ML20150F265
| ML20150F265 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/05/1988 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20150F258 | List: |
| References | |
| NUDOCS 8807180206 | |
| Download: ML20150F265 (12) | |
Text
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Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated July 5,1988
- Proposed Changes Marked Up on Existing Technical Specification Pages Exhibit B consists of'the existing Technical Specifications pages with the-
. proposed changes written on those pages. Existing pages affected'by this-
- change,are. listed below:
TS-x TS.3.10-1 TS.3.10-2 TS.3.10-9 (and insert)
Figure TS.3.10-8 (to be deleted) 1
'; 8807180206 880705 PDR ADOCK 05000282 '
P. PDC
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, TS.3.10-1
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3.10 CONTROL ROD AND PC'.iER DISTRIBUTION LIP.ITS Aeolicability, ,
Applies to the limits on core fission power distribution and to the 11=1ts on control rod operations.
Objectives To assure 1) core suberiticality af ter reactor trip, 2) acceptable core power distributions during power operation, and 3) li=ited potential reactivity insertions caused by hypothetical control rod ejection.
Soecification A. Shutdown Reactivity
! The shutdown margin with allowance for a stuck control rod assembly
( shall exceed the applicable value shown in Figure 75.3.10-1 under all stesdy-state operating conditions, except for physics cases, from zero to full power, including eff ects of axial power distribution. The shutdown margin as used here is defined as the a=ount by which the .
reactor core veuld be suberitical at hot shutdown conditions if all control rod asse=blies were tripped, assu=ing that the highest worth control rod asse=bly re=ained fully withdrawn, and asse=ing no changes in xenon or boren concentration.
B. Pever Distribution Linits
- 1. At all times, except d'q ring ley power physics testing, measured hot channel factors T and T. , as defined below and in the bases, shall meet the holleving li=its:
N f 450 l T.Q x 1.03 x 1.05 <(7 'TaH}/P]K(Z)
Q 7
rNaa
- l 0' 1 an Q- T )
- U + 0*3Cl-?)I
- f. 70 where the folleving definitions apply:
I
- K(Z) is the axial dependence function shown in Figure TS.3.10-5. .
- Z is the core height location.
l - P is the fraction of yrsted power at which the core is operating. In the T limit determination when P < .50, see S
P = 0.50.
- !!.e 7 0 ""' "" '" ' = "" ~
- ir. Tigare TC.3.10-Q. AH aH AH Q
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t
, ,c
> . TS.3.10-2 nj ""
J 01 7/0/07 N
-F or F is defined as the measured 7 or F wkththesmallestma,rginorgreatestekcassb$respectively,. li=1t.
e
- 1.03 is thg' engineering hoc channel f actor, F", applied measured T to account for manufacturing tolerr.nce.
to the q
N -
- 1.05 is applied to the measured T coaccounciformeasuremene q
uncertaincy.
- 1.04 is applied to the =easured _N 1 3g to account for measurement
, uncertaincy.
N. N
- 2. Hot channel factors T oand T g, shall be =easured and the target flux dif f erence deter =rned, at equilibrius conditions according to the following conditions, whichever occurs first:
(a) At least once per 31 ef f ective full-power days in conjunction with the carget flux difference deterzination, or (b) Upon reaching equilibrium conditions af ter exceedL=g the reactor power at which target flux difference was last decernined, by 10~ or more of raced poveJ.
f%
' -' (- ,1) shall meet the following li=it for the middle axial 80~
i'9, o : ore: 2.60
. T9q (equil) x V(Z) x 1.03 x 1.05 5 (~q(~331/P] x K(Z) l vhere V(Z) is defined Figure 3.10-7 and other ter=s are defined in 3.10.3.1 above.
- 3. (a) If either measured hoc channel factor exceeds its li=it specified in 3.10.3.1, reduce reactor power and the high neutron f1* g x trip setpoint by 1 for each percent that the measured FY exceed 7"hthe3.10.3.1Itnit.or byThen 3.33"follov for each percent that the =easured 3.10.3.3(c).
-H N
(b) If the measured T exceeds c~ae 3.10.3.2 li=1:s but not the 3.10.3.1 limi2,(equil) take one of the following accions:
l
- 1. Within 48 heurs place the reactor in ar. equilibrien configuracion for which Specification 3.10.3.2 is sacis-t fied, or 2.
Reduce reactor power and the high neutron flux trip
~
sqcpoint by 1" for each percent that the measured '
l TQ (equil) x 1.03 x 1.05 x V(t) exceeds the 11:1c.
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! TS.3.10-9
?:7 '!3!35 (q,:r,,,
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- mechanical properties to within assumed design criteria. In 2dditier,
. -limitin;; the p::h lin::: p=:e-dencity durin;; C:nditi:n ! cvenee- g 7 .j pr cid:: ::::::::: th:t th: initi:1-c:ndici :: ::::::d f:: th: LOC.'.
reslyce 2re ser ?-d the ECCS ::ceptsree criteri: 14-4: ef 2200*F 1:
- t exceeded.
During opegation'Nthe plant staff comperes the measured hot channel and F factors, transient F'knd LOCK a,na(described lyses. later)ontothe The terms theright limitsside determined of the in the equations in Section 3.10.B.1 represent the analytical limits. Those l
terms on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.
F is the measured Nuclear Hot Channel Factor, defined as the maximum 19ealheatfluxonthesurfaceofafuelroddividedbytheaverage heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment ,
a c B b%et The K(Z) function shown n Figure TS.3.10-5 is a normalited function d that limits Fn axia11y. The K(Z) 'specified for the lowest six (6) feet l of the core 13 abritrarily flat since the lower part of the core is generally not limiting. Above that region, the K(Z) value is based on
. large tad small break LOCA analyses, o
j V(Z) is an axially dep i
measured yFI to bound F'qndent 's that couldfunction applied to at be measured thenon-equilibrium equilibrium conditions 9 This funck1on is based on power distribution control analyses that evaluated the effect of burnable po , rod position, ardel efftets, and xenon worth. ; ffay .
v '
F", Engineering Heat Flux Hot Channel Facpfr, is defined as the allow-ance on heat flux required for =anufactyfing tolerances. The engi-nearing factor allows for local ve;iati:ns in enrich =e'ut, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is s factor of 1.03 to be applied to fuel red surface heat flux.
The 1.05 eultiplier accounts fe uncertainties associated with measure-ment of the power distribution with the me ::ble incere detectors and the use of those measurements to establish the asse=bly local power distribution. is
' 1YoVa vfl (equil) is the measured limiting N obtained at equilibrium conditions d ring target flux determination. :9 N
F' , Nuclear Enthalov Rise Hot Channel Factor, is defined as the ratio oh the integral of linear power along the rod with the highest integrated power to the average rod power.
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The ECCS analysis
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One calculation at the 95%
was performed in accordance with SECY 83-472.
probability level was performed as well as one calculation with all the a 1-required. feat.ures of 10 C' R Part 50, Appendix K. The 95% probability level calculation used a peak linear-heat generation rate of 14.2 kw/ft. The Appendix K calculation used a peak linear heat generation rate of 15.8
- limit kw/ftoffor2.5 Fn limit theduring of 2.5. Maintaining 1) peaking factors below the Fq all Condition-I events and 2) the peak linear heat I generation rate below 14.2 kw/ft at the 95% probability level assures compliance with the ECCS analysis. ,
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FZGURE '5S.3010-8 E.
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/ I Op pf ation NOT E'
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F 3g(Fo ) W F
FIGURE 3.10'-8 Acceptable Values of Fq(FAH) and FAH(Fq) I l
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_ ______.____.__.__._________.______________m_ _ _ _ _ - _ - _ _ . _ _ _ . _ _ _ _ _ _ __m.m-_ma__ -_____mm._.m_-.A-__. _m b_L AAME
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l Exhibit C Prairie Island Nuclear Generating Plant License Arnendment Request Dated July 5,1988 Exhibit C consists of the proposed Technical Specification pages with the changes shown in Exhibit B incorporated. The proposed pages are listed below:
TS-x TS.3.10-1 TS.3.10-2 TS.3.10 9
TS-x APPENDIX A TECHNJCAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, inermal and hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations l 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations
! 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus' Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I 131 I 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents
( 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration i' 3.10-2 Control Bank Insertion Limits l 3.10-3 Insertion Limits 100 Step overlap with One Bottomed Rod
! 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10 7 V(Z) as a Function of Core Height 4.4-1 Shield Building Design In-Leakage Rate l 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Organizationa 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization '
for On-site Operating Group l
- y. n: e TS.3.10-1
- 3.10 CONTROL ROD AND POWER DISTRIBUTION LIHITS Acolicability
-Applies to the limits on core fission power d'istribution and'to the limits.
-on control rod operations.
' Objectives s
To assure '.) core subcriticality after readtor trip, 2) acceptable core power distributions during power operation, and 3) . limited potential reactivity insertions caused by hypothetical control -rod ejection.
Soccification A Shutdown Reactivity
'The shutdown margin with. allowance for a stuck control rod assembly-shall exceed the applicable value shown in Figure TS.3.10 1 under all steady-state operating conditions, except for physics 2t ests, from zero to full power, including effects of axial power distribution. The-shutdown margin as used here is defined as the amount by which the reactor core _would by suberitical at hot shutdown conditions if all control rod, assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration.
- B.-Power Distribution Limits
- 1. At all times, except during low power physics testing, measured hot. channel factors, andF$H,asdefinedbelowandinthe bases, shall meet the o11owing limits:
Phx1.03x1.055(2.50/P)K(Z)
F$H x 1.04 51.70 x [1 + 0.3(1-P))
where the following definitions apply:
- K(Z) is the axial dependence' function shown in Figure TS.3.10-5.
i l - Z is the core height location.
- P is the fraction of rated power at which the core is operating. IntheFhlimitdeterminationwhenPS.50, set P - 0.50.
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TS.3.10-2 e t
- L q N
, -F OorFfig is defined as the measured Fq or F3gg respectively with~the smallest margin cr greatest excess or' limit.
-1.03istheengineeringhotchannelfactor,Fh,appliedtothe measured (toaccountfor--manufacturingtolerance.
-1.05Isappliedtothe, measured (toaccountformeasurement uncertainty.
-1.04~isappliedtothemeasuredFfg;toaccountformeasurement uncertainty.
- 2. Ilotchannelfactors,(anddH,shallbemeasuredandthetarget flux difference determined at equilibrium conditions according to the following conditions,whichever occurs first:
(a) At lecer once per 31 effective full-power days in conjunction with the . target flux difference' determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10% or more of rated power.
F$(equil)shallmeetthefollowinglimitforthemiddleaxial80%
of the core:
((equil)xV(Z)x1.03x1.05<(2.50/P)xK(Z) where V(Z) Is defined Figure 3.10-7 and other terms are defined'in 3.10.B.1 above. ,
- 3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1. reduce reactor power and the high.
neutron f x trip setpoint by 1% for each percent that the <,
measured or by 3.33% for each percent that the measured N
Fagg exceed the 3.10.B.1 limit. Then follow 3.10.B.3(c).
(equil) exceeds the 3.10.B.2 limits but not (b)Iftheueasured(it the 3.10 B.1 lim , take one of the following actions:
- 1. Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> place the reactor in an equilibrium configuration for which Specif! cation 3.10.B.2 is satis-fled, or
- 2. Reduce reactor power and the high neutron flux tr!p s tpoint by 1% for each percent that the measured (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.
TS.3.10 9 mechanical properties to within assumed design. criteria, The.ECCS analysis was; performed in-accordance with SECY 83 472. One calculation.at the-95% ;;
< probability. level was' performed as well as one calculation with all tha
' required.~ features of 10 CFR Part 50, Appendix K. The 95% probability level calculation used.a_ peak linear heat. generation rate of 14.2 kv/ft, The Appendix K calculation used a peak linear heat generation rate of 15.8 kw/ft for the Fq limit of,2.5. ~ Maintaining 1) peaking factors.below the Fq
-limit of 2.5 during all Condition'I events and 2) the peak linear heat
_ generation rate below 14.2 kw/ft at the 95% probability level assures compliance with.the_ECCS analysis.
During ope ation,' the plant staff compares the measured hot channel factors, and F N , (described later) to the limits determined in.the transient ~
GA unalyses. The terms on.the right side of the equations in Section 3.10.B.1 represent the analytical limits Those terms on.the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.
N Fg is the measured Nuclear Hot Channel Factor, defined as the maximum local heat flux on the surface of a fuel rod divided by The average heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment.
The K(Z) function shown in Figure TS.3.10-5 is a normalized function that limits Fq axially. The K(Z) Specified for the' lowest six (6) feet
.of the core is arbitrarily flat since the lower part of the core is -l generally not limiting. Above that region, the K(Z) value is based on small break 1hCA analyses. I V(Z) is a axially dep ndent function applied to the equil!brium N
measured Fg to bound 's that could be measured at nonequilibrium
, conditions. This func ion is based on power distribution control
' analysis that evaluated the effect of burnable poisons, rod position, axial effects, and xenon Worth.
Ph, Engineering'HeatFlux'HotChannelFactor, is defined as t'ne allow-ance on heat flux required for manufacturing tolerances. The engi-neering factor allows for local variations in enrichment, pellet I
, density and d'.ameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
The 1.05 multiplier accounts for uncertainties associated with measure-mont of the power distribution with the movable incore detectors and j the use of those measurements to establish the assembly local power discribution.
d(uringtargetfluxdetermination.(equil)isthemeasuredlimitingFhobtainedatequ N
F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio ob;e,rintegraloflinearpoweralongtherodwiththehighestintegrated power to tho' average rod power.
Exhibit D V
Prairie Island Nuclear Generating Plant License Amendment Request Dated July 5,1988 Safety Evaluation of Increased FQ and FAH NSPNAD 8705 Revision 1 I
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