ML20236L544

Rev 0 to Analysis of Revised Low Setpoint for High Flux, Power Range

ML20236L544
Person / Time
Site:	Prairie Island
Issue date:	11/03/1987
From:	Richard Anderson, Bonneau C, Shandeth Montgomery NORTHERN STATES POWER CO.
To:
Shared Package
ML20236L521	List: ML20236L519 ML20236L528 ML20236L544 ML20236L589
References
NSPNAD-8725, NSPNAD-8725-R, NSPNAD-8725-R00, NUDOCS 8711100398
Download: ML20236L544 (12)

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.k Prairie Island Unit 1 and 2 l.

Analysis of ths' ,

Revised Low Setpoint for High Flux, . Power Range 1

' N SPN AD-8725' Revision 0 +

October 1987 l

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Prepared by jd ha 0

Date # /16/97 Reviewed by [

-, ,e-Date f,o/rA/r 7 Approved by / #d,,f[, 2/ M - Date /// h-

- - a z 8711100398 871103 4 PDR ADOCK 05000282 P PDR Page 1 of 12 1

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LEGAL NOTICE l 4 L

l This report was. prepared by, or.on behalf, of. Northern States" l Power Company.(NSP).- Neither NSP,.or' any.' person acting ontbehalf.

of NSP: '

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a. Makes any warranty or representation, express or'

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implied, with respect to th'e accuracy, dompletene'ss,,

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l usefulness, or'u's of any information, apparatus, method:

or process disclosed or' contained in this report, or 1 that the.use'of any such information, apparatus, method .

or process may not infringe privately owned rights; or-

b. Assumes any liabilities with're'spect to the use of, 3

or for damages resulting from'the use of,~any. .

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l information, apparatus, method, or process disclosed in the report. -I l

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. 7 TABLE.0F' CONTENTS

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SUMMARY

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'2;0 CALCULATI0'NAL H0DELS' AND METHODOLOGY:

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2.2- Methodology w1 5 3.0~ ACCIDENT AND TRANSIENT ANALYSIS '6?

3.1 Uncontrolled RCCA Withdrawal From Subcritical 6- q 1

3.1.1 Input Parameters- Y 6l -

3.1.2 Results 7 '- )

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3.2 Rod Ejection Analysis 17 0

4.0 REFERENCES

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SUMMARY

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-l This repo'tr summarizes the calculations performed by the Northern' States Power Nuclear' Analysis Department (NSPNAD)11n support uf.a change in the setpoint for reactor trip- 4 on High Flux, power range (low setpoint) in the Technical Specifications from 25% to

~40% of rated power. ,

NSPNAD has analyzed the transient;_ response of Prairie Isli.nd for.the.40% setpoint.

The results show that all transient acceptance criteria are' met'with the new setpo' int.;

Therefore, no unreviewed safety questions are involved. This analysis covers 'both the Westinghouse OFA-fuel assemblies and the ENC.TOPROD assemblies.

J Section 2 of this report describes the methodology, assumptions'and calculations models used in this analysis. )

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l Section 3 contains transient analysis results.

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'2.0 CALCULATIONAL MODELS AND METHODOLOGY. l l

2.1- Calculational Models '

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'Theseanalyses'wereperformedusing.themodellingmelhSdsdescribedin;the'.NSPNAD

)

ReloadSafety'EvaluationMethodsforPWRs.(Reference 1)..Thesemethods'havebeen~

reviewed and approved by the.NRC.

2.2 Methodology i

After reviewing all of the transients ih the USAR (Reference 9), NSPNAD has

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analyzed-the events which are affected by a changelin the.setpoints for a' reactor. ,j trip on High Flux, power range' trip'(low setpoint). 'These transients were the

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Uncontrolled RCC Assembly Withdrawal From a Subcritical. Condition 'and Rupture of l]

l' Control Rod Drive Mechanism' Housing (RCCA Ejection) from hot;zero power'(HZP) conditions. '

SinceanUncontrolledRCCAssemblyWithdrawalLFromaSuberiticalConditionisan.

unplanned event postulated to occur at least once during the life of the plant, it is considered a' condition III event. The Rupture of a: Control Rod Drive l L Mechanism Housing event is much less probable and is a condition IV accident.

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3.0 ACCIDENT AND TRANSIENT ANALYSIS f

3.1 Uncontrolled RCCA Withdrawal From Subcritical ,

This documents the NSP Nuclear Analysis Department's analysis of plant ;i operational transients effected by the proposed increase in the maximum allowed i High Flux, Power Range (low setpoint) from 25% of rated full power to 40%.

This transient is a condition III event with the following acceptance criteria:

i) Primary and Secondary system pressure will not exceed 110% of their  !

design values. These specific values of the limiting pressure are' 2750 psia (110% of 2500 psia) for the primary systen and 1210 psia j (110% of 1100 psia) for the secondary system.-

ii) Fuel cladding integrity will be maintained by limiting the MDNBR to be greater than 1.30 (calculated using the W-3 correlation) for ENC fuel and 1.17 (calculated using the WRB-1 correlation) for Westinghouse fuel.

3.1.1 Input Parameters The only plant operational transient affected by the change in the High Flux, power range trip (low setpoint) is the Uncontrolled RCCA Withdrawai i From Subcritical. The value used was the nominal Tech. Spec. value, 40%.

Another case was run with a 10% uneartainty factor added to the setpoint to account for instrument drift and error, however the effect of this change was negligible. At the time of the trip, the neutron flux is increasing extremely quickly (so that the time to reactor trip is not significantly affected) and the flux increase is turned around by the Doppler feedback and not by control rod motion. The conclusions drawn from the results of the analysis using a 40% setpoint are still valid.

In the analysis, this transient is initiated from HZP conditions as the heat generation rate before the reactor becomes critical is very small compared to the rate during the power excursion. The core.is assumed to be critical at 10 -12 nominal power at the time the rod withdrawal begins.

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This analysis was performed using the most conservative combination of j physicsparametersfromPrairieIsland1. Cycle 12(References 2through6) q and Prairie Island 2 Cycle 12 (Reference 7 and 8) and bounds both cycles.

One change was made to the data in references 2 through 8: the reliability l factor for the Doppler coefficient was increased to 25% in accordance with NSPNAD answers to NRC questions regarding the Monticello reactor physics l methods topical (Reference 10).  !

l 3.1.2 Results I This transient assesses plant response to a control' rod withdrawal, with

~4 a reactivity insertion rate of 8.2x10 ok/sec, from a.P2P condition.

The assumed reactivity insertion rate is greater than.the simultaneous withdrawal at maximum speed of the combination of the control rod banks having the greatest worth and, therefore, is a bounding value for Prairie l Island analysis. All automatic reactor contro1' systems are' assumed to be inoperable during this transient.

The transient response of the NSSS for this case is shown in the I l

Figure 3.1. The reactor trips at 7.58 seconds on high flux (low setpoint). The maximum pressurizer pressure is 2459 psia at 11.2 seconds, l The maximum steamline pressure was 1077 psia at 18.8 seconds. Thus satisfying both acceptance criterian for the primary pressures. The peak heat flux is less than 50% of the nominal value, while the av'erage coolant temperature in the vessel is 562 F compared to 566 F for the nominal conditions. Although the power increases rapidly, the power distribution

( becomes only slightly more highly peaked as FAH increases to 103% of the nominal value used in the RSE analyses. FQ is still bounded by the nominal value. This combination of thermal power, coolant temperature and only slightly increased peaking factors results in MONBR values well above l the limiting values for the W-3 and WRB-1 CHF correlations. The peak clad temperature (590 F) is less than the nominal full power value (roughly 625 F) and, thus, there is no possibility of clad damage.

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e Therefore this transient meets;all applicible acceptance criteria.

3.2 Rod Ejection Analyses A Control- Rod Ejection Accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a. Rod ClusterL

. Control- Assembly-(RCCA) and drive shaft. The consequence of.this mechanical ,

failure is a rapid. reactivity insartion together with an adverse core. power distribution, possibly leading to. localized fuel . rod damage.

The acceptance criteria for the control rod ejection accident-('a. condition IV event) are as follows:

1. The average hot spot fuel.enthalpy.must be less than 280' calor'ies/ gram.

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2. The maximum reactor coolant. system pressure must tib less than the pressure .

that will cause stresses to exceed the emergency condition stress limit; assumed to be 120% of design pressure (2500 psia). (In this analysis, the-highest pressure calculated in any of those cases was 2502 psia). i

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3. The maximum clad temperture calculated to occur at the core hot spot must- l l

not exceed 2750 F. l

4. The number of fuel rods calculated to experience a DNBR of less than 1.3 (W-3) or 1.17 (WRB-1), whichever.is applicable, should not exceed.the '

number which are required to fail in order that the doses due to released activity will exceed the limits of 10CFR 100. This limit is currently the q maximum number of failed fuel rods calculated in the USAR (Reference 9). .

As the control rod ejection accident is a very localized.' power excursion, the number of failed fuel rods is not explicitly calculated in every: case.

The rod ejection accident has been evaluated with the procedures developed'in ,

Reference 1. The ejected rod worths, hot pellet peaking factors,: delayed j

neutron fractions and Doppler coefficients were taken as conservative values J which bound Prairie Island analysis. The values for the High~ Flux (low .

setpoint) and Doppler reliability factor were the same as those listed in Section 3.1.1. -

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d-2.Theejected. rod.fromHZPinitial/ conditions.is.itierminated.by.'the!1cwsetdoint.

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for the high neutronLflux trip. -The' pellet stored energy and peak-clad l y+, > >

.' temperature;were evaluated explicitly;at BOC'and E00'o conditions...;The'BOCrease usedthe-mostconservativeJcombihation.ofvalues,1fromLPIl1 Cycle:12-andPIK20 * , !

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. Cycle 12.' Themaximum.fuelstoredenergywascalcula6.....tedtobe"146. cal /glxl? '

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(limit: f 280 cal /g) and the ' peak clad temperature was' 2656*Fq(limit: L27507F).

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'The rod ejection from'HZP at EOC conditions was' analyzed sejiarately;for.PIf1) '

Cycle?l2 and PI 2 Cycle 12L For PIL 1;Lth'elmsdimum fuellstor$d  ! energpfwab / 'q

~ +. ,. ., , it , L' [h ' :M< I i 125 cal /g~(limit _280 cal /g) and the<pe'ak clad' temperature was.2125*F f  :

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and Lthe peak clad temperature was 2600 F. g" 7 y gj6 ,4.1;L

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Therodejection'accidentwasfound.toresult'..inamaximumstored/energpioflless.

than 280' cal /g limit inLa11 cases. Thehakcladtemperatures.are!1esdthanthe' i limit of 27501 F. The- significant' par $betersL for the"imalysis:,calodhw'ith the ,

results are summarized in. Table 3.1.

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i TABLE'3.1 Prairie Is1'and Units 1 and 2,' Cycle'12' Ejected Rod ~ Analysis, HZP-BOC EOC PI 1 & 2 l Cycle 12- PI 1 Cycle'12 PI 2 Cycle 12 Inpul Maximum Control Rod Worth (pcm) 831 831 804 Doppler Defect (0-100% power pcm)- 1288 1899 \ 1798 Shutdow Margin (pcm) 2415 2737 2383 Delayed Neutron Fraction 0.005634 0.004837 0.004863 Power Peaking Fractor, Fg 6.277 10.6 10.95 I

Results Maximum Fuel Enthalpy (cal /g) 146 125 151 Peak Clad Temperature ( F) 2656 2125 2600 l

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Rod Wittidrakal from Subcritical P 3 Absolute Power J Core Average H: sat Flux 10 -.

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g s't e l' 4.0~'REFERENCESL 1.) NSPNAD-8102A, Rey, 5, " Reload Safety Evaluation. Methods for Application to PI,-

' Units",' March 1987 2.) . Internal Correspondence, C A Bonneau to D A Rautmann, " Transfer. of CAS Input to

~

SAS for PI 1 Cycle'12 Reload Evaluation", December 12,.1986.

. 3.) ' Internal Correspondence, C A Bonneau to'D A'Rautmann, " Error in FRDR'CAS-SAS' Transfer Letter", December:19, 1986, 4.) Internal Correspondence, .T L Breene' to' ClA Bonneau, Transfer of. Revised CAS.

Input to SAS for PI 1 Cycle 12-Reload Safety Evaluation, Rev '1", March 25,.1987.'

i 5.) Internal Correspondence, T L Breene to C-A Bonneau, " Transfer-of Additional: CAS' Output to SAS for PI 1 Cycle 12 Reload Safety' Evaluatidn, Rev. -2",' July.14,1987.

l 6.) Internal Correspondence, T L'Breene to C A Bonneau, " Error in FRDR.Rev. 2' CAS-SAS Transfer Letter", August 7, 1987.

7.) InternalCorrespondence,TL'BreenetoCABonneau,"TransferoflCASInputto .

SAS for Prairie Island 2 Cycle 12 Reload Safety Evaluation", August.19, 1987. i 1

JJ 8.) Internal Correspondence, T L Breene to C A Bonneau, " Transfer of Additional CAS f Input to SAS for Prairie Island 2 Cycle 12 Reload Safety Evaluation",

l. September 10, 1987.

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9.) Northern States Power Company, Prairie Island Nuclear Power Plant, Updated Safety l- Analysis Report.

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10.) Letter, D. M. Musolf (NSP) to' Director, Office of Nuclear Reactor Regulation ]

(NRC), " Response to NRC Questions on the.NSP Reload Safety Evaluation Methods for Monticello, NSPNAD-8608," September 29, 1987. ,

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