USED IN PTSPWR2 ANALYSIS OF PRAIRIE ISLAND UNIT 1 AND 2 Analysis I.,put Value Core Total Core Heat Output, MW 1,683.0 Heat Generated in Fuel, %

97.4 System Pressure, psia 2,220.

Hot Channel Factors Total Peaking Factor, F 2.32 Enthalpy Rise Factor, F 1.55

,H 6

Total Coolant Flow, Ib/hr 68.20 x 10 6

Effective Core Flow, lb/hr 64.54 x 10 Reactor Inlet Temperature, F 534.5 Heat Transfer 2

Calculated Average Heat Flux

Btu /hr-ft

90,973 Steam Generators 6

Calculated Total Steam Flow ** lb/hr 7.233 x 10 Steam Temperature, F 510.9 Feedwater Temperature, F 427.3

Calculated from total thennal power and total cladding surface.

- Calculated from thermal power, feedwater and steam conditions.

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M TABLE 2.2 PRAIRIE ISLAND UNIT 1 AND 2 TRIP SETPOINTS Setpoint Used in Analysis Delay Time High Neutron Flux 108%

118%

0.5 sec Low Reactor Coolant Flow 90%

87%

0.6 sec High Pressurizer Pressure 2400 psia 2400 psia 1.0 sec Low Pressurizer Pressure 1830 psia 1700 psia 1.0 sec High Pressurizer Water Level 90% of Span 100% of Span 1.5 sec Low-Low Steam Generator i

Water Level 5% of Span 0% of Span 1.0 sec

?

T T

567.3 F AVE = 567.3*F 6.0 sec Overtemperature AT*

AVE

=

g g

2250 psia P

= 2250 psia P

=

g g

M gh Pressure Safety a)1830 psia coincident 1800 psia coincident 25 sec Injection with 5% level in with 0% level in pressurizer pressurizer b)3.32 ft steam 3.32 ft steam 25 sec generator level generator level

The overtemperature AT trip is a function of pressurizer pressure, coolant average temperature, h

setpoints are contained within the functional relationship.

AVE

and P and axial offset. The T n

g M

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TABLE 2.3 PRAIRIE ISLAND UNIT 1 AND 2 FUEL DESIGN PARAMETERS FOR EXXON dVCLEAR FUEL, CYCLE 5 Fuel Pellet Diameter 0.3565 Inch Inner Cladding Diameter 0.3640 Inch Outer Cladding Diameter 0.4240 Inch Active Length 144.0 Inch Number of Fuel Rods in Core 21,659 I

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I I XN..F-78-35 TABLE 2.4 PRAIRIE ISLAND UNIT 1 AND 2 ENC KINETIC PARAMETERS I

I Symbol Parameter Value Beginning-End-of-I of-Cycle Cycle "M

Moderator Coefficient

+2.0*

-35.0 (pcm/F)

"O Doppler Coefficient

-1.25

-1.60 (pcm/F)

"P Pressure Coefficient

-0.02

+0.40 (pcm/ psi)

"V Moderator Density Coefficient

-l,800.

+31,500.

3 pcm/(g/cm )

"B Boron Worth Coefficient

-7.70

-8.60 (pcm/ ppm)

Beff Delayed Neutron Fraction (pcm) 610 510

" CRC Total Rod Worth (pcm)

-2,120.**

-2,830.**

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This value applies to hot standby conditions. The coefficient is calculated to approach zero g)about 70 percent power, and

-3.69 + 2 pcm at full power

- These are conservative values, for analysis purposes only. The I

actual plant values are significantly higher.

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I XN-NF-78-35 3.0 TRANSIENT ANALYSIS 3.1 FAST CONTROL R00 WITHDRAWAL The withdrawal of control rods adds reactivity to the reactor 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 a DNB flux ratio of less than 1.3.

While the inadvertent withdrawal of control rods is unlikely, the reactor protection system is designed to terminate such a transient while maintaining an adequate margin to DNB. Two potential causes for such an incident are:

1) operator error, and f

2) a malfunction in the reactor regulating system or rod drive control system resulting in continuous withdrawal of a control rod group.

In this incident, the reactor is tripped by the nuclear over-power function. The rod withdrawal rate was chosen to give the most severe thermal response based on established core limit curves.(3)

The analysis is presented here to provide a check on those limits. '

The fast rod withdrawal was analyzed from an initial power level of 1683.0 MWT.

The reactivity insertion rate used is consistent with the rates analyzed in the reference cycle analysis.( ) Beginning-I of-cycle kinetic coefficients were used with an appropriate multiplier applied to the Doppler coefficient (see Table 2.5).

Figures 3.1 to 3.6 show plant responses for a fast rod with-

-3 drawal, iK = 1.0 x 10 1/sec,from full power. A -nuclear overpower trip (118% setpoint) occurs at 1.36 seconds. The DNB flux ratio drops I

from an initial value of 2.32 to 1.97.

Pressure increases to a maximum I

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of 2229 psia, with core average te...perature increasing by less than 2 F.

3.2 SLOW CONTROL R0D WITH0RAWAL The slow control rod withdrawal results in a smooth heatup of the primary system, limited by the overtemp'Nture AT or the over-power AT function long before any significant level of overpower is reached.

Based on the reference cycle analysis, a withdrawal value of

-6 at = 25.0 x 10 1/sec was chosen.

The plant responses for the slow rod withdrawal are presented in Figures 3.7 to 3.12.

The overtemperature AT trip setpoint is reached at 35 sec, and the shutdown rod insertion starts after a 6.0 sec delay.

The minimum DNB flux ratio is 2.03 at about 35 sec.

3.3 LOSS OF REACTOR COOLANT FLOW Flow coastdown accidents resulting from a loss of electric power to the primary coolant pumps result in.a rapid increase in coolant temperature, which cm1bined with the reduced flow, reduces the heatflux margin to DNB. Only the most severe case is analyzed:

Loss of both pumps from the reactor system operating at 1683.0 MWT resulting from simultaneous loss of power to the pumps.

Beginning-of-cycle values for kinetic coefficients are assumed.

For conservatism a multiplier of 0.8 was applied to the Doppler coefficient. The loss of power to all pum.s will result in a reactor trip due to either undervoltage or underirequency at the bus.

For conservatism, however, the trip was taken to be on a low flow signal.

This allows a further flow reduction at full power, and a more conservative calculation of heatflux margin to DNB.

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I XN-NF-78-35 Figures 3.13 - 3.18 present plant responses after the loss of both pumps. A reactor trip occurs at 2.7 sec. A minimum DNB flux ratio of 1.87 is reached at 3.7 sec after beginning of coastdown. At about 5 sec, a pressure peak of 2246 psia is reached.

3.4 LOCKED PUMP ROTOR In the unlikely event of a seizure of a primary coolant pump, flow through the core is drastically reduced. The reactor is tripped by the resulting ius TL., rianal. The coolant enthalpy rises, decreasing the heat flux margin to DNB. The locked rotor transient was analyzed I

assuming two loop operation with instantaneous seizure of one pump from 102% of rated power.

The effect of the pressurizer spray and pressurizer relief valves on reducing system pressure was ignored in the analysis. Also, steam dump to the condenser was not allowed, and the feedwater pumps were assumed to trip with the reactor.

Kinetic parameter values for the beginning of Cycle 5 have been used since they cause the most adverse plant response. A multiplier of 1.2 has beeg applied to the doppler coefficient.

Two cases have been analyzed foi this transient, one with

-0

-0 5k=+20.0x10 /Fandonewith5k=-16.9x10 /F for the moderator temperature feedback coefficient (see Taole P.4).

The first case is conservative since it combines the most positive hot standby feedback coefficient with assumed full power operation.

The conservative case (positive rnoderator coefficient) results in a DNB flux ratio of 1.09, the more realistic case yit.lds a ONB flux ratio of 1.19.

Results for the conservative case are reported. The transient responses are shown in Figures 3.19 to 3.24.

The reactor is tripped at 0.7 sec by a low flow signal. The core average temperature I

I XN-NF-78-35 I

increases by 13 F with a system pressure reaching 2,277 psia, wel' below the power operated relief valve setting of 2,350 psia. The number of fuel rods expected to experience departure of ".ucleate boiling has been cal-culated to be less than 1 percent for the conservative case.

3.5 LOSS OF EXTERNAL ELECTRIC LOAD The Frairie Island plant is designed to accept a 50 percent step decrease of electric laad without a reactor trip.

For a complete loss of electric load at full power, the reactor is tripped by a signal derived from the turbine stopvalves.

In the analysis of this trans sent, it is conservatively assumed that only the turbine is tripped on the Loss of Electric Load signal,but not the reactor.

I In addition, the pressurizer spray system and the power-operated relief valves are assumed to be inoperative.

On the secondary side, the turbine bypass into the condenser as well as the actuated steam relief valves are assumed to be inoperative.

Neutrr.ri-data for tne beginning of the cycle are used (positive moderater temperature feed-back), and unavailability of the automatic reactor control is assumed.

In addition, a factor of C.8 is applied to the doppler coefficient.

The criteria for this transient are 1) the ability of the passive pressurizer safety valves to limit the reactor coolant system pressure to a value below 110 percent of the design pressure (2750 psia) in accordance with Section III of the ASME Boiler and Pressure Vessel Code and 2) a sufficient themal margin in the hot fuel assembly to assure that no departure of ru'leate boiling occurs throughout the transient.

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Figures 3.25 through 3.30 show the plant responses for a complete loss of electric load at 102 percent of full power without a direct reactor trip. After closure of the turbine stop valves, the pressure in both steam generators increases at an average rate of 20 psi /sec, reaching 1090 psia at 13 sec, when the first set of steamline safety valves opens (see Figure 3.28). At 16 sec, the second safety valve setpoint of 1105 psia is reached. After that point, the steam pressure continually decreases.

In the primary system, the pressure increases at the same average rate as in the secondary system, only delayed by about 5 sec (see Figure 3.28).

The reactor is tripped on the over-pressure signal at 13 sec, the peak pressurizer pressure is 2537 psia.

The pressurizer safety valve is open from about 15.5 see to 17.5 sec.

The average primary coolant temperature increases by about 23 F.

The lowest value for the minimum DNB heatflux ratio is 2.16, at about 13 sec.

3.6 LARGE STEAMLINE BREAK The break of a steam pipe (or safety valve failure) results in a sharp reduction in steam inventory in the steam generator.

The resulting pressure decrease causes an energy demand from the primary coolant which reduces coolant temperature and pressure. With a negative moderator tempera-ture feedback coefficient (at the end of the cycle), this causes a reactivity insertion into the core 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 magnituie of the primary loop cooldown.

For conservatism, the most reactive I

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I XN-NF-78-35 I

control rod is assumed t; be stuck out of the core when evaluating the shut-down capability of the control rods.

The reactivity as a function of core average temperature and the variation of reactivity as a function of core I

power used in this analysis are shown in Figures 3.31 and 3.32.

The moderator and Doppler feedback coefficients are valid for Cycle 5 fuel.

Minimum capability of the boron injection system was assumed, which implies that only one of the two 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 trip point (1800 psia). The time required to sweep the lines of low concentration borated water prior to the introduction of 20,GJ0 ppm borated water from the Boric Acid Tanks has been 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.

A large break at the exit of the steam generator with offsite power available was analyzed. A 10 sec delay was used to cover the startup time for the high prersure injection pump. An initial break flow of 600 percent of rated flow was chosen.

Figures 3.33 to 3.38 show the plant responses.

The core returns to criticality at about 10 sec.

The power reaches a peak value of 52 percent of nominal full power at 56 2

sec with a corresponding peak in core average heat flux of 98,000 Btu /(hr x ft ),

At this time, the borated water from the high pressure safety injection system reaches the core, initiating a power decrease. A conservatively large local hot rod peaking) factor of Ff = 10.0 was used.

The lowest value for the heatflux margin to departure of nucleate boiling was 1.35, at about 54 sec (W-3 correlation). A shutdown reactivity of 1,800 pcm has been used.

I

I XN-NF-78-35 I 3.7 SMALL STEAMLINE BREAK The small steamline break transient is intended to envelope a valve failure.

For instance, an actuated steamline relief valve or a turbine bypass valve could fail open and release steam. A small break at hot standby conditions, two-loop operation, with an initial steamflow of 25 percent of nominal full flow with offsite power available has been coalyzed. The most significant parameter responses are presented in Figures 3.39 to 3.44.

The boron injection is triggered by the same signal as in the large break case.

The borated water reaches the core at about 150 sec, and the core does not become critical. A shutdown margin of 1,800 pcm has been used.

I I

I

M

,a' x?Ebh M

0 2

wN N

1 a

l

~_

a 0

w 0

3 a

.N X

3 r

8 4

d h

7L t

a i

W 0

do X,~N N

8 8

R 7

/

M 0

l 1

o

/,

r a

1 tn 0

o C

X\\*

N' r

7 M

YO t

L K

s R

a a

I M

F W

Q M

O 0

E o

L S

F f

G

+

TN W W 4

O O s

i C R sr N

8 U L F L

w 7

0 F o

0 2

E l

C E

C F

I 2

O Y T R A L E

N A

r A

O S m

A 3

a M

C J

N e

1

, x R,

E c

1 7

5 t

F $

wP E

+

L s

4 M

E1A I

y C

a F c T E

n n T 7 T S

O 0 O T

S a E, T 7 T

/

p d

1 n

3 0

a

-E

. a 3 4 M

t 5

4 x

0 u

1 l

f T

2 t

A N

a L

e n

W L

0 H

n 1

\\

3 R

I r

F T

e I

\\_

w 3

W o

1r M

P D

I

\\

0 P

I 2

T

^

1 E

T w

3 F

+

E o

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U I.

0 G

L G

L I

[

I F

E IR IA P

M P

0 0

0 4

2 0

o 0

0 4

2 0

8E mo v 1

1 1

8 M

M

'r",

xY*

o m

0 0

V a

1 a

z l

M 0

0 3

w 3

W a

6 r

4 d

7 h

1 t

M a

i W

0 d

W V

8 4

o 7

R

/

0 1

l

/

o 3

2 1

r tn 0

o 7

Y C

01 t

W s

I a

t F

z_

Q t

E r

0 S

o

+

RI W N

6 f

i E l

T e

8 R

7 RT s

E N E n

WA RL T

C o

C L

E O T E

p O

T O A 1

C R N

S s

0 3

T E

e 1

M' 5

R L P N

+

N D E E

I C T M

e C

E

. A 3

D I

r E

R E T

u O V L

/

C A C t

1 a

1 r

0 3

e M

N C

M e

F

. 2 3 4

p 1

0 m

1 T

TA er t

t i

o o

W O

3 O

^

D 2

E e

/

2 3

D L

t E

o.

P R

T 2

U E

G T

I m

F F

+

O R

0 x

I S

1 I

E IR 3

IR R

m P

1 0

0 0

6 4

2 0

8 6

2 6

6 6

6 5

5 5

nd3E!

m m

PRRIRIE I190 + FMT RCD WIT 100WFL AT 1.oE-31/9r + 13 oCT 78 +

CGo 1.

hot tcc encomTtste Lcor t 2.

COLD LEc TOPERATUltf LodP 1 l

G4e C2a i

l' 1

1 9

1

%sso N

t g

5Go 7a R

1 2

g g

5 528 0.0 1.o z.o 3.o 4.0 s.o G.o 7.o s.o s.o to.o a

TIME, SEC SEQ. MIRIC10Y 11/16/74 17.4s.30.

FIGURE 3.3 - Primary Loop Temperature Response for Fast Control Rod Withdrawal M

M M

M PPERIE IEJYO + FAST RO) WITr{RAWrt AT 1.0E-31/SEE + 13 OCT 78 +

250 1

STEM 00iE PRESSURE 00NCE. LOOP 1 2.

STEM 00iE PRESSUPE OnNCE. LOOP 2 3

PRESSURI'Gt PRESSURE OWWCE 200 2

Initi al Steam Dome Pressure:

758 psia Initi al Pressurizer Pressure:

2,220 psia 2

t 150 t

2 1D0 y

1 a

s

/

E" l

3 1 2 _

1 2,s P

R 0

w N

% 3 x

-50 z

M b

a

~ b.0 10 2.0 3.0 4.0 5.0 6.0 7.0 80 9.0 10.0 TIME. SEC SED. atIRK10Y 13/10/78 17.45.30 FIGURE 3.4 - Pressure Changes in Pressurizer and Steam Ge'nerators for Fast Control Rod Withdrawal

PRRIRIE ISUM] + FAST PID WITHRAWR AT 1.0E-31/SEC + 13 OCT 78 +

10 1.

DONCE L' STEAM CEN. eRTER LEVEL. LMP L 2.

UmHCE Li STETH CEN.

Rl'ER LEVEL. LUP &

u 3

DONCE L' PRESSLRIZER WpitR LEku.

1 2 0

g w

3

-10 4N N

=ji-20 g

w H

ui u -30 w

"i

-40 3

~50 x?

2?

w 0.0 10 2.0 30 4.0 50 6.0 7.0 8.0 90 10.0

?

w TIME, SEC m

SEQ. MIPK10Y 13/10/74 17.48.30.

FIGURE 3.5 - Level Changes in Pressurizer and Steam Generators for Fast Control Rod Withdrawal M

M M

M W

W W

m'

M

.NT x?z,Oc0U 1

n 1

0 M

01 0

.o J

9 6

4 7

1 M

0 la 8

w a

L r

/

d 3

h 1

t i

0 W

i 7

w Y

o O

d n

L o

K R

n.

RI a

m it l

o i

Q r

c E

+

2 0

S t

6 n

w o

8 a

C 7

a T

u t

C n

s C

O w

E a

S F

3 e

d 1

o 5

r

+

m E,

o M

C s

M f

i I

E n

i T

o

/

m i

1 t

3 a

J.

R E

O 4

t x

L u

l T

F A

M B

1 i

4 h

0 r

J D

P 3

I m

g u

FT m

IV in D

i C

M R

9 0.-

T 2

E T

6 F

+

3 O

E R.

R 0

U S

L G

I I

E F

IR IR RP 1

0 0

4 2

0 8

c.

4 2

0 2

2 2

L t

L 1

1 oGE x3w pa 5E m

PRRIRIE IStJ10 + S.DW RCD WITHRWFL AT 25.0E-61/TE + 13 MN 78 +

140 1

P0wcR t,c, et 1

HEATFLUM 3

TOT A PR MARY COOLANT FL4W 4

TOTA FE1 DWRTER FLOW 5

TOTA STI AHLINE FLOW 12.0 L

5_

3 3

8 100 c80 3

a' s

b qGO u

5 (1.

40 5

5 5

( \\

s 2D q

z 5

i 1

7

' a 5

t,

y b

4 4

o 0

10 20 30 40 50 GO 70 80 90 100 TIME. SEC SEQ. MIRJC102 13/ L1/76 14.43.18.

FIGURE 3.7 - Power, Heatflux and System Flows for Slow Control Rod Withdrawal M

M M

h N$*

M M

00 M

1 8

0 NRx 9

9 4

la 4

1 w

M ar d

h 0

t AN 8

8 i

7

/

W M

L L

d

/

3 o

1 R

l 0

o M

N r

7 2

r 0

t 1

K n

R o

a I

M C

QE o

0 S

l 6

Kn S

t F

+

wEH T T r

8 n

7 PT 3

o t N E f

/

> A P C

o

O T L U NE E

e E

t T O A S

s C F 3

0 n

i E

1 6'

5 o

uM P

E

+

N O E p

M I L T M

s C

I E. D T

e T

f R E A R

o V L cC C U

a s

er 6

0 M

u u

4 E

. 2 J O

1 g

tar 52 e

p T

m A

2 M

e L

T 0

M 3

e JP ro M

C IW D

8 C

0 R

2 3

W O

M L

E S

R 3

U

+

2 G

U I

F Y

0 J

1 M

S I

EH 3

G a_

M fH F

0 0

0 6

4 2

0 8

6 6

6 5

5 rE h

M t

M

F45 ERIE II2J10 + SLOW PLO ufITtOY7drt. AT 25.0E-6 USEE + 13 MN 78 +

C60 t.

,c7 u;; - w o,nr e u op t a.

COLD LEC TEWEPAT M i.00P 1 640 62D 1

600 C<

N g

r 5rr E500 m

t s'N SEO a

/

a x

m 54o a

N 2

d 9

u 0

10 2D 30 40 50 60 70 80 90 100 TIME, SEC so.nrw.1cz tuttne 1e.es.te.

FIGURE 3.9 - Primary Loop Temperature Response for Slow Control Rod Withdrawal m

m m

M M

M Hii25 M

M M

M

M M

'M M

M M

M-M M

M M

PRRIRIE TSRO + SLOW POD WIT 10PRWAL RT 25.0.I--61/EC + 13 POV 78 +

400 L.

STERN DOIE PRESSURE O.NCE. uGJP L E.

STERN COLE PRESSURE OnNGF. LJ)P 2 3

PRESSURI EE PRES 3UPE ClWNCE

/

300 Initia l Steam D)me Press Jre:

758 p s.,i Initia l Pressur izer Pres sure:

2, 22 psia 200 N

IDO i

3 L E C

L L L 2 3s f

H 0 N

N

-100 N3

-200 N

x

?

25 b

-30S 10 20 30 4a 50 60 70 80 90 1D0 h

TIME, SEC SEQ. MIRK 102 13/11/78 18.43. 4.

FIGURE 3.10 - Pressure Changes in Pressurizer and Steam Generators for Slow Control Rod Withdrawal

FHRTxIE N + SLOW FG WITHHWR_ AT 25.0E-6 VSEE + 13 FD/ 78 +

0 1

00NCE Di STEAM CEN. iATER LEVEL. LXIP L 1

OSNCE DI STEAM CEN. aATER LEVEL. LAP &

3 OMNCE Di PRESSLRIZER WATER LEVEL 2

L 2 1 2 3/

L 2 1/

a-#

\\

0

_g s

,N

\\

N r

L da is

?

E d-60 lb 3

?>

a

\\

-80 s

-100 x7 5

Y N

10 20 30 40 50 60 70 80 90 100 h

TIME, SEC SE3. MIPX102 13/11/76 18.43.it.

FIGURE 3.11 - Level Changes in Pressurizer and Steam Generators for Slow Control Rod Withdrawal mm m

m m

m

.M M

M M

M-M M

M M

M 4

x2i27e$

M M

0 M

01 M

8 0

1 9

9 4

l e

a M

l war d

0 h

8 8

t 7

i M

/1 W

1

/

d 31 o

R M

i 7

o 0

l w

2 o

0 r

r 1

l t

I

(

n.

R n

I o

u M

C m

M a

O S

o Q

c E

w x

G l

+

8 S

w 7

o r

M r

o N

Ta C

f K

w ES o

m 31 0

i 5

t

+

n E

a M

u C

n M

R I

I E

n T

x I

/

m u

1 l

4 F

M 0

E 4

B O

1 N

52 D

T m

A u

M m

G L

i 0

n D

3 i

M D

D M

I W

2 D

1 t

E 0

3 F

2 N

W E

D.

M R

N S

UG

+

I O

F R.

t M

0 S

1 I

tuU R

M RP J

0 8

6 4

2 2

0 4

L 1

1 1

c' [E 5u. L E

2 2

ga 6r M

M

RNURIE ISRO + PRDTRY PlW C0RSTCIM4 + 12 ocT 78 L40 A.

eawee te.a a.

riEATFLOW 3

TOTAL PR MARY COOLANT Flow 4.

TOTAL FElDwATG4 Flow TOTFIL ST'.AMLItaE Flow 120 4 3 1

4 5 1

1 2 3

s

__g 4 2

a w

N X

N N

8 9

w N

X E

a X

m W

y 4

N

~

w 5

n.

40 4 N.5 i

L g

t 5

e-T 0

M 0.0 1.0 20 30 4.0 50 G.0 7.0 8.0 90 AD.0 d2 on TIME, SEC sea. %IRn oa ta/torre ts.os.ac.

FIGURE 3.13 - Power, Heatflux and System Flows for Coolant Pump Trip g

g g.

m g

g-M M

M-m M

M M

W W

M M

M M'

M M

N IS_RO + PRDfRY PLM2 CDRSTIINN + 12 OCT 78 GGO 1

core INu.r TseenAT#e 2.

AVE. CO3ii, COOLANT TEMF 3

C. LAD TEH'tPAT SE 3

G40 p

3 3

/

3 G20 N

w 600 w

N2 b

w "H

$580 E

a a

2-t N

{

N, a

N_a 560 540 25 1

L L

7 t

1 1

1 L

t t

h

$2D W

0.0 1.0 2.0 3.0 4.0 5.J 6.0 7.0 80 9.0 10.0 m

TIME, SEC SEO. MIRKt02 12/10/78 t6.04.40.

FIGURE 3.14 - Core Temperature Response for Coolant Pump Trip

PRAIRIE ISRO + PRIFRf PUH) 00ASLMJ + 12 OCT 78 660 2

.m LEC t @ ERAT M LCOP L 2

COLD LEC TEWtRATURE LOOP 1 640 620 L

g 1

t L

600

~

5 m

E580 u.

560 540 x

2 a

2 2

g 3

g 2

z

?

M 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 TIME. SEC sEm enez tutane ts. o 4. u.

FIGURE 3.15 - Primary Loop Temperature Response for Coolant Pump Trip M

M M

M' M

M W

M M

M

M M

M PRERIE ISJto + PRDTRY PUW CDASTDWN + 12 oCT 78 240 1

STEM DC4E PRESWRE dGCE. LOOP 1 2

STkN DC1E PRESSURE d@ CE. LOOP 2 3

PRESSURIZER PRESSURE 114ANCE 200 Initi al Steam Dome Pressure:

753 psia Initi al Pressurizer Pre ssure:

2,220 psia ISO L

i 120 r

i

/

e P

/

2 40 2

3 1 2,

L 2 a 1

[-,

f O

~ -

I e

d 0.0 1.o z.o 3.o 4.0 s.o s.o 7.o s.o s.o io.o TIME, SEC sea. nrRyto2 L2none tc.os.zo.

FIGURE 3.16 - Pressure Changes in Pressurizer and Steam Generators for Coolant Pump Trip

m EhTMeU

.$i M

0 01

\\

M 0

0 M

2 N

9 4

d s

i M

0 r

o 8

4 f

7

/

O s

N L

M r

/

1 o

1 t

L 2 a

0 r

W&

e 7

'N 2

n L L M

D e

L

,l L,

L E X G E E V P

V V E I

E.

L M

m EL R

. a R R E Q

e L L T

E t

0 g

M A

T IA W S

s

)

F e

G N.

w.

dn E E C C a

t fNs C

r M

M AE E f E

e I

T P

S 8

S S P 0

z 7

If 5

i F

r T

I I I'-

C E

u M

O E E W

s C C I

M N N s

2 OO T

1 e

DD r

P

+

0 p

N ni W

0 L

4 0

. s 3 i r M

T T

s G

f ep 0

gm C

nu e

aP t

0 h

M l

Ct P

3 n

M M

f l a R

el K

vo eo IR LC P

0

+

2 O

7 1

R M

S 3

I E

E a

R I

0 U

R I

G Q

1 M

I H

~

F~

F M

M a

0 i

0 0

D 0

0 0

0 4

4 0

1 2

3 4

1 O

~

1M p5b i?a.

j c

M M

M

.wNi N,2, M

M 0

01 M

M 0

01 9

4 0

s i

M 0

8 8

7

/

0 M

1

/21 0

p M

i 7

2 i

n 0

r o

1 u

K T

R n

I p

i M

a m

m u

Q 0

E P

M a

C t

c S

n w

a a

l o

o n

M C

o a

R E

C S

8 0

r 7

o 5

T E.

o f

C m

M O

e M

n I

o 2

T n

i 1

I t

a R

N 0

W 4

x O

3 M

D 1

u T

l sn F

o N

B c

N e

0 D

M u

3 m

F u

YR m

m in I

m 0

i RP M

+

2 O

8 R

1 M

G I

3 t.

E u

R 0

w U

1 G

M I

F F

0 M

1 0

4 2

0 8

6 4

2 0

2 2

2 t

L L

1 1

OpaE sd ga 6I M

M

FRRIRIE IRRO + PRD5fY PLH) SE32LRE + 17 OCT 78 +

N 1.

POWOI LEVEL 2

HEATFLIId 3

TOT A PRDWUtY C00LRidT FLOW 4

TOT A FEEDWATER FLOW 5

TOTA ST_% LINEi FLOW 140 120 1

l h

4 Y

M DM t

E

~

x M

O 2

g80 3

7 s

2 E

Ns

~

2 x

~3

?

a 40

\\

U L

t t

0.0 0.5 1.0 1.5 2.0 25 30 35 4.0 45 5.0 TIME, SEC SEQ. MIRK 1FE 17/10/78 12.34.2G.

FIGURE 3.19 - Power, Heatflux and System Flows for Coolant Pump Seizure g

g g

g

.M W

M M

Mltl 5.. n M

M 0

M 5

M 6

5 2

N 4

+

J 21 M

L 0

N 4

8 7

/

0 e

M 1

r

/

u 71 z

2 i

L e

5 S

M 3

EF p

1 m

KR u

I P

M t

Q M

L E

n 0

S a

3 lo o

7a C

u e

M c,

m 2

C r

e i

E o

n 1

S f

a 5

+

e 8

e p

2 E

s mma 7

ct M

n M

T I

o C

e

, o T

p O

p e n o

v L s

7 c

a c 2

e 1

L R

+

0 2

e M

E 1

2 r

f u

n t

3 a

+

re M

L p

P}

5 m

l 1

e P

7 T

YR e

f r

H M

o I

2 R

C P

L 0

+

1 v

Y G

O 0

2 R

M L

3 SI f

E t

1 R

u 5

U m

0 i

I R

F P

M L

0 0

5 o

5 0

'0 5

7 s

2 0

5 2

5 s

rW's=E G

S 6

6 M

M

FPMRIE ISLR0 + PRIMEY PLH) SEIZLE + 17 OCT 78 +

660 1

,iOT tec itneexntuwt Lixw t 2

LOLD LEC TEt4*ERRIJfE l40P 1 640 620 L

1 t

~

600

_t

~

t t-g

?

5 rr E580 a

560 540 5

2 2

2 g

T M

i 520 0.0 0.5 1.0 1.5 2.0 2.5 3.0 35 4.0 4.5 50 TIME, SEC SEQ. MIRKLF2, L7/10/74 12.J4.25.

FIGURE 3.21 - Primary Loop Temperature Response for Coolant Pump Seizure e

e e

a m

m M

M M

M 7Mh

b' M

M 0

M 5

G2 4

4 J

2 1

M ro f

0 4

8 s

7 r

/

M 7

o o

L t

/

7 a

L r

2 ne 5

e M

/

3 2

G t '

F1 p

K m

a R

a w_

I y

M e

t

.. E 2

Q S

s E M

c c E

u N 0

S d

mE 3

n c C a

eE n R u U r

s S e

M s S eE C

z R H E

i P P

+

t S

r 5

u E E 8

4 M

/

2 s

7 c

C R

E s

oD U a

M e

M T

S C

MMS i

I r

O eE E s

T t

T R P

7 s S P a

p e

i 2

1 nr s 0 iu

+

p 2 0

2 z

M 1

7 2

si E

2 2 ee 5

n gSn i

ap

+

hm e

Cu P

e r M

P F

r u

/

5 e

U u

s r

P 1

rt s s un Y

s e R

e r sa sl M

r P 3

eo M

I P

R r

2 ro PC F

e e

+

m z 0

a i 1

y O

D r

2 A

u M

2 m s LS a s 3

3 I

e e 2

E t

r E

I S P R

R 5

U IA l

l M

/~

0 G

F a a I

F i i F

t t v

i i n n 2

I I

L 0

O 2

O O

0 0

0 o

GnA G

5 4

3 M

M

mfERIE IG.RO + FRImRY PttP EIZLFE + 17 oCT 78 +

30 1.

cin<ce I4 stEnM CEN.

i'RiER LEVEL. L00P L 2

LHRNGE Id STEAM GEN. inTLR LEVEL. i.00P 2 3

O ONCE I4 PRES $URIZE34 W.1ER LEVEL 20 3

s 3

2 10 M

/

7

,a 2

2 a

I t 2 a 2

2 "O

m i

?

1 A

~10 W

b N

-20 xz

-30 7'

M

~400.0 0.5 1.0 1.5 2.0 2.5 2.0 35 4.0 4.5 5.0 TIME, SEC SEQ. MIRK 1F2 L7/LO/76 12.J4. s.

FIGURE 3.23 - Level Changes in Pressurizer and Steam Generators for Coolant Pump Seizure g

m m

m M

M M

M iw' 5

4 bw

~

M 0

M 5

C2 4

4 3

2 1

M L

0

/

4 8

7 M

/ e 0

r 1/ u 7

z L

ie t

5 S

M 3

2, p

no F

r 1

m i

K u

n.

R P

I a

M m

t M

i S

o Q

n c

L 0

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l 3

o n

o o

C M

r t

r n

C o

e

+

E f

t S

a

/

5 8

o 7

2 i

n E

M T

u C

n M

t I

I a

O n

I T

R 7

n 1

x u

+

L 0

l M

E F

t 2

n B

N D

+

P m

M H

u U

5 m

P i

1 Y

n R

i m

M M

IR P

4

+

O 2

L O

s 3

A M

S E

I R

U E

G I

I RE

\\

5 F

M V

0 R

M l

0 L

0 8

G 2

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2 0

L 1

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2 2

l pE se gQ 6I M

m

PRRIRIE ISLR0 + TLRBItE TRIP AT 102 PCT POWER + 1G OCT 78 +

40 l

3 povoi te c.

t.

Mcpretur 3.

70T4. PR J@RY C00LapT Flota 4.

TOTAL. FElW ATER FLOW S

TOTN. STCArd.INE FLOW D

L L

1 2 L 2 L 2 _

2 L

x e

2 1 '

c 2

s s

t

\\

4

\\

5 r

5 g

m 2

p S

\\

m

^

\\

N 4

s N__

s n5 40 s

N

\\

=

4 N

l A

2

~

l

=

s s

0 2

4 6

8 10 12 14 16 18 20 TIME. SEC sea. nInsi2o scitoire te.ss.tr.

FIGURE 3.25 - Power, Heatflux and System Flows for Turbine Trip M

M M

M WE M

M M

M

M a

?=0?W M

M D

2 M

a M

N 7

8 1

1 5

5 N

t L

a M

s B1 M

N

/

8 7

/

0 1

/s i

N a'

t 4

1 M

D p

3 L

i K

r RI T

M M

e D

n E

i 2

S 1

b a

I' V

V r

M E

u i T T

A

+

R T E N E M

/

r A

eL W 8

C o

7 E 0 T E

f T 3 A S

T C P C

T t

01 e

O

$M

/

t R E

s 6

uO E M

n zC T M

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o E. D T

p n E A oV L s

R c A C e

E R

WO P

8 e

M t a3 r

T V

~.

u F

ta 2

r 0

e 1

M p

T m

A 6

e T

P IR e

T r

3 M

E o

N a

C IS F

T I.

i

!I i

l' 4

t 6

+

2 M

O 3

3 R

L L

E S

R 1

I U

E 2

G M

I u

f F

u B

3 F

M x

0 0

0 6

4 0

6 4

6 6

6 g

5 5

d$

it M

M

PPfERIE ISJ10 + TURBItE TRIP GT 102 FYT POWER + 1G TT 78 +

660 !

,e0T LEC LMTmTUW L -

t 2.

COLD LEC TEMPGATJRE WdP 1 l

t l

I l

l l

640

[

i i

I l

i I

i i

620 i

y

/

N t

600 U

i b.

x E580 2

1 t/

\\,

560 f

I i

/

540 ~

4

?

I.

i a,

.z0

~ 0 2

4 6

8 10 12 14 1G 19 20 TIME. SEC sco. mInxtso scitone te.ss.17.

FIGURE 3.27 - Primary Loop Teraperature Response for Turbine Trip W

M M

M

M i?

x2 aa#

M M

0 M

l 2

z M

L 7

8 1

1 N

5 5

ro f

f i

M sro 6

t 1

4 a

M V

7 r

/

0 e

1 n

/

e s.

/

G m

4 a

M 1

0 e

L z 3

t 1

P P K

S O O P

O O 3

I L L d

m n

E.

E 2

a G

M Q

C C N N N A 1

E 4 A H 2

S r

1 C O e

C 1

E z

i l

EE R i

R R U

+

U U S r

3 S S S u

M 8

S S E C

s 7

E E R P R P E

s P P T

S e

R C

E E E a

0 r

O l

01 I 1

f P

O i

G D 0 R s

E 1

U a

p M

n M

M M S I

A A !

i i

+

T E E E s 0 T T P R

S S P p 2 s

f 2

2, E

e W

8 g

O 1

5 2 n

P 8

ap M

}'.

3 T

. 2

l7 C

hi P

C r e

T 1

e r e

0 r u re 1

2 u s un M

T s s si A

s e sb 6

e r

er P

r P ru IR P

PT T

r e e E

M m z r

I o i 8

B D

r t

2 F

u l

s 4

r 3

T r

s a

+

e e E

M t

r R

O 3

S P R

l iG L

l l

S I

I a a F

i i t.

t t

2 M

l.

i i xu n n I

I e

F t

0 0

0 J.

0 0

t

~

6 5

4 2

i e$1 M

M

PPNRIE ISLAO + T114TK TPlP AT 102 PCT POWER + 16 OCT 78 +

75

" 7 umpcr Li stran erw. sten tevet. twt umpcc Li stenn crN. hTERLEVEL.i UP 2 s.!

t.wascc ti peesstmI2En !wnttu Levet i

I i

N I

/[

3 25 3

l I

f l

r 1 2 E0 h

H s

W E

I 5-u

^

E!

I l

J 2

i

-50

't 7

N At

/

i

-75 T

~

i m

j a

i i

m i

I

~ 0 a

4 6

8 10 12 14 16 16 20 TIME. SEC tro. minx 120 istions te.ss.u.

FIGURE 3.29 - Level Changes in Pressurizer and Steam Generators M

M M

M E

=L M

M M

i 02 M

7 8

1 1

5 5

0 1

M 61 4

?

/

M 0

1

/

C 1

4 1

D p

M i

v e

wo M

i n

K r

R T

e I

t M

ca n

m C

M i

x oc E

i 2

S b

1 r

-w u

T a

+

oz r

M 8

r 7

a C

o m

E f

T S

C e

0 o

O i

1 pN i

E t

G M

a 1

M I

R p

T I

R n

x E

u W

l O

F P

8 M

T 1

C B

P N

D 20 m

1 u

M T

m t

A i

6 n

v i

K I

M E

NI 0

M,'

3 I

4 T

3 M

. V

+

E M

O R

U R

G L

I SI F

t 2

u D

A9 F

M j,

O t

4 2

0 c.

4 2

8 2

2 2

L t

1 t

CQe x5a pa Mr M

M

I I

1.0 0.9 0.8 0.7 b

250.6 "E

M c

g 0.5 M%

g1 0.4 0.3 0.2 x

0.1 2

5 1

1 I

I I

I O

O O

10 20 30 40 50 60

?$

Thermal Power Percent of 1650 MW)

FIGURE 3.31 - Variation of Reactivity with Power at Constant Core Average Temperature M

M M

M

I

-M-I g.tW-78-35 I

i i

i I

i i

I I

4 I

I 3

I E-52 I

58 35 2

l 3.5 Ua 23 I

1 I

O 400 420 440 460 480 500 520 540 560 I

Average Core Temperature (in Fahrenheit)

I FIGURE 3.32 - Variation of Reactivity with Core Average Temperature at the End of the Cycle I

e tO?

xzey ?U e

m 00 5

1 a

3 a

ka L

S m

e 0

L s

3 r

4 B

2 e

?i 2

n i

m a

lm L

0 ae 8

4 s

7 t

/

S 0

1 m

3

/

e 4

1 g

a ra L

0 L

5 7

X r

M C

t o

Kk f

3 I

M w

L o

ir A

Q l

o E

M L

0 S

F F

N<

6 5

m TN W w e

A O o V

t L L l 0 F F 2

s 0C R.

E C

y M

f N

z

+

T T I E

S 8

2.

R R L L

S A

a H

7 1

M iD A 0

d l

'I 5

5 n

O X R,

E T E

a F S T

L C

WP M

AAA I

x O

R F M

E T W R T T T 2

T u

4 O E O O O l

2 P

4r T T T a

f

+

t t

0 a

K

. a 3 4 5

H 4

e 1

2 B

M P

B e

,r 2

E e

t I

w L

o t

M R

0 P

3 E

5 T

S 3

T R

3 2

M fL 3

+

\\

0 E

5 2

R 0

U r

T G

L I

M S

2 F

I t

uu O

R

_ L R

M P

2 t

i t

M 0

0 0

0 O

8 4

0 E

2 0

0 E

2 2

l 1

6 4

0 c Ex s EU5n.

M M

M m

M M

PRAIRIE Ia.R0 + LAE STEPH_DC BREDK + 24 OCT 78 +

600 1

cms ruu r it e s Ai m a.

AVE. cm: cc0GNT TEW, 3

(1# rtM' GAT URE 580 560 540 t

$20 5

5 E

500 m

3 4

~-

N 3

3 3

3 5

480 z

c a

2 2

2 L

L L

ww 0

10 20 30 40 50 60 70 80 90 100 TIME, SEC SEO 64 IRK 1CX 24/10/74

17. 2 4. L3 FIGURE 3.34 - Core Temperature Response for Large Steamline Break

M 5a 4e$

M M

001 M

z k

M t

t a

0 i

e 9

4 r

2 B

7t en M

2 i

lm t

0 a

8 4

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/

t M

0 S

1

/

4 e

1 2

gr r

a 0

L 7

M KC r

t K

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f I

4 1

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. s L

Q M

P E

n P O 0

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6

~

UL p

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T A

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e

+

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u 8

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5 C

a T

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r C

E L M

M I

e O

L D.

T p

laX m

42 H U 2

e

+

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K

. z o

p M

S l.

4 t

P o

B L

E y

D 2

r L

a M

N L

m I

0 i

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r r

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m M

5 L

3

+

t 02 3

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R

\\

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R S

U I

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0 tr 1

M R

P M

0 5

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4 5mNtu W

W

M xY27Nh M

001 M

2 2

M t

0 3

4

1. r 7

o 1

2 f

M 2

sr t

0 o

t 8

4 7

a

/ r M

0 1

e 2

/ n 4

2 e

2 G

t 0

ma 7

M X

t 2 C

e 1

t P P K

O O R

S O O 3

I L L M

d E

z n

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G Q

EC C N t

E a

N N A 0

S R n H 6

r Oot e

E z

E E R R R U i

U U S 2

r S S S M

S S E C

u E E R 2

+

R R P E

s P P t

S s

R 0

e 8

E f E r

5 7

l e

T a i aO E

Pk 4

T D D K i s M

a M

C LS s

p I

ne M M O

A A S n

L E E 2

T i r l

l k 0

B 4

2 S S P 2 9 2

s 0 2 ee

+

t 1,

0 gn K

4 ni

)

M T

1 E

. 2 3 al 1

R h m B

C a e

e E

e r et N

r u 2

rS l

u s u

M h

t s

s, 0

s e U

3 sg s

T e r er S

r D ra P

PL E

r G

M i

R e e f

m z a

L o i 6

+

D r t_

3 02 O

m s

3 R

a s

M L

e e E

S t

r R

I S D U

2 G

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l l

i I

R a a A

i i 0

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1 M

R t t P

i in n I

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M

{

d 0

0 0

0 2

4 6

8 4

2 o

gMr.t M

M

9 PRrHRIE ISLDO + LfRT STEN 4_INE BRETK

24 OCT 70 +

200 1

u nNGE DI STE M GEN. bRTER LEYEL. L10P L 2

Q ONCE Di $1LM CEN. TATER LEYfl. LhP 1 3

D ONGE DI PWES$JtIZER wmtJt LEYG.

1D0 0

2 2

2 3

x 3

3 3

l200 N

u-

_J x

-300

\\

N

-400

^

L 5

k i

00 h

0 LO 20 30 40 50 60 70 80 00 100 h

TIME, Stic SEQ. MIpatics(

14/10/78 (7. 4 4 is.

FIGURE 3.37 - Level Changes in Pressurizer and Steam Generators for Large Steamline Dreak M

M M

M

M M

FHRIRE ISLR0 + LFRCE STEDHLTtC BREN + 24 OCT 78 +

4 L

1 MODERATcll RERCTI g g

2 DOPPLER I ERCTIVITY 3

BORON RElicTIVITY 4.

TOTRL REllCITVITY 3

2 l

1

.m G

g 8

4 J

L 1

1 3

3 4

s 4

4 4

do

~

3 3

-1

~

~,

2 1

A Y

N T

N t

G h

' 0 ID 20 30 40 50 G0 70 80 30 100 TIME, SEC

$EQ. MIRKtCK 14/10/78 L7. 2 4..$.

FIGURE 3.38 - Nuclear Reactivity Feedback Effects for Large Steamline Break

PRRIRIE ISJto

9AL STERLDE BREDK + 24 OCT 78

140 L.

Powen Le n.

i 2

Henrrww 3

TOT A PR MARY COOLANT FLOW 4.

1011 FC DWRIER FLOW 5

TOTE STEnnLIHg FLOW 120 3

3 3

2 10 0 i

c80 b.

g o

70 ri 5

n.

40 20 5

5 5

s s

e L

L t

4_

a t a e

t a 4

L a e

t a e

t a 4

L a e

aa e

0 20 40 GO 80 100 120 140 ISO 180 200 TIME. SEC EEQ. MIRK 1PD 28/t0/78 La.s4.05.

FIGURE 3.39 - Pot.

, Heatflux and System Flows for Small Steamline Break M

M M

M

M M

M PRRIRIE ISRO + 9RL STERUtE BEK + 2B OCT 76 +

600 1

CORE INU T TEWORT@E 2

AVE. C RI CQXANT TEMF J.

CLAD TEWEM' TWE 590 580 570

=

Y5GO 5

E E

550 m

540 z

e O

2 2 a z3 1

, a a ts

,2 a

, 2 a 1

530

?

O 20 40 GO 80 100 120 140 160 180 200 uw TIME. SEC SEQ. MIRKLPD 2Gn004 12.54.05.

FIGURE 3.40 - Core Temperature Response for Small Steamline break

PRIERIE ISLRO + SmLL STERt.DE BREW

2B oCT 78 +

580 1.

Hai Ltc execantuas Lcoe t 2.

COLD LEC TEWERATJtE LOG 8 1 i

570 560 550 4

N

?

E 5x

$540 u.

1 L

L t

t t

t g

530 N

n 520

~

a n

n n

n

>57 5

O 20 40 60 80 100 124 140 1s0 180 200 TIME. SEC SEQ. 84 IRK 1PO ns /10/74 t2 54.05.

FIGURE 3.41 - Primary Loop Temperature Response for Small 5teamline Break M

M M

M

,3,

%e M

00 M

2 t

0 5

8 0

1 o

t 2L M

r z

o f

t 0B 1

8 s

7 r

/

M O

o L

t

/

a sE r z

e t

0 n

4 e

1 L 2 D G P

P 1

m ML O

K O

R a 3

I e M

t

.. E z

E E G

, S N

Q

%CA t

0 E

A A H 2

S d DOC 1

n E

a E E R R R U U U S r

S S S 3

M e

S S E E E R C

z

+

R R P e

E i

P P R

a a t

0 S r

8 E.

E E i i I

0 u

7 1

ct I s

s 1

0 s

T 0 O R p p E

s CO M M S M

e L

A A S 2 0 I

r E E E 0 8 a

T Pk G

T T R 1,

2, 2

S S P a

z

+

ne 1

2 L

i r K

0 B

G

. 2 3 L

8 s

e ee iB e r gn r u ni E

u s i

al b

s s h m 3

r P

,C I.

z s e C a M

M e r e

0 et T

l P

^

6 rS S

r u

e e sl m z sl U

a o i ea H

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z u

PS

+

m s a s 0

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t r

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M S P 4

LSI l

l 3

a a E

z i i E

R' t t R

I i i 0

U M

A n n 2

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I I

F M

\\

z 0

0 0

0 1

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5 6

7-0 2-eHmo.

M

g

7MLw z

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g g

00 2

1 g

0 5

8 0

4 5

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S I

8 r

7

/

o 0

f sg 1

/

S 1 2 E

r P P o

3

)

t X X 0

t 41 a

L L w

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L L.

L P

E e

E E V K

V V E R

e E E L I

L L M

G R

R R E m

E E T Q

t t A 1

ag 0

E

n. n W 2

S e

i t

N 1

.. 0 t

w w 2 S

E E I c r R U d

mms n

S

+

E t E C

a t i R E

s sP 0 S r

6

, i I 1

z 7

i l

0 e

rI TC E t E E

i D

c c ac M

r u w sg s

cmAt I

u 4

T 2

oa0 s

+

ea K

re I

0 Pr E

t z3 8

ng R

B i

E P

si l

el M

ag g

L H

0 n

1 6

5 h

C L

I i

ll F

e M

+

eg S

v

+

L 0

U W

4 U

3 M

SI 4

t 3

uu L

E a

0 R

2 R

P UG I

F

\\

L N

0 0

0 2

E 0

0 6

1 1

8 4

=

6o o YW M

M M

PRRIRIE IS.RO + SMU. STERLItE BRfmK + 26 OCT 78 +

4 1

MODERATGt REACTIVITY 2

f>0PPLOt QEACTIVITY 3

BORON REETIVITY 4

TOTAL Rt'4CTIVIT Y 3

2 i

t t

i i

k-V

'^'.

E J

t 3

2 3 2

2 2

2 8

a 3

3 3

-1 y

4 4

-2.

v z7 w

C"

-3 0

2D 40 60 80 100 22.0 140 ISO 180 2D0 d,

TIME. SEC SEQ. MIRKLPD 2s/10/78 12.54.d5.

FIGURE 3.44 - Nuclear Reactivity Feedback Effects for Small Steamline Break

E XN-NF-78-35 4.0 DISCUSSION OF RESULTS The transient analy.;is as performed by ENC for the Prairie Island Unit 1 Nuclear Power Plant ensures adequate margin to regulatory limits for the ENC fueled core for anticipated operating conditions. The following transients were analyzed using the ENC PTSPWR2 model.

1)

Fast Rod Withdrawal

2) Slow Rod Withdrawal 3)

Loss of Power to both Primary Coolant Pumps 4)

Locked Rotor in 0'e Primary Coolant Pump 5)

Loss of Electric Load 6)

Large Steam Line Break 7)

Small Steam Line Break These transients were considered because they were shown in the reference cycle analysis (3) to have the least margin to technical specifi-cation limits.

Table 4.1 provides a tabulation of the operational transients as analyzed in the FSAR for the reference cycle. The technical specification limits for the transients are a minimum DNB ratio of 1.30 and a peak pressure of 2750 psia.

In addition, for the small steam line break, an adequate shutdown margin must be demonstrated such that the reactor does not become critical following the break.

Table 4.2 shows a comparison of general operating parameter values for the reference fuel cycle and for the ENC fuel cycle. The data in Tables 4.1 and 4.2 illustrate that the parameter values used in the Cycle 5 analysis for most cases are either equal to the reference data, or they are enveloped by them.

The only exception is the positive inoderator coefficient at the beginning of Cycle 5.

This means that under most comparable transient conditions, the response of the ENC fuel is either XN-NF-78-35 enveloped by or equivalent to the response of the reference cycle fuel.

The transient analysis of the reference cycle indicated that the heat flux margin to DNB is mt limiting in the locked pump rotor case, 2 Class IV event.

Likewise, the ENC analysis showed it to be the most limiting event.

It is the only transient where the DNB heat flux ratio is calculated to be below 1.3.

The minimum DNB value calculated is 1.09 (see Table 1.1).

A statistical analysis shows that fewer than 1 percent of the fuel rods are like'y to experience DNB during this event.

If the same transient is analyzed with the calculated full power

-6 moderator coefficient of -16.9 x 10 /F, the heat flux margin to departure of nucleate boiling is 1.19.

The conservative case with a moderator

-6 coefficient is +20 x 10 /F has been reported here, and the following points apply:

1.

The design value of the moderator coefficient at the beginning

-0 of Cycle 5 is +16.6 x 10 1/F. To cover measurement uncertainties,

-6 an analysis value of +20 x 10 1/F has been chosen.

The design value, however, applies to hot standby conditions. As reactor power increases, the coefficient decreases, turt lg negative at about 70 percent power, and reaching a calculated 31ue of

-6

-36.9 + 20 x 10 /F at full power.

6 2.

For the Plant Transient Analysis, the design value of 63.2 x 10 lb/hr has been used for the reactor flow, which is only 9c' percent of the actual plant flow as measured and documented in the Technical Specifications (4).

XN-NF-78-35 3.

The setpoint for the low loop flow trip function is decreased for the analysis in order to envelope instrumentation errors; 87 percent is used rather than the plant value of 90 percent.

In addition, the design flow value of 68.2 M lb/hr has been used instead of the measured plant flow which is at least 75.8 M lb/hr.

In the reference cycle analysis, the following transients also showed a reduction in MDNBR from steady state conditions:

e Startup of Inactive Loop e Feedwater System Malfunctions e Excessive Load e loss of AC power For the reference cycle, the lowest MDNBR during a Class II or III incident was 1.61 for the 2 pump trip incident.

These transients were not reanalyzed because they did not result in as large MDNBR changes as those analyzed in this report and thus were not limiting in the reference cycle analysis and would not be limiting for an ENC fueled core either.

Since the system response for these transients is insensitive to the fuel type, the only variation in results would be the DNB ratio.

Table 4.3 compares the neutronic parameter values of the reference cycle analysis to the ones for the Cycle 5 analysis. As pointed out earlier, conservative values for the moderator and Doppler feedback coefficients have been used in the analysis.

In the reference cycle analysis, the rod withdrawal transient has been analyzed for a spectrum of reactivity insertion rates from ik = 10-6 1/sec to lk = 10-3 1/sec.

This spectrum of insertion rates is covered by two trip functions:

the overtemperature AT trip for low inser-tion rates and the high nuclear flux trip for high insertion rates.

I XN-NF-78-35 At full power, the crossover point between these 2 functions is at 2.7 x

-5 10 1/sec which corresponds to the point of lowest DNB margin for the high nuclear flux trip regime.

For insertion rates below the crossover value, the the minimum DNB heat flux ratio stays almost constant, and for insertion rates above it, the MDNBR rapidly increases due to the fast acting high flux function. Going to partload operation changes the plant response somewhat.

In the reference cycle analysis, the response spectra for 60 percent power and 10 percent power are shown. The crossover value moves to ik = 1.5 x 10-4 1/sec for 60 percent power and up to dk = 2.8 x 10-4 1/sec for 10 percent power.

Going to partload increases the margin to DNB flux in the regime of the high nuclear power function, and it slightly lowers the MDNBR in the overtemperature AT regime from a typical value of 1.34 down to 1.30.

The ENC analysis has shown that the rod withdrawal transient is not limiting at 102 percent of nominal power.

Since the change in plant response caused by lower power levels is mainly dependent on the plant protection system, it can be expected that the response trend in partload cases for Cycle 5 fuel is analogous to the trend for the refer-ence cycle fuel.

Therefore, adequate protection is ensured over the complete range of; power levels.

A malfunction of the chemical and volume control system is also en-veloped by the rod withdrawal transient.

During this malfunction, reactivity is added to the core by addition of unborated primary coolant makeup water. The plant response is similar to that for the slow rod withdrawal I

I XN-NF-78-35 I

transient analyzed in Section 3.1, except that the rate of reactivity insertion is lower. A typical boron dilution event would cause a reactivity insertion at ik = 10-5 1/sec.

At all power levels, this insertion rate I

falls into the regime of the overtemperature AT function.

The plant response for this event would be identical to the slow rod withdrawal case (at ik=10-5 1/sec) analyzed in Section 3.1.

Certain operational incidents are not dependent on fuel type. These include:

I e RCCA Misalignment e Turbine Generator Overspeed o Fuel Handling Incident e Accidental Waste Gas Release e Radioactive Liquid Release e Steam Generator Tube Rupture These incidents as discussed in the reference cycle analysis were shown to be protected for any fuel type by administrative controls, redundancy of alarms, and/or integrity of system components.

The conclusions drawn for these incidents as given in the reference cycle analysis are valid for Cycle 5 and all future reload cycles with ENC fuel.

I I

I

TABLE 4.1 COMPARISON OF TRANSIENT-SPECIFIC INPUT PARAMETERS Reference PTS Analysis for Cycle Cycle 5 ENC Fuel Moderator Doppler Moderator Doppler Coefficient Coefficient Coefficient **

Coefficient 6

0 6

6 (to/ F x 10 )

(to/ F x 10 )

(M/ F x 10 )

(ts/ F x 10 )

Rod Withdrawal From Full Power 0.0 small

+20.0

-10.0 From Reduced Power 0.0 small

,m Loss of Flow

?

Pump Coastdown 0.0

-16.3

+20.0

-15.0 Locked Rotor 0.0 NA*

+20.0

-15.0 Inactive Loop Startup

-40.0

-10.0 Loss of Load 0.0 NA*

+20.0

-10.0 Loss of Feedwater NA*

NA*

Excessive Feedwater 0.0 and -40.0 NA*

E Excessive Load increase 0.0 and -40.0 NA*

3 Steam Line Break Variable Variable Fig. 3.32 Fig. 3.31 a

Information not available in reference cycle analysis

- See disc.ussion of moderator coef ficier, on page 65 M

M M

E Xft-flF-78-35

~70-TABLE 4.2 COMPARISON OF OPERATING PARAMETERS PRAIRIE ISLAND UNIT 1 AND UNIT 2 I

Reference Cycle 5 With Cycle ENC Fuel Core Total Core Heat Output, MW 1650 1650 Heat Generater; in Fuel, percent 97.4 97.4 System Pressure, psia 2250 2250 Hot Channel Factors

Total Peaking Factor, Ff 2.80 2.32 Enthalpy Rise Factor 1.58 1.55 Axial Peaking Factor, F l.72 1.45 Z

Location of Axial Peak, ft NA**

6.2 6

6 Coolant Massflow, lb/hr 68.20 x 10 68.20 x 10,,,

6 6

Effective Core Massflow, lb/hr 64.64 x 10 64.64 x 10 Reactor Inlet Temperature, F 535.5 530.5 Heat Transfer Average Heatflux, Btu /hr-ft 190,973 190,973 Hot channel factors as applied to safety analysis and thennal-hydraulic analysis only.

Information not available in reference cycle analysis.

- - This is the design value. The actual reactor flow as stated in the Technical Specifications and confirmed by measurements is at least 6

75.8 x 10 lb/hr.

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I

-71 XN-NF-78-35 I

TABLE 4.2 (Continued)

COMPARISON OF OPERATING PARAMETERS FOR PRAIRIE ISLAND UNIT 1 AND UNIT 2 I

Reference Cycle 5 With Cycle

_ ENC Fuel Steam Generators 6

6 Total Steam Flow, lb/hr 7.080 x 10 7.091 x 10 Steam Temperature, F 510.8 510.9 Steam Pressure, psia 750.0 750.0 Feedwater Temperature, F 427.3 427.3 I

I I

I

M M

M TABLE 4.3 COMPARISON 0F KINETIC PARAMETER VALUES FOR PRAIRIE ISLAND UNIT 1 AND UNIT 2 Cycle 5 Reference Cycle with ENC Fuel BOC E0C B0C E0C Moderator Temperature

-6*

-350 x 10

+20 x 10

-6

-6*

-350 x 10 Coefficient in 1/5

+30 x 10 Moderator Pressure

-6

-6 Coefficient in 1/ psia

-0.3 x 10

+3.5 x 10

-0.2 x 10

+4 x 10

-6

-0

-6

-6 Doppler Coefficient in 1/F

-10 x 10

-16 x 10

-12.5 x 10

-16 x 10 L

-3

-3 7

Delayed Neutron Fraction 7.1 x 10 5.1 x 10 6.1 x 10 5.1 x 10 Value for hot standby conditions.

E ua w

XN-NF-78-43 5.0 SIMULATION CODE CHAliGES The basic digital plant simulation code as documented in Reference 1 has been used in performing the plant transient analysis for the Prairie Island plant.

Starting from the version PTS-PWR2-NOV76A, several code changes have been inplemented resulting in the version PTS-PWR2-NOV78.

All changes were restricted to the initialization modules of the code.

Therefore, the dynamic plant model of the PTS code was not affected. The purpose of the changes was (a) to remove a number of redundant variables from the input list and generate them internally in the code and (b) to redefine some input parameters such that hand calculations are eliminated or reduced. All code changes have been checked individually.

In addition, the pump seizure transient for the R. E. Ginna plant has been rerun, and the results were found to be very close to previous results.

Some key results are shown in Table 5.1.

An alphabetic list of the affected variables is shown in Table 5.2.

I I

i XN-NF-78-35 I

TABLE 5.1 Comparison of Results for the RE Ginna Pump Seizure Transient Version Version E

PTS-PWR2 NOV 76A PTS-PWR2 NOV 78 5

Minimum DNB Flux Ratio 1.23 1.23 Maximta Reactor Power, %

102.

PeakValueofAgerageCoreHeatflux, 181,163.

181,160.

btu /(hr.s ft )

Reactor Flow at 5 sec., %

49.

Peak Core Average Temperature, F 590.

5 91.

I I

XN-NF-78-35 TABLE 5.2 LIST OF CODE VARIABLES REMOVED REMOVED FROM INPUT OR REDEFINED BETA KSHTB NBETAl WBDGlI KSLSH1 NBETA2 WBDG2I CFSPR KSLSH2 NBETA3 WDOSLR CPWPR KUPSP NBETA4 WDOSL1 CFWPRS Kill 2 NBETAS WDOSL2 K2122 NBETA6 WFWMAX DOSLll WFW1 DOSL2I LEVGlI WFW2 I

LEVG2I Q0AR WIVIIr.

FWCIC2 LT0PSG WIV2IL SLSHll WLPCRR HFW1IC MD01IC Si.SH2I WLP1IC HFW2IC MD02IC WLP2IC HGUP1I MSG 1IC T1POIC WTBMAX I

HGUP2I MSG 2IC T2P01C WS0 HUP 1IC MSHIC T1P3IC HUP 2IC MSLllI T2P3IC HWPRIC MSL12I I

MSL21I UPSPll KBDSG MSL221 UPSP21 KBDTD MSPRIC I

KDOSL1 MUPlIC KDOSL2 MVP2IC KSGUP MWPRIC REDEFINED VARIABLES Variable New Definition CFWPRS Setpoint for High Pressurizer Level Trip, in Percent of Span HFWIIC Feedwater Temperature for Both Coolant Loops, In Fahrenheit MWPRIC Initial Pressurizer Level, in Percent of Span NBETAl Delayed Neutron Fraction, Group 1 NBETA2 Delayed Neutron Fraction, Group 2 NBETA3 Delayed Neutron Fraction, Group 3 NBETA4 Delayed Neutron Fraction, Group 4 I

NBETA5 Delayed Neutron Fraction, Group 5 NBETA6 Delayed Neutron Fraction, Group 6 I

All variables not redefined have been removed.

I I

I XN-NF-78-35 I

6.0 REFERENCES

1.

Kahn, J. D., Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors PTSPWR I

XN-74-5, Revision 1, May 1975.

2.

Galbraith, K.

P., et al., Definition and Justification of Exxon Nuclear Company DNB Correlation for Pressurized Water Reactors, XN-75-48, October 1975.

3.

Northern States Power Company, Prairie Island Nuclear Generating Plant, Units 1 and 2, Final Safety Analysis Report.

4.

Northern States Power Company, Prairie Island Nuclear Generating Plant, Units 1 and 2, Technical Specifications, Docket 50-282 and 50-306.

I 5.

Kahn, J.

D., Assumptions Used in the Plant Transient Analysis for the Donald C. Cook Unit 1 Nuclear Power Plant, XN-76-35, Suppiement 1, November 1976.

6.

Lyle, J. M. and Killgore, M.

R., Prairie Island Unit 1 Cycle 5 Fuel Management Analysis, Exxon Nuclear Company, XN-NF-78-43, October 1978.

I I

ML19276D708

Contents

Text

1.0 INTRODUCTION

SUMMARY

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

SUMMARY

6.0 REFERENCES

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ML19276D708

Text

1.0 INTRODUCTION

SUMMARY

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

SUMMARY

6.0 REFERENCES

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