ML19332D181
| ML19332D181 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/17/1989 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML19332D177 | List: |
| References | |
| NUDOCS 8911300117 | |
| Download: ML19332D181 (59) | |
Text
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- -: Exhibit B ,
I Prairie Island'tiuclear Generating Plant
-License Amendment Request Dated November 17, 1989 Proposed Changes Marked Up 7
On Existing Technical Specification Pages r
l' Exhibit B consists of existing Technical Specification pages wit.h the proposed p changes written on those pages. Existing pages affected by this License Amendment Request are listed below:
I .-
L Eas,e TS ix fr TS xiii TS.1-2 P[ TS.3.1-12 TS.3.10 1 TS.3.10 2 TS.3.10 3 TS.3.10-4 TS.3.10-5 TS.3.10-7
!- TS.3.10-8 Figure TS.3.10-2 (delete)
Figure TS.3.10 3 (delete)
Figure TS.3.10 4 (delete)
Figure TS.3.10 5 (delete)
Figure TS.3.10 6 (delete)
Figure TS.3.10 7 (delete)
- j. TS 6.7 4 B.2.1 2 B.3.10-1 B.3.10 3 B.3.10-4 B.3.10-6 B.3.10-10 8911300117 891117 2 ADOCK 0500
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TABLE OF CONTENTS (Continued) i
- l. .
TS SECTION M PACE. ,
i 6.7 Reporting Requirements TS.6.7 1 .i A. Routine Reports TS.6.7-1 ,
- 1. Annual Report TS,6.7 '
'l l
- s. Occupational Exposure' Report- TS 6.7-1 i l- b. Report of Safety and Relief Valve Failures and Challenges TS.6.7-1 ,
- c. Primary Coolant Iodine Spike Report TS.6.7 1 ,
- 2. Startup Report. TS.6.7-2
- 3. Monthly' Operating Report TS.6.7-2 t
- 4. Semiannual Radioactive Effluent Release Report .
TS.6.7-3 l
- . 5. Annual Summaries of Mcteorological bata TS.6.7 4 B. Reportable Events TS.6.7 %
F
.C. Environmental Reports TS.6.7 K
- 1. Annual Radiation Environmental Monitoring 5 -[
Reports 75.6.7-% ,
- 2. Environmental Special Reports TS.6.7 5 t
- 3. Other Environmental Reports TS.6.7-5 (o ,
(non-radiological, non-aquatic)
D. Special Reports TS.6.7 1 i
- 6. Core Operating Limits Report TS.6.7-4
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TS-xiii 3 i
'l APPENDIX A TECHNICAL SPECIFICATIONS ,
1 LIST OF FIGURES a
1 1
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- TS FICURE II,ILE
, 2.1 1 Safety Limits, Reactor Core, Thermal.and Hydraulic Two Loop )
l 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 Limitatiens ~j
'3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit ;
Versus Percent of RATED THERMAL POWER with the Primar" Coolant l Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I 131 ,
3.9 1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid ,
Effluents ;
'3.9-2 Prairie Island' Nuclear Generating Plant Site Boundary for l Caseous Effluents
'3.10 1 Required Shutdown Margin Vs Reactor Boron Concentration !
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2.10 5 D: ficti:r fr:: T:rg:t Flun Diff:::::: :: : Fun:ti:n ;f Th::::1 Bewee-3,10 ' '?(2) :: : F ::ti:r :f Cer: M:ight 4
4.4 1 Shield Building Design In Leakage Rate 6.1-1 NSP Corporate Organizational Relationship to On Site Operating l Organizations 6.1 2 ~ Prairie Island Nuclear Generating Plant Functional Organization ,
for On Site Operating Group 1
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TS.1-2 ;
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i CONTAINMENT INTEGRITY CONTAINMENT INTECRITY shall exist when:
- 1. . Penetrations required to be isolated during accident conditions are either.
- a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or . ,
- b. Closed by manual valves, . blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specifications 3.6.C and 3.6.D.
- 2. Bilnd flanges required by Table TS.4.4-1 are installed. !
- 3. =The equipment hatch is closed and sealed. ,
4 . Each air lock is in compliance with the requirements of Specification t
3.6.M.
- 5. The containment leakage rates are vithin their required limits.
COLD SHUTDOWN f A' reactor is in the COLD SHUTDOWN condition when-the reactor is suberiti.
cal by at least 1% . k/k and the reactor coolant average temperature is less than,200'F.
-CORE ALTERATION CORE ALTERATION is the movement or manipulation of any component within ,
the reactor pressure vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity. Suspension of CORE ,
ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.
These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.7.A.6. Plant operation within these operating limits is addressed in individual specifications. ,
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s, 3 .1. F. . ISOTHERMAL TEMPFRATURE COEFFICIENT (TTC) ht 1. When the reactor is critical, the isothermal temperature coefficiert ,
shall be less .than 5 pem/*F with all rods withdrawn, except during j p low power PHYSICS TESTS and as specified in 3.1.F.2 and 3, 1 :
a
, 2. When the reactor is above 70 percent RATED THERMAL POWER with all ;
rods withdrawn, the isothermal temperature coefficient shall be !
P negative, except as specified in 3.1.F.3.
V . l f 3. luf the limits of 3.1.F.1 or 2 cannot be met, POWER OPERATION may l h continue provided, the following' actions are taken. ;
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- a. Establish and maintain control rod withdrawal limits sufficient l to restore the ITC to less than'the limits specified in i specification 3.1.F.1 and 2 above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT l !
SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.- These withdrawal limits '
shall be in addition to the insertion limits :f Figu :
T0.3.1^ 2.
l
- b. Maintain the control rods within the withdrawal limits l h established above until a subsequent calculation verifies that the ITC has been restored to within its limit for the all rods withdrawn condition. ,
I
- c. Submit a special report to the Commission within 30 days. - l !
describing the value of the measured ITC, the interim control '
rod withdrawal limits,t and the predicted average core burnup necessary for restoring the ITC to within its limit for the all- [
rods withdrawn condition. !
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specified in the CORE OPERATING LIMITS REPORT
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, 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS :
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Applies to the limits on core fission power distribution and to the limits f on control rod operations. ;
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To assure 1) core suberiticality after reactor trip, 2) acceptable core l
- power distributions during power operation, and'3) limited potential
,< reactivity insertions caused by hypothetical control rod. ejection, j
. t i Soecification j h A.- Shutdown Marrin-l' -
The shutdown margin with allowance for a stuck control rod assembly i shall exceed the applicable.value shown in Figure TS.3.10-1 under all i
' steady state operating conditions,- except for PHYSICS TESTS, from zero l l to full power, including effects of axial-power distribution. Th6 :
- shutdown margin as used here is defined as the amount by which the - +t
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reactor core would be suberitical'at HOT SHUTDOWN conditions if all -l 1 control rod assemblies were tripped, assuming that the highest. worth. '!
contro11 rod assembly remained' fully withdrawn, and assuming no changes _ -!'
in xenon or boron concentration.
B.=i-fower Distribution Limits
- 1. At all times, except d ring 1 w power PHYSICS TESTING, measured l hot channel factors, and g, as. defined below and in the ;
bases, shall meet the ollow ng limits:
a' normalized function that Ph x l'03 x 1.05-$(e-61/P)K(Z) ,
RTP w limits Fo (z) axially as ;
' kB specified in the Core !
(H x 1.04 $hM x (1+ +-6(1 P)] Operating Limits Report.
< [ vhere the following definitions apply:
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- K(Z) is th: 2xi:1 dq=d=:: En::in ;h:n ir Ti;;n: TS,2.10 5.
Z is the core height location.
- P is the fraction of RATED THERMAL POWER at which the core is l operating. In the Ph limit determination when P 50.50, see P - 0.50.
RTP <
- Fo is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.
RTP Fay is the Fag limit at RATED THERMAL POWER specified in the CORE OPERATING ,
LIMITS REPORT.
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- PFDH is the Power Factor Multiplier for F,g specHid in de QE EMUM ,
LIMITS REPORT. ~
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.3.10.B.1. - P7 orthe is defined as the measured Fq or F (H smallest !
wkth margin or greatest excess.3brespeccively, o 1'mit. i
-1.03isthgengineeringhotchannelfact:,r,Th,appliedtothe ,
measured qF" to account for manufacturing tolerance. ,
-1.05isappliedtothemeasured%toaccountformeasurement uncertainty.
- 1.04 is applied to the measured P[H to account for measurement uncertainty. , ,
- 2. Hot channel factors, F$ and FfH , shall be measured and the target j flux difference determined, at equilibrium conditions according '
to the following conditions, whichever occurs first:
(a) At least once per 31 effective full-power days in conjunction r with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the $
reactor power at which target flux difference was last f determined, by 10% or more of RATED THERMAll F0WER. l F$(equil)shallmeetthefollowinglimitforthemiddleaxial80% ';
of the core: rr 5 Q (equil) x V(Z) x 1.03 x 1.05<(ih M /P) x K(Z) .
where V(Z) is d: fin d Tign: 3.1^ 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 /
neuernn Nv trio set point by it for each percent that the !
m asured Py or by"+-4M for each percent that the measured H exceeds the 3.10.B.1 limit. Then follow 3.10 B.3(c).
(b) If the measured Ph (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions:
L 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 j l' configuration for which Specification 3.10.B.2 is satis- l fied, or p 2. Reduce reactor power and the high neutron flux trip L s tpoint by 14 for each percent that the measured l- (equil) x 1.03 x 1.05 x V(Z) exceeds the limit, j specified in the CORE OPERATINO LIMITS rep 0RT ;
1 the factor specified in the CORE OPERATING LIMITS REPORT l
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3.10.B.3 (c) If subsequent in core mapping cannot, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, demonstrate that the, hot channel factors are met, the reactor
. shall be brought to a HOT SHUTDOWN condition with return to power authorized up to 50% of RATED THERMAL POWER for the purpose of PHYSICS TESTING. Identify and correct the cause-of the out of limit condition prior to increasing THERMAL POWER above 50% of RATED ERMAL POWER. THERMAL POWER may then be increased provided or dH is demonstrated through in core
. mapping to be within its limits.
(d)'If two successivg measurements indicate an increase in the peak. rod power PSH with exposure, either of the following l actions shall be taken:
1.. % (equil) shall be multiplied by 1.02 x V(Z) x 1.03 x 1.05 for comparison to the limit specified in 3.10.B.2, or
- 2. (equil) shall be measured at least once per seven afectivefullpowerdaysuntiltw(osuccessivemapsH indicate that the peak pin power, is not increasing.
'4. Except during' PHYSICS TESTS, and except as provided by specifica. l-it tions-5 through 8 below, the indicated axial flux difference for
-- mm v at least three operable excore channels shall be maintained within hetyrgetband - ); 150 had about the target flux difference.
- 5. . Above 90 vercent of RATED THERMAL POWER; l If the indicated axial flux difference of two OPERABLE excore
_ _ channels deviates frompee. target band, within 15 minutes either l eliminate such deviation, or reduce THERMAL POWER to less than 90 percent of RATED THERMAL POWER. ,
. 6, . Beeveen 50 and 90 percent of RATED THERMAL POWER l
- a. The indicated axial flux difference may deviate from it; i!;-
target band for a maximum of one* hour (cumulative) in any 24-
' hour period provided that the difference between the indicated axial flux difference about the target flux difference does not exceed the envelope jMn. in rig ;;. TO . 2.10 5.
- b. If 6.a is violated for two OPERABLE excore channels then the THERMAL POWER shall be rec' ced to less than 50% of RATED THERMAL POWER and the high neutron flux setpoint reduced to less than 55% of RATED THERMAL POWER.
i
- May be extended to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> durin5 incore/excore calibration. _
The target band is specified in the CORE
^
.epecified in the CORE OPERATING LIMITS R -
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I i i 3.10.B.6. c. A power increase to a level greater than 90 percent of rated i
(= power is contingent upon the indicated axial flux difference ;
of at least three OPERABLE excore channels being within the j 7~
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target band. l
- 7. -Less than 50 vereent of RATED THERMAL POWER; g
- a. The indicated axial flux difference may deviate from 444r ,
target band. ;
- b. A power. increase to a level greater than 50 percent of RATED THERMAL POWER is contingent upon the indicated axial' flux
[ difference of at least three OPERABLE excore channels not being outside the target band for more than one hour (cumula- ;
j tive) out of the preceding 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, f
- 8. In applying 6a and 7b above, penalty deviations outside the'264, !
. target band shall be accumulated on a time basis of: ;
[. . a .- One minute penalty deviation for each one minute of power !
operation outside of the target band at THERMAL POWER levels ,
equal to or above 50% of RATED THERMAL POWER, and ;
- b. One half minute penalty deviation for each one minute of power operation outside of the target band at THERMAL POWER levels
~between 15% and 50% of RATED THERMAL' POWER. ,
O. -If alarms associated with monitoring the indicated axial flux _l difference deviations from the-444L target band are not operable, ,
the indicated axial flux difference value for each OPERABLE excore l channel shall be logged at least once per hour for the first 24 i hours and half hourly thereafter until the alarms are returned to ,
an OPERABLE status. For the purpose of applying this specifica- l tion, logged. values of indicated axial flux difference must be assumed to apply during the previous interval between loggings.
C. OUADRANT' POWER TILT RATIO .
- 1. Except for PHYSICS TESTS, if the QUADRANT POWER TILT RATIO exceeds l-1.02 but is less than 1.07, the rod position indication shall be l
monitored tnd logged once each shift to verify rod position within
~
l- each bank assignment and, within two hours, one of the following steps shall be taken:
I a. Correct the QUADRANT POWER TILT RATIO to less than 2.02. l
- b. Restrict core power level so as not to exceed RATED THERMAL POWER less 24 for every 0.01 that the QUADRANT POWER TILT RATIO exceeds 1.0.
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[ 3.10.C.2. If the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than 1.07 for a sustained period of more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or if such a tilt recurs intermittently, the reactor shall be brou5ht to the HOT SHUTDOWN condition. Subsequent operation below 50% of rating, for testing, shall be permitted,
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- 3. Except for PHYSICS TESTS if the QUADRANT POWER TILT RATIO exceeds 1.07, the reactor shall be brought to the HOT SHUTDOWN condition.
i subsequent operation below 50% of rating, for ' testing, shall be
);. : permitted.
L 4 If the core is operating above 854 power with one excore nuclear
[ channel inoperable,_ then the core quadrant power balance shall l-o be determined daily and after a 104 power change using either 2- I movable detectors or 4 core thermocoaples per quadrant, per Specification 3.11. , - - - - - --e C-L limited in physical insertion as specified' D. Rod Insertion Limits in the-CORE OPERATING LIMITS REPORT-
_ _ . - _v_
- 1. The shutdown rods shall be -fally ri:hdr.un when the reactor is p critical or approaching criticality.
- 2. When the reactor'is critical or approaching criticality, the l' control banks shall be limited in physical insertion; in::::i n ch::rn:1 :p::::ing ::nditi:n:.
trol bank insertion may be further restricted by specification 3.10.A , measured control rod worth of all rods, less
~-
the worth of the worst s s less than 5.52% reactivity at the beginning of the first cycle or the value if measured at any other time, or (2) if a rod is inoperable (Spec te- __
3.10 G).
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- /(. Insertion limits do not apply during PHYSICS TESTS or during periodic exercise of individual rods. The shutdown margin shown in Figure TS.3.10 1 must be maintained except for low power PHYSICS' TESTING. For this test the reactor may-be critical with l all but one high worth full-length control rod inserted for a L period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per year provided a rod drop test is L run on the high worth full-length rod prior to this particular low L
power PKYSICS TEST. l i
as specified in the CORE OPERATING LIMITS REPORT.
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F '3.10.C. Inocerable Rod Limitations a 1. An inoperable rod is a rod which (a) does not trip, (b) is
[ declared inoperable under specification 3.10.E. or 3.10.H. or ;
. (c) . cannot be moved by its drive mechanism and cannot be !
corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.. :
L . I
- 2. Th's reactor shall be brought _ to the HOT SHUTDOWN condition within l
- 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> should more than one inoperable rod be discovered during ;
POWER OPERATION.
- 3. If the inoperable rod is located below 'the 200 step level and is capable of being tripped, or if the rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits,fr. Tigu;; T0.3.1^ 3 apply.
- 4 If the inoperable rod cannot be located, or if the inoperable rod is located above the 30 step level and cannot be tripped, then the insertion limits J
......... ....... apply, 5.. If POWER.0PERATION is continued with one inoperable rod, the
. potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within.30 days unless the rod is earlier made OPERABLE. The analysis shall include due allowance for nonuniform fuel depletion 'l in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, THERMAL POWER shall be reduced to l a level consistent with the safety analysis.
H. Rod Dron Time At operating temperature and full flow, the drop time of each RCCA shall.be no greater than 1.8 seconds from loss of stationary gripper
.c - j. coil voltage to dashpot entry. If the time is greater than 1.8 seconds, the rod shall bs declared inoperable.
Y ~l specified in the CORE OPERATING LIMITS REPORT e
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t 3.10.I. . Monitor Jnocerability Reeuirements
- 1. = If the rod bank insertion limit monitor is inoperable. . or if the rod position deviation monitor is inoperable, individual rod !
positions shall be logged once-per shift, after a load change a greater than 10 percent of RATED THERMAL POWER, and afrar 30 ;
inches or more of rod motion. !
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.2. If both the rod position deviation monitor and one or both of the ;
quadrant power tilt monitors are inoperable for 2. hours or more, ;
the nuclear overpower trip shall be reset to 93% of RATED THERMAL i POWER in addition to the increased surveillance requirements.
- 3. . If one or both of the quadrant power tilt monitors is inoperable.
- individual upper and lower excore detector calibrated outputs and
, the calculated power tilt shall be logged every two hours after a load change greater than 10% of RATED THERMAL POWER 1
J. DNB Parameters i
The following DNB related parameters limits shall be maintained ,
during POWER OPERATION: , , _
}
the value specified in the i
- a. Reactor Coolant System Tavg $564'F CORE OPERATING LIMITS REPORT
- b. Pressurizer Pressure >2220 psia *' - -
- c. Reactor Coolant Flow E1?",000 ;;; 4 /
'With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER -
to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Compliance with a. and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
L Compliance with c. is demonstrated by verifying that the parameter is !
within its limit after cach refuelin5 cycle. ;
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- Limit not applicable during either a THERMAL P'0WER ramp increase in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step increase.in excess of (10%) RATED THERMAL POWER p
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Control Bank Insertion Limits 1 1
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- REV 77 4/3/86' I
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.n i REV 77 4/3/86 :
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- t. 1!
- l i: ' t v\S e.v h e C-No n b .7, @ , h 6.7. A.S. Annual Siimmaries of Meteorolocical Data l An annual summary of meteorological dats shall be submitted for the previous calendar year in the form of joint frequency distributions ;
of wind speed, wind direction, and atmospheric stability at the j request'of the Commission. j B. REPORTABLE EVENTS The following actions shall be taken for REPORTABLE EVENTS:
.. L
- a. . The Commission shall be notified by a report submitted pursuant i to the requirements of Section 50.73 to 10 CFR Part 50, and ]
- b. 1Each REPORTABLE EVENT shall be reviewed by the Operations l Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.
C. Environmental Reverts The reports listed below shall be submitted to the Administrator of the appropriate Regional NRC Office or his designate: ,
- 1. Annual Radiation Environmental Monitor'ac Revert (a)' Annual Radiation Environmental' Monitoring Reports covering the operation of the program during the previous calendar year shall be submitted prior to May 1 of each year.
(b) The Annual Radiation Environmental Monitoring Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental i surveillance activities for the report period, including a I comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The , f reports shall also include the results of land use censuses L required by Specification 4.10.B.1. If harmful effects or l' evidence of irreversible damage are detected by the l- monitoring, the report shall provide an analysis of the L problem and a planned course of action to alleviate the l problem. (c) The Annual Radiation Environmental Monitoring Reports shal) include summarized and tabulated results in the format r: Regulatory Guide 4.8, December 1975 of M1 radiole ,Na1 environmental samples taken during the report period. In the event that some results are not available for inclusion l with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possibic in a supplementary report. g
? ,
m h .n . MCW b e.(d*6eN b.l. k . b $ p .1 , l- 1
- * ~
- 6. Core Operating Limits Report bu
- a. Core operating limits shall be established and documented in-the CORE i OPERATING LIMITS REPORT before each reload cycle or any remaining .
i part of a reload cycle for the following: I l R1T
- 1. Heat. Flux Hot Channel Factor Limit (T o ), Nuclear Enthalpy l l RTP L' l Rise Hot Channel Factor Limit (F g
' (Specifications 3.10.B.1,3.1bn),PFDH,K(Z)andV(Z)
.B.2 and 3.10.B.3) i UW 2. Axial Flux Difference Limits and Target Band b
p (Specifications 3.10.B.4 through 3.10.B.9) ! b, .[ '
- 3. Shutdown and Control Bank Insertion Limits (Specification 3.10.D) ,
[ 4. Reactor Coolant System Flow Limit ti (Specification 3.10.J)
- b. The analytical methods used to determine the core operating limits I shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: '
} NSPNAD-8101-A, " Qualification of Reactor Physics Methods for
[ Application to PI Units" (latest approved version) NSPNAD-8102-A, " Prairie Island Nuclear Power. Plant Reload Safety k Evaluation Methods for Application to PI Units" (latest approved [ version) WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation 1 0' Methodology", July 1985 i I WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model , Using the NOTRUMP Code", August 1985 i-
, WCAp-10924-P-A, " Westinghouse Large-Break LOCA Best-Estimate j Methodology" December 1988 (
XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear - Power Distribution Control for Pressurized Water Reactors Phase ; II", May 1981
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core
)
j thermal-hydraulic limits ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be supplied upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. ;
x - l
-c i i
b: . y B.2.1-2 l E 2.1 SAFETY LIMIT. REACTOR CORE Bases continued .
. power levels. of 914 and 74% respectively. For the 2235 psig and 2385-psig curves, the coolant. average temperature at the core exit is equal L to 650*F below power levels of 644 and 73% respectively.
The third and fourth criteria are evaluated using standard DNB metho-t dology. For all four curves the DNBR is limiting at higher power levels. ! 4 The area of safe operation is below these curves.- The plant conditions required to vio'1ste the' limits in the lower power 'I range arc precluded by the self actuated safety valves on the steam r generators. The highest nominal setting of the steam generator safety . [ valves is 1129 psig (saturation temperature 560*F). At zero power the difference _ between primary coolant and secondary coolant is zero and at , full power it is 50*F. The reactor co'nditions at which steam generator l safety valves open is shown as a dashed line on Figure TS.2.1-1. Except for special tests, POWER OPERATION with only one loop or with l natural circulation is not allowed. Safety limits for such special tests will be determined as a part of the test procedure, i PFC>td
- The curves are conservative for the following nuclear hot channel factors.
WTP 3 k RTP \ AHj N H =}-r4 [1 + 454(1-P)) ; and Fh' = 4 r64-Use of these factors results in more conservative safety limits than , would result from power distribution limits in Specification TS.3.10. This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to ' maximum allowable control rod insertion. .The control rod insertion limits are covered by-Specification 3.10. Adverse power distribution , factors could occur at lower power levels because additional control i rods are in the core. However, the control rod insertion limits specified ' "'--- *" ' ' ^ ' assure that the DNB ratio is always _ greater at part power than at' full power ' theCOREOPERdNGL[IMITREPORT 7 The Reactor Control and Protective System is designed to prerent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel, i whore: RTP
- Fo is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.
RTP
-Fgais the Fan Iimit at RATED THERMAL POWER specified in the CORE '
OPERATING LIMITS REPORT. N
- PFDh is the Power Factor Multiplier for F 3e srscified in the CORE OPERATING LIMITS REPORT.
u W-r
--- -- . . - e-- -
- g. ;
4 {, y B.3.10-1 ,
,3.10 CONTROL ROD AND p0VER DISTRIBUTION LIMITS Eases .
r Throughout the 3.10 Technical Specifications, the terms " rod (s)" and g t
"RCCA(s)" are synonymous.
A. Shutdown Margin [ '[ Trip shutdown reactivity is provided consistent with plant safety , analyses assumptions. One percent sbutdown margin is adequate except l l for the steam break analysis, which requires more shut & wn reactivity , due to the more negative moderator temperature coefficient at end of life (when boron concentration is low). Figure TS.3.10-1 is drawn accordingly, , B. Power Distribution Control l The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR in the. core of greater than or equal to 1.30 for Exxon fuel and 1.17 for Westinghouse l fuel during normal operation e.nd in short term transients, and (b) ! l limiting the fission gas release, fuel pellet temperature and cladding > l l' 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 the required features of 10; CFR Part 50, Appendix K. The 95% Mg ,
. probability level calculation used vpeak linear heat generation rate ,
vf l'. 2 h/f:. The Appendix K calculation used # peak linear heat ' generation rath;f 15.0 h/ft for the Fq limit ef 2. 7 Maintaining 1) l peaking factors below the Fq limit :f 0.5 during all Condition I events and'2) the peak linear heat generation rate below,aj. S/f:Te the 95% J. I- - - - - probability level assures compliance with the ECCS analysis.'@kOd u.D - th 1 During factors,ope and ation,f F H ,e(described plant staff later) compares the measured to the-limits hot channel determined in the l transient and LOCA analyses. The terms on the right side of the l 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. Py is the measured Neelear 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.
~TheK(Z) functio):hm in ri;;;; TC.2.10 5 is a normalized function that limits Fq axially. S.: 5(2) :p::ifid f:: "h: 1::::: :in (5) f:::
f :he ;;;; i: ::bi::: ily f1:: ;in:: :h 1:r:: p::: :f :h: :::: i:-
;;n:::117 :: li;iting. */.t;;; th:: ::gi;n, khe K(Z) value is based on j '
large and small break LOCA analyses. f- W_ specified in the CORE OPERATING LIMITS REPORT l
n - - .
, , ~
y'.g;,y$ Q ys as ' L\ '
,f .B.3.10-3 0:s',- "' i
- 3.10 lDONTROL ROD AND ' POWER DISTRIBUTION LIMITS '
g , . ilyttg. continued- i pRh*,.ameasurementofF[H is - taken, measurem v.c iarror must be allowed
~ "
;for,and 4 percent is the appropriate allowance for a-full core. map ',
~ taken with:the-movable incore detector flux:mspping' system.-
i
...asurements of the hot channel. factors are. required as part of startup '
c PHYSICS TESTS, at least'once each effectiva full power month of operation,
> and whenever abnorms1 power distribution conditions require a reduction of
-[ 'i core power. co- a level based on measured hot channel factors, The incore
- map taken following initial loading provides confirmation of the basic . _
7 uuclear: design bases including proper fuel loading patterns. The periodic' , monthly incore mapping provides additional assurance that the nuclear. design, bases remain inviolate and identify. operational anomalies which would
' otherwise affect these bases. '
For-normal-operation, it is not necessary to me.isure these quantities.- Instead it has been determined that, provided curtain. conditions are
. observed, the hot channel factor limits will be. met; these conditions are .
as follows:
. J
- 1. Control rods in a single bank move together with .no l
. v,
~
individual rod insertion differing by more _ than 15 . inches from the bank demand position. An accidental misalignment li .t of 13 steps precludes a rod misalign- - ment gret;sr t; a 15 inches with consideration of maximum
'instrum atatiot. et 'r.
- 2. control rod banks are sequenced with overlapping banks as. described in Technical Specification 3.~10. '
3 .- 'Thecontrolbankinsertionlimitgarenotviolated. , s . 4 4 Axial power distribution control procedures, which are given in , f terms of flux difference control and control bank insertion t limits are observed. Flux difference refere to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial' offset which is defined as the difference in
, normalized power between the top and bottom halves of the core.
spe:ified in the CORE OPERN1ING LIMITS REPORT 4
w q
%e 4 B.3.10 4 >
I 3.l0' CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases = continued- l Bj> Power Distribution Control'(continued) s ThepermittedrelaxationinFfgand allows for radial power shape
~
Lchangeswithrodinsertsontothein(ertionlimits. s It has been b determined that provided the above conditions 1 through 4 are obse ed, e these hot channel factor limits are met. In specification 3.10, is arbitrarily' limited for P'less than or equal to 0.5 (except for lo power PHYSICS TESTS). ! > t o > The procedures for axial power distribution control referred to above are designed to minimize the effects of ; tenon redistribution on tbr axial power distribution during load-follow maneuvers. Basically. control'of flux difference is required-to limit the difference between
'the current value of Flux Difference (AI) and a reference value which !
corresponds co the full power equilibrium value of Axial Offset (Axial Offset - AI/fracticnal power). .The reference value of flux difference varies with power'1evel and burnup but expressed as axial offset it U . varies-only with burnup.
.The technical; specifications on power distribution control assure that theLF 11mit.is not exceeded and xenon distributions are not developed which4at a later. time, would cause greater local power peaking even'though ,
- the: flux difference is then within the limits specified by the1 procedure.
+ The' target-(or reference) value of flux difference is determined as f follows:/ At any time that equilibrium xenon conditions have been established, the indicated flux difference is 'noted with the full length rod control rod bank more than 190 steps withdrawn (i.e. , normal
= full power. operating position appropriate for the time in life, usually
" withdrawn farther as burnup proceeds). This value, divided by the fraction -of full poaer at which the core was operating is tho. full -
. power value Lof the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional' power. Since the indicated equilibrium was noted, no u allowances for excore detector-error are necessary and indicated > deviation,;f i5 p::::r.: cr; p;;;itt;d from the indicated reference valug. '
!Sg.;; TC . 2.10 5 ;h;r \healloweddeviationfromthetarget flux' difference as -the function of THERMAL POWE l
. - - - -'_ _ _~
~
'but within the target is specified in the CORE is permitted OPERATING LI]!ITS REPOR h
e 1 l-
>< ll b
qa>; ; 1.
~
ff j spcciffsd in th3 CORE OPERATING LIMITS REPORT B.3.10 6-a 3'10 CONTROL ROD AND POWER DISTRIBUTION LIMITS e Bases continued' B '. Power Distribution Control (continued) c In some instances of rapid plant power. reduction, automatic rod lr motion will cause the: flux difference to deviate from.the target band'when the reduced power level is reached. This does not necessarily affect the . xenon distribution: sufficiently sto change the envelope of: peaking 1 factors whichican be reached on a subse-quent . return .to full power within the target band, however to simplify the ' specification, t limitation 'of one hour in any period of 24 hours is placed on operation outside the band. =This ensures i 37 that the- resulting xenon. distributions are not significantly different -from those' resulting' from operation within the target band. ? The consequences of being outside the $&T target band but within the
, "ieur: T0.2.10 5 limit %for. power levels between 50% and 904 has been evaluated and determined to result in acceptable peaking factors. There-fore, while the- deviation exists the power level is limited to 90 percent or lower depending on the indicated axial flux difference, In all cases the 15 ;;;;;;t ' target band is the Limiting Condition for Operation.- Only when the target band is violated do the limits wwwker "ig.:e T0.2.10-0 apply.
"If, for any reason, the indicated axial flux difference is not control-led within the t5 p:::;nt band for as long a period as one hour, then l xenon distributions may be- significantly changed and operation at- or l below 50 percent is required to protect against potentially more severe L consequences of some~ accidents.
SAs discussed above,.the essence of the procedure is to maintain the xenon distribution .in the core as close to the equilibrium full power condition as possible. This is accomplished by using the' boron system to position the full . length control rods to produce the required
; indicated flux difference.
1For Condition II events the core is protected from overpower and a )
-minimum DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by l
- an automatic protection system. Compliance with operating procedures .;
we is assumed'as a precondition for Condition II transients, however, ! I' operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered. I 1 C. QUADRANT POWER TILT RATIO , 1 O' QUADRANT POWER TILT RATIO limits are based on the following considera- ; i tions. Frequent power tilts are not anticipated during normal operation since this phenomenon is caused by some asymmetric perturbation, e.g. I rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F, and l. l' l t i
y 3 -
- p. <
,+g t
.c l t'
( . B.3.10 10 f,, P P . . . A i3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS N Bases continued-y - D H.. Rod Drop Time The. required drop time to dashpot entry is consistent with the safety , analysis . - i
's '
.I..LMonitor Inoperability Requirements l' If either the rod bank insertion limit monitor or rod position devia-
, tion monitor are inoperable, additional surveillance is required to h .. ensure adequate shutdown margin is maintained.
If the rod position deviation monitor and quadrant power tilt monitor (s) are. inoperable, the overpower reactor trip setpoint is reduced (and also J power) to ensure that adequate core protection is provided in the event
~
that unsatisfactory conditions arise that could affect radial power
, ' t distribution.
E Increased surveillance is required, if the quadrant power tilt monitors are> inoperable and a load change occurs, in order to confirm satisfac-tory power' distribution behavior. The automatic alarm functions related to QUADRANT POWER TILT must be considered incapable of alerting the l ,~ operator: to unsatisfactory power distribution conditions. J. DNB Parameters and Pressurizer Pressure requirements are based -
;TheRCS-flowrate,'T,y$s,sumptions.
.on transient analyses The flow rato shall be- verified by .
calorimetric flow. data and/or elbow. taps. Elbow taps are used in the . reactor coolant system as an instrument device that indicates the status of the_ reactor coolant flow. The basic function of this device is to provide information as to whether or not a reduction in flow rate has occurred. If a reduction in flow rate-is indicated below the :p::ifirs-, v
.t4en.value 'ndi:;r.d, shutdown is required to investigate adequacy of core cooling during operation.
i specified in the CORE OPERATING LD1ITS REPORT
![i o :
h' jl ! - ' 6;(rt :. j h
, r !: Exhibit C t n; Prdirle Island Nuclear Generating Plant i V
License' Amendment Request Dated November 17, 1989-
~
og REVISED' TECHNICAL SPECIFICATIONLPAGES 1 + . Exhibit C consists of revised pages for_the Prairie Island Nuclear Generating Plant Technical: Specifications with_the proposed changes incorporated as B" listed below: Enga d' L-TS ix TS-xiii
. TS.1-2 1 TS.3.1-12 TS.3,10-1 TS.3.10-2 TS.3.10 3 TS.3.10-4
- s. TS.3.10 5 TS.3.10-7 TS.3.10-8 TS.6.7-4 TS.6.7-5 TS.6.7 B.2.1-2 B.3.10 1 B.3.10-3 B.3;10-4 B.3.10-6 'l B.3.10-10 l1 i
s i
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i ry, V& rs. ' s ,1 s V . '* j r us m c
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Eu - A F( $.Ifj 6
, ] $. I l TS-ix ,
l, ' li' d . i Gr: v; 7 TABLE OF CONTENTS (Continued) . I
*"p~
F> : . Oc :-TS SECTION- IIILE PAGE ' hi- 1 6" -- 6 . 7 . Reporting Requirenents TS.6.7 1 , l(g"a K A'. Routine Reports ~ s TS.6.7 19 I Qkn i 1. Annual. Report .TS.6.7;1; -i fn' i .
- a. Occupational Exposure Report TS 6,7-1, l sc ?[ ^ b. Report _of Safety and Relief Valve .
t
- a 1 Failures and Challenges .
TS.6.7 l' - t 3 m; c. Primary Coolant Iodine Spike _ Report - TS 6.7-1
- y@ ,
L2. Startup Report;, TS.6.7-2
'pe
- 3. Monthly Operating Report TS.6.7 2 fjff%' 4. Semiannual Radioactive Effluent Release
. .c ' Report TS.6.7-3 )
I ~5.' Annual Summaries of Meteorological Data- TS.6 7-4 .!
.6. Core Operating Limits Report TS.6.7-4: .,
9 - B. Reportable: Events TS.6.7-5~ '; osa ,
.C. Environmental Reports TS 6'.7-5 .
- 1. Annual Radiation Environmental Monitoring Reports. .TS.6.7-5 ,
'2. Environmental Special Reports TS.6.7-6 ggg 3.~Other Environmental Reports TS.6.7 6 h
(non radiological, non-aquatic) ?
,D.(Special. Reports ' TS 6.7-6 s a
. j.
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0 ' [ 50 ,i TS xiii ( L 4 APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES l
.TS FIGURE. '
TITLE 11 , .
~n and Hydraulic Two Loop
- 2;1 , - Safety Limits, Reactor Core, it Operation '
. - 4 '
t J: 3.1-1. Unit 1 and Unit 2 Reactor' Coolant System Heatup Limitations i 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. ,l:
=3.9-1L PrairieL Island Nuclear Generatir.g Plant Site Boundary for Liquid J Effluents
- p ,
3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for- 'I
' Gaseous < Effluents 3.10-l' Required-Shutdown Margin Vs Reactor Boron Concentration .t 4.4-1 Shield Building Design In Leakage Rate [
- 6.1 -l' - NSP-Corporate Organizational Relationship to On Site Operating ,
; 0rganizations
-6,1-2 Prairie Island Nuclear. Generating Plant Functional Organizationi E!
7 .for On-Site Operating Group - e
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e
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f o - C t a t b YFYs - "
m 3- - l ;; y c 3
& TS.1-2 .f; e
s ?
,, 3 CONTAINMENT INTEGRITY: !
CONTAINMENT INTEGRIW shall exist when: , 1 ' Penetrations required to be. isolated during accident conditions are either:' -
- a. Capable of being closed by an OPERABLE containment automatic! j
' isolation valve system, or b, -Closed by manual valves, blind flanges, or deactivated automatic f
" f valves secured in-their closed positions,.except as provided in
.'y-Specifications 3.6 C and 3.6~.D.
l' 2. Blind! flanges required by Table TS.4.4 1 are installed, s
- 3. 'The. equipment hatch is closed and sealed, j
t
-4.- Each air lock is in compliance with the requirescuts of Specification l3.6 M.
~
- 5. The containment leakage. rates are within their required limits, q COLD SHUTDOWN- i A reactor'is in the' COLD SHUTDOWN condition when the reactor is subcriti-k ~ cal..by' atx least 1%Ak/k and=the reactor coolant average temperature is fless than' 200*F... .;
-t
' CORE ALTERATION- ,
CORE ALTERATION'is the movement or manipulation of any component. Nichin the' reactor pressure vessel with the vessel head removed and fuel in j
> the vessel, which may affect. core reactivity. Suspension of CORE ,
L -ALTERATION shall not preclude completion of movement of-a component to ; a safe conservative position. -: CORE OPERATING LIMITS REPORT
'The CORE.0PERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. i H , 'These cycle-specific core operating limits shall be determined for each l< . reload cycle in accordance with Specification 6.7 A 6. Plant operation
,a lwithin these operating limits is addressed in individual specifications. u
t
$3n @ ,
E hb ;.1 . i Ll v:s. lL'5 TS.3.1 12' l 2 ' i 14 (:
. 3.1.FJ ISOTHERMAL TEMPERATURE COEFFICIENT (ITC) 'I @$ ~1. When the reactor is= critical, the isothermal temperature coefficient shall be less than 5 pcm/*F with all rods. withdrawn,.except duringi low power PHYSICS TESTS and as specified in 3.1.F.2 and 3. >
ic, , I
;p. -2. When the reactor is above 70 percent, RATED THERMAL POWER with all i rods: withdrawn, the' isothermal temperature' coefficient shall;be , ?( , negative', except'as'specified in 3.1.F.3~.
s 3. IfJthe limits of?3.1.F.1 or.2 cannot be met, POWER OPERATION may NcvTJ continue ~provided-the following actions'are taken:
?
~'a.: Establish and maintain control rod withdrawal ~ limits sufficient '
. toJrestore'the lic to less than the.. limits specified in .
ESpecification 3.1.F.1 a.ed 2 above. within 24 hours ~ or be in' HOT ~! l; SHUTDOWN within'the next 6 hours. These withdrawal-limits 1 l , shall be-in addition to :the insertion'11mits specified in then ,
" t
.00RE OPERATING LIMITS REPORT.
n b. Maintain the control rods within the withdrawal limits: L established.above until a subsequent calculation verifies that thez ITC ha's been restored-to within its limit for the all rods = 5 ,w ' withdrawn condition. , , g. ' 4 i c. Submit a special report toLthe. commission within 30 days,-
, ' describing the value of the measured ITC, thefinterim < control 0
rod withdrawa1' limits, and the-predicted average ccre burnup-- > l necessary for restoring the ITC to within its limit for the'all crods-withdrawn condition, j
~
-i
.- g ,
q [- .. P TS.3.10 1. E 3.10' CONTROL' ROD AND POWER' DISTRIBUTION LIMITS
- Aeolicabilitv:
r
- ( .
r . Applies to the limits on core. fission power distribution and to the limits l
. on contt 11 rod operations.
,g ,
p , .. : Obiective
'i g ,,
To assure;1).cofeEsuberiticality'after reactor trip, 2).~ acceptable core < q power distributions during power operat! ion, and 3). limited potential '
,o reactivity -insertions caused by hypothetical control rod ejection.
-Soecification
[ A'. > Shutdown Margin- I f: ' i The shutdown margin with allowance for a stuck control rod assembly - :I shall exceed the applict.ble value shown in Figure TS.3.10-1 under all- 3 steady state operating conditions,- except for PHYSICS TESTS. 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- :
- L reactor core <would be subcritical'at HOT SHUTDOWN conditions <if all control: rod assemblies were tripped, assuming that the highest worth.
controltrod assembly remained fully withdrawn, and assuming no changes' r in xenon or boron concentration.
,B. Power Distribution Limits ,
.1.1 At:all times, except d ring low power PHYSICS TESTING, measured o ' hot channel factors, andFfH,asdefined.belowandinthe
, bases, ah-11 meet the ollowing limits:
% ' x 1.03 x 1.05 '$ (Fq P) x K(Z)
RTP F$Hx1.04.$F L AH x [1+ PFDH(1-P)) where.the following definitions apply: l' RTP ;
- i. ,Fo is.the F limit O at RATED THERMAL POWER specified in.the CORE
^
OPERATING LIMITS REPORT. L RTP
-F AH is the FAH limit at RATED THERMAL' POWER specified in the CORE OPERATING LIMITS REPORT.
-PFDHisthePowerFactorMultiplierforFfHspecifiedintheCORE OPERATING LIMITS REPORT.
- K(Z) is a normalized function that limits F0(z) axially as
[' specified in the CORE OPERATING LIMITS REPORT. 1, L - Z is the core height location. L 1 l: - P is the fraction of RATED THERMAL POWER at which the core is operating. In the limit determination when P $0.50, set j' P - 0.50.
, c A< ,,
3
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Qgllh f ,
- c
'l %o.
eau
.s .
Syf, s ~
.TS.3.10 2 R .
f I 3.10'B.l'.' .--FNorPfH is defined as the measured Fq or F3 respectively,. wkththe'smallestmarginorgreatestexcess-oh'~1imit. n
.s g, ,
E l.03 is'the engineering hot channel factor, F ,' applied to the
. measured %toaccountformanufacturingtolehance. >
k p g.: ,1,05.isappliedtothemeasured(toaccountformeasurement uncertainty. , [' fr. 1.04'isappliedtothemeasured-(H to account for measurement 5 F ,
~ uncertainty.
N
; 2. Hot. channel ^ factors, o F and(H'.shallbemeasuredandthetarget' pl flux difference determined,.at equilibrium conditions accordingi '
c,h ,g -to.the following conditions, whichever. occurs first: (a) At least once per 31 effective full power ~ days in conjunction s
- with the . target flux' difference determination, - or .
m. (b)-Upon. reaching' equilibrium conditions after exceeding the' reactor l power at which target flux difference was-last- '
- determined, by 10% or more of RATED THERMAL POWER. -
F$L(equil)shallmeetthefollowinglimitforthemiddleaxial-'80% l of.the. core:- RTP k(equil)xV(Z)x1.03x1.05$(,q/P)x.K(Z) r where V(Z):is specified in the CORE OPERATING LIMITS REPORT.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 reacter power and the high' j 1: neutron.f x trip set-point by 1% for each percent that the [
g" measured or by the factor specified'in,the CORE j OJERATING IMITS REPORT for each percent-that the measured
? qH exceeds the 3.10.B.1 limit. Then follow 3.10.B.3(c).
o-(b) If;the measured (equil) exceeds the 3.10.B.2 limits but not the'3.10.B.1 limi , take one of the following actions:
- 1. Within 48 hours place the reactor in an equilibrium
, , configuration for which Specification 3.10.B.2 is satis-fied, or
- 2. Reduce reactor power and the high neutron flux trip !
s tpoint by 1% for each percent that the measured (equil) x 1.03 x 1.05 x V(Z) exceeds the limit. 9 0 .- - __________._m
+
e
~
ph; ' '
+
l png y- q f{! lq' n ;- t W ' TS.3.10 3 t l :1 E i .
't M ' '
1 j3 10 B~3. . (c)lIf subsequent
. t 'in ccre mapping = cannot, within ai 24_ hour period,:
V ' demonstrate that the hot; channel' factors:are met,_the reactor' ,
- g. y '
- shall'be brought to a HOT SHUTDOWM condition with return to V" . power' authorized up to 50% of RATED THERMAL POWER for:the i
_ '_ . -purpose,of PHYSICS = TESTING. Identify and correct the cause:.< ' of the out of limit condition prior to increasing THERMAL POWER; above 50% of RATED ERMAL POWER.- THERMAL POWER may then be: , 9'w' ' increased provided or FU ~'isidemonstrated through in, core '
,{
f mappingtobewithinits1 Nits. j sj (d)J If: two successiva r.easurements indicate- an increase in the- i peak rod power Fh with exposure, either ofithe'following 4 actionsshallbeWaken: q
- v
' 1.: FN-(equil)shallbemultipliedby1.02xV(Z)x1.03lx
- q. '
L1,05 for comparison to the limit specified in 3.10.B 2,' or i 2. ~ FN.(equil)~shall.be measured at least once_'per,seven, _[ l' ekfectivefullpowerdaysuntiltwosuccessive-maps ) indicatethatthepeakpinpower,F$g,isnotincreasing. .[
- 4. Except during PHYSICS TESTS, and except.as provided by specifica<
;tions 5~through 8 below, _ the indicated axialiflux-difference for e i: atsleast three operable excore channels shall. be raintained within.
the. target band about. the target flux difference. -The target band
'is specified in:the CORE OPERATING LIMITS REPORT.
- 5. .Above 90 nercent of-RATED' THERMAL POWER:
- If the indicated axial flux difference of two OPERABLE'excore
, . channels deviates from the target band, within 15 minutes either l ,
4 L eliminate such deviation, or reduce THERMAL POWER.to less than 90. H parcent of RATED' THERMAL POWER. I A '6. 'Between 50 and 90 cercent of RATED THERMAL POWERI.
- a. The indicated axial flux difference may_ deviate from the l I
target band for a maximum of one* hour,(cumulative) in any 24- 1 hour period provided that the difference between the-indicated- - axial flux' difference about the target flux difference does i not exceed the envelope specified in the CORE OPERATING LIMITS REPORT. , h 3
- b. If 6.a is violated for two OPERABLE excore channels then the THERMAL POWER shall be reduced to less than 50% of RATED THERMAL POWER e.nd the high neutron flux setpoint reduced to less than 55% of RATED THERMAL F0WFR.
*May be extended to 16 hours during incore/excore calibration.
4 <w a =w- e
- q e-m.qA9 . r -
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~
S 3ll ~ , TS.3.10 4 , c V,, I, , > A power.. increase. 'to a level ' greater than ' 90 percent of rated 3.10.B.6. . c. : y!
'd cpower-is' contingent upon the indicated axial flux difference-i hp ,
[of'at least three OPERABLE excore channels being within the ' target band.
- 7. ELess than 50 nercent of RATED THERMAL POWER:
4,, .c a.. TheLindicated axial flux difference may deviate.from'the-
~
f
' -l b', : target band, ,
e a p
" y' ' ,'
- b. A power: increase to a level ~ greater than 50 percent of RATED THERMAL: POWER is contingent upon the-indicated axial flux difference of at least three OPERABLE excore channels not JC I
being outside the-target band for more than one hour'(cumula- ;[ Etive) out of the-preceding 24 hour period._
~
r ' }, , 8. 'In applying 6a and 7b above,' pen'alty deviations outside the ll , l target band shall be accumulated.on a time basis of:
; a. 2 0ne minute' penal:y deviation for each'one' minute of power y- operation outside of the target band at THERMAL POWER levels y.
1; equal to or above 50% of RATED THERMAL: POWER, and-l- V .
. b. - One-half minute penalty deviation. for each one' minute of power p . operation outside of the target band at THERMAL: POWER levels ;
between.15% and 50% oft RATED THERMAL POWER. 9 , i 9 .- If alarms associated with monitoring the indicated axial flux l' difference deviations from the target band are not operable, the- l indicated exial flux difference value for each OPERABLE excore~ , , channel'shall.be logged at least once per hour for the first 24-I hours and half-hourly thereafter until' the alarms are returned to l sy an OPERABLE status. For the! purpose of' applying this specifica-tion, logged values of indicated axial flui difference must be (,
-assumed to apply during the previous 7i nterval:between loggings. '
C. OUADRANT POWER TILT RATIO
- 1. Except for PHYSICS TESTS,-if the QUADRANT POWER TILT RATIO-exceeds.
1.02 but is less than 1.07, the rod position indication shall'be monitored and logged once each shift to verify rod position within each bank assignment and, within two hours, one of the following o' steps shall be taken: D '
- a. Correct the QUADRANT POWER TILT RATIO to less than 1.02.
, b. Restrict core power level so as not to exceed RATED THERMAL POWER less 2% for every 0.01 that the QUADRANT POWER TILT RATIO exceeds 1.0.
e, : h' b' ah , s , . . - , . . . , . . . , , , , ,-
~
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, , El V
H-l .. r b ch ! NNet , R;&ou$kb,
,y'~
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\
i TS.3.10-5 l
, , , ,s t y ' '
0, , 4 s 3.1'0','C;2..'If"the. QUADRANT. POWER-TILT RATIO exceeds liO2 but is less thanL ;
,, 11.07 for a sustained period of more than 24 hours, or if such mL l tilt recurs intermittently, the reactor shall be brought to the: l s
~ HOT. SHUTDOWN condition. Subsequent operation below 50% of=
rating, for-testing,'shall'be permitted,
- : t.-
j pO
'3. .Except;for PHYSICS TESTS if-the QUADRANT POWER TILT RATIO exceeds l s
f, 1.07, the reactor.shall be brought to the HOT SHUTDOWN condition.1 l [
~
., Subsequent operation below 50% of rating,- for testing, shall be d "3
. permitted.
- 4. If the core is operating above 85%-power.with one excore nuclear gfi channel inoperable then the core quadrant' power balance shall [
be-determined daily and-after a 100-power change using either 2
.}
movable-detectors or 4 core thermocouples'per quadrant,,per l m , Specification'3.11. 4 N: D. Rod Inrertion Limits f cc Wl 1. The shutdown rods shall be limited in physical insertion as ? specified in the CORE.0PERATING, LIMITS REPORT ~when.the reactor is
~
criticalior approaching criticality.
' t
.2. When the reactor'is critical or approaching criticalihy,-ths' _
._ t 1
co'n trol banks shall be limitedlin physical insertion as specified 1 in the CORE OPERATING LIMITS REPORT.
'jn ,
,,' ~
- 3. Insertion limits do not apply during PHYSICS TESTS'or durinF
~ periodic exercise of individua1Lrods. The shutdown margin shown in Figure TS.3.10 1 must be maintained except;for low' power PHYSICS: TESTING. For thisttest the reactor may be critical with 1 all'but one high worth full-length control rod inserted for a.
period not to exceed 2 hours per year provided a rod drop test is-g , run on the high worth full-length rod prior to this particular low 1 ' power PHYSICS TEST. t u 1
- r. ..
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[' ' dt 3.10,G. Inocerable Rod limitations f m -
. 1. An inoperable rod-is a rod which (a) does not trip, (b)11s '
declared inoperable under specification 3.10 E. or-3.10.H. or~ ', (c) cannot be moved by its drive mechanism and cannot be ,; corrected within 8 hours.- (
.2. The reactor shal1~ be. brought- to the HOT SHUTDOWN condition within 6- hours: should more than one inoperable rod be discovered during -
p 1 POWER OPERATION.' M u - 3. If the inoperable-rod is located-below the~200 step level and is -
. capable. of being tripped. or if the rod. is Elecated below the 30. ;
step level whetherlor not it is capable of being tripped,' then the:
~
-insertion' limits specified in the CORE-OPERATING LIMITS. REPORT- lh 1 apply. ,
fj ^
- 4. cIf the inoperable rod cannot be located, or'if the inoperable rod )
is located above the.30. step level and cannot.be tripped;!then' , the insertion limits specified in the CORE OPERATING LIMITS REPORT l. apply. L - 5. If POWER OPERATION-is continued with one inoperable, rod,;the i [', ~ .
-potential- ~ ejected rod worth and associated transient power distribution'poaking, factors.shall-be' determined byianalysis~
within 30; days unless the. rod is earlier lmade OPERABLE. . The , analysis shallrinclude due allowance'for nonuniform fuel: depletion >s A L in the. neighborhood'of.the= inoperable l rod. If'the' analysis resultsf ni aimore -limiting hypothetical transienti than the' cases . , pi : reported in the safetyLanalysis, THERMAL POWER shall be reduced to- n l !a level consistent with' the . safety analysis. ~
.H. Rod Dron Time-3 3
At operating ~ temperature and full-flow,.the drop time of each RCCA' r shall bL no greater than 1.8 seconds from loss of stationary. gripper'
~
L coil voltage to dashpot entry. If.the time ~1s greater than 1.8 seconds, the rod shall be declared inoperable. ; 1.
=
5
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TS.3-10 8 h; / I : 3.10.I. Monitor Inocerability Reauirements 4 ; e . .
, 1. - If-the' cod bank insertion ~1imit monitor is inoperabic, or if the -
(' ' rod position' deviation monitor is inoperable, individual-rod.
, positions shall be logged once per shift, after a load change i be' , greater than 10 percent of RATED _ THERMAL POWER, and after 30 !
W '
-inches or more of rod motion.<
- f p;n
- 77. 3 2. 'If both the rod position deviation uonitor and one or both of- the KEb quadrant power' tilt monitors are inoperable for 2 hours or inore .
t? " the nuclear overpower trip shall be reset to 93% of: RATED: THERMAL ,
"O( POWER in addition to the increased-surveillance requirements, l pny
- 3. -If one or,both of the quadrant power tilt monitors is' inoperable, (6 / W;' ' individual upper and lower excore detector calibrated outputs and the calculated power tilt she,11 be logged every twq hours:after a-li z load change' greater than.10% of RATED THERMAL POUER- , ;
N 2E J..DNB-Parameters (
;dP The following DNB related parameters limits shall be maintained durin5 POWER OPERATION: ;
i
- a. Reactor Coolant System Tavg <564*F l
'b. : Pressurizer Pressure 22220 psia
- 1 in c'- Reactor Coolant Flow 2the value specified in.the CORE f OPERATING LIMITS REPORT: 4
.c
'V "With any of the above parameters exceeding its-limit.: restore the- ?
, , ' parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
I
- Compliance with a. . and b, is demonstrated b'y verifying'~ that each of.
the_ parameters is within its limits at least once'each 12 hours. c 9 . Compliance with c. is demonstrated by verifying that the parameter is
! -within its limit after each refueling cycle.
o' '
.ij r
n- ,
;r. '
r s
.*Limit not applicable during either a THERMAL POWER ramp increase l in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER l fstep increase in excess of-(10%) RATED THERMAL POWER 4
EP. I o 4 ______-_-_-__-___--__________l
m - . a v...- ' '
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.
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' "] ' TS.6.7 4 99l > +
' 6.7.A.S.> Annual Som=mries of Meteorological Data
( 93 ! vb ,~ _
-+ fan-annualesuarary of meteorological data shall be submitted for thel - .
- y Eprevious _ calendar year in,the form of joint frequency distributions q; of wind speed, wind direction,' and atmospheric ' stability at the ig *
- request of the Commission. .
S 1 l j- L6.7 A 6. Core Oneratine Limits,Recort
.t a.. Core: operating limits,shall bc established and documented in'the y
CORE OPERATING LIMITS REPORT before each reload' cycle or:any ; remaining part of alreload cycle for: the following:
~
- 1. Heat Flux Hot Channel Factor Limit (FqRPT), Nuclear Enthalpy- l Rise Hot Channel: Factor, Limit (F ),,PFDH, K(Z) and V(Z)-
-(Specifications 3.10B,1,3.10N2and3.10B.3)
Wh 3 -
'~
2.'AxialnFlux Difference Limits and-Target Band (Speelfications:3.10.B.4 through 3.10.B 9)
.3. Shutdown and Control Bank-Insertion Limits -
J(Specification 3.10.D)- l
- 4. Reactor CoolantLSystem Flow Limit t
'(Specification,3.10,J)
- b. The 'analytica1' methods used to determitie the core operating .
limits shall be those previously reviewed and approved by the 7 NRC,vspecifically those described in.the following-documents: NSPNAD-8101-A, " Qualification of Reactor Physics Meth'ds o for s Application to PI Units" (latest approved version)' NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation' Method. for Application to PI Units" (latest approved version) WCAP-9272-P-A, " Westinghouse Reload Satety Evaluation Methodology", July, 1985 1: *
- WCAP-10054-P-A, " Westinghouse Suall Break ECCS Evaluation Model Using'the NOTRUMP Code", August, 1985 WCAP 10924 P-A, " Westinghouse Large-Break LOCA Best-Estimato Methodology", December, 19"3
., ; XN-NF-77 57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981
- c. .The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such rs shutdown uargin, and transient and accident analysis limits) of ,
the safety enalysis are met. 4 7
+
; , O k{gh([N
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s TS.6.7-5 [ R , i
- cf?' ,
& d.:The' CORE' OPERATING LIMITS REPORT, including any mid. cycle revis. ions ;
)' 4ii? or supplements thereto, shall be supplied upon(issuance, for each e
>< reload cycle, to the NRC Document Control Desk with copies:to the- L Regional Administrator and Resident Inspector. ;
5 g , B. REPORTABLE EVENTS The following actions shall be-taken for REPORTABLE EVENTS:
;{' ;
- a. 'The-Commission shall be notified by a report submitted pursuant: ,
" -to the. requirements of Section 50.73 to 10 CFR Part 50, and- q
-1 i '
'b. Each REPORTABLE EVENT'shall be reviewed by the Operations- ;
Committee and the results of this review shall-be submitted tol i the Safety. Audit Committee and the Vice President Nuclear
~
l 1 Generation'. <
.j C, Environmental Reports .h N. -.?W The reports listed below shall be submitted to the Administrator of 1 the appropriate Regional NRC Office or his designate:.
- yt .;
1
, 7 ,
- 1. Annual Radiation Environmental Monitorine Reoort "
i (a) Annual Radiation Environmental Monitoring Reports covering. the operation of the program during the previous calendar-p '
, year shall be' submitted prior to May 1 of each year. ;
o 1
, (b) The Annual Radiation Environmental Monitoring Reports-shall j k include summaries. interpretations, and an analysis of trends of the results of the radiological environmental -
- surveillance activities for the report period,. including a comparison with preoperational studies, operational controls (as appropriate), and previous; environmental ,
surveillance reports and an assessment of the observed
~
7 '% impacts of the plant operation on'the environment. The reports shall also' include'the results of land use censuses . required by Specification 4.10.B.1. If harmful effects or evidence of irreversible damage are detected by the M monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem. t g 4 (c) The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of
- Re6ulatory Guide 4.8, December 1975 of all radiological y environmental sampics taken during the report period. In the event that some results are not available for inclusion }
( with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possib'e in a supplementary report. s
=^
i-
l 1- l
@p' ,
7s %y, f.[g Yf 4 < i q%g : , l , TS.6.7 6- ; n k ,.j,
* " '(d). The reports shall also include the following: a summary' '~w description'of the radiological environmental-monitoring j
e program; a map of all sampling locations keyed to a' table - 4 c giving distances and directions from one reactor; and the results of licensees participation-in the.Interlaboratory ; 7 : Comparison Program, required by Specification 4.10.C.1.: i
^
m, , 1 ic ,2. Environniental Soecial Renarts ' l [ s 9 m _ (a) =When radioactivity _ levels in samples exceed limits '; P- specified in Table l4.10-3, an Environmental Special Report l
- shall be submitted within'30 days from the end of the y affected calendar quarter._ For certain cases involving: ..; ;) long analysis _ time, determination of quarterly averages.
H ' may. extend beyond the 30 day period. In these' cases the- " a potential-for exceeding the quarterly limits wil1}be reported within the 30 day period to be followed by thes R Environmental Special Report as soon as' practicable.
- s i U .' L3. Other Environmental Renorts (non-radiological. non-aountic) o y
lWritten reports for the following items shall.be-submitted to- '. f the' appropriate NRC' Regional-Administrator:
- a. Environmental events ~that indicate or could= result in a- i significant: environmental impact casually related to plant 4
. operation. -The following are examples: excessive bird ;;
- impaction; onsite plant or' animal disease outbreaks;_- y
, unusual mortality of;any species protected by the ;
Endangered Species Act of_1973; or-increase in nuisance l l' organisms or conditions. This report shall be submitted i within 30 dayt of the event and shall-(a) describe, analyze, ;
lD and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describc. the probable cause of.the event,-(c) indicate the +
[ -action taken to correct the reported event,'(d) indicate - the' corrective' action taken.to preclude repetition of the i event and to prevent similar occurrences involving-similar. ) j, components _or systems, and (e) indicate the agencies
.,1 L notified and their. preliminary responses.
~ l b. ' Proposed changes, test or experiments which may result in a significant increase in any adverse environmental impset which was-not previously reviewed or evaluated in the Final L, j Environmental Statement or supplements thereto. This report shall include an evaluation of the environmental impact of the proposed activity and shall be submitted 30 , days prior to implementing the proposed change, test or ; experiment. D. Special Reports Unless otherwise indicated, special reports required by the Technical - Specifications shall be submitted to the appropriate NRC Regional Administrator within the time period specified for each report. , j
?
i ,, , - - -
, J
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- B.2.1 2= ,;
r
.[
bI %
.2.1" SAFETY LIMIT: ' REACTOR CORE sie -;
y 3. . Bases continued <
<u.
power levels of 91% and(74% respectively. For the 2235 psig and 2385- ; l psig curves', the coolant average temperature at the core exit is; equal ; to 650*F below power, levels of 64% and.73% respectively, j
, , c
. 1 -The third and fourth criteria are evaluated using standard-DNB metho ;
dology.- For alllfour curves the DNBR is limiting.at higher power levels.
~
.The; area of_ safe operation is below these curves. .,
J The plant conditions required to violate the limits in the lower power j range are; precluded by the self-actuated safety valves on the steam -; generators.L The. highest nominal setting of the'~ steam generator safety l valves:is 1129'psig,(saturation temperature 560*F). :At zero power the
' difference;between primary coolant and secondary' coolant is:zero and at: ;
p , full powerfit is 50'F. The reactor conditions at which steam generator safety valves open is shown as a dashed line-on Figure TS.2.1-1. , Except for special tests, POWER OPERATION with only one loop or with ;
, . natural circulation'is not allowed. Safety limits for such special- /,
1 x tests:will be determined as a part of the. test procedure. , O The curves are conservative for the following nuclear hot, channel factors:
?
+ N TP (H ~ TP-[1 H
+ PFDH(1-P)] ; and Fq, Lwhare: ,
4 tt f
- 1 TPis the . limit at RATED THERMAL POWER specified in the CORE LO ERATING LI ITS-REPORT.
' RTP
-F AH is the FAH limit at RATED THERMAL POWER specified in the CORE -{
OPERATING - LIMITS' REPORT. 'i
-PFDHis.thePowerFactorMultiplierforPfHspecifiedintheCORE "
OPERATING LIMITS REPORT.
~.i Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.
This combination of hot chcanel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10. Adverse power distribution factors could occur at lower power levels because additional control , rods are in the core. However, the control rod insertion limits spacified in the CORE OPERATING LIMITS REPORT assure that the DNB ratio l is always greater at part power than at full power. The Reactor Control and Protective System is designed to prevent any
; anticipatad combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel.
\
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( 6,
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i
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'* . B.3.10-1 3i10; CONTROL ROD AND POWER DISTRIBUTION LIMITS 3 > 1 '
BAERS. a' [Throughoutlthe3.10 Technical.Specif$ cations,the. terms" rod (s)"and
- D4 L
"RCCA(s)" are synonymous, f
-.A; Shutdown Margin: ,
-s Trip shutdown reactivity is'provided consistent with plant safety.
- analyses assumptions. One percent shutdown margin is adequate except for the steam break analysis which requires more shutdown reactivity. ;
c due to the.more negative moderator temperature coefficient at'and of ' t 111fet(when boron. concentration is low). Figure TS.3.10h1 isidrawn:
! Jaccordingly, ;
~
. Power Distribution Control
~
B. The specifications of this section provide assurance of fuel integrity; i: during Condition I (Normal Operations) and II (Incidents of Moderate
. frequency) events by: (a) maintaining the minimum DNBR in the core of
'l
- y greater than or equal'to 1.30'for Exxon fuel and 1.17 for Westinghouse y
~^ < fuel during' normal operation and in short term transients, and (b), , l 1 ,11:r.iting the' fission gas release, fuel; pellet temperature and cladding
. mechanical properties to within assumed design criteria. The ECCS '
- analysis'was performed in accordance with SECY 83-472- One calculation
. -(
j 1at the 95% probability level was performed as well.as'one' calculation with all..the required foatures of 10 CFR Part 50,: Appendix K. The,95%
- probability level calculation used the pe'ak linear heat-generation-rate specified'in the CORE OPERATING LIMITS REPORT. The Appendix' K :
icalculation ' used the peak linear: heat generation rate. specified in the CORE OPERATING LIMITS REPORT for1 the.Fg limit specified in the CORE
.0PERATING LIMITS REPORT. Maintaining I) peaking factors below the Fq . t
' limit specified:In the CORE 0PERATING LIMITS REPORT during all Condition !
'I e' vents and 2)'the: peak linear heat generation rate below the value specified!in the CORE' OPERATING' LIMITS REPORT at the 95% probability level assures compliance with the ECCS analysis.
, a During. operation, the plant; staff compares the measured hot channel ,
factors, FN- d AH, (described later) to the limits' determined in the j
. transient qand an LOCA analyses. 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 uncertaintiec.
-%isthemeawc+1NuclearHotChannelFactor,definedasthemaximum 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 specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fq axially. The K(Z) value is based on , large and small break LOCA analyses, i l 1
~
' ' ^
h }l:, jf J , .
' 'O r 6
so ] u> ' I _l h,' ' B.3.10-3 j i he I h -3.~10' CONTROL ROD AND POWER DISTRIBUTION LIMITS
- y. .;
Bases -continued. WhenameasurementofFfH is taken, measurement error must be allowed i for and 4' percent is the appropriate allowance for a fu11Leore map y
?' taken with the movable.incore detector flux mapping system. .;
. Measurements:of the hot channel factors are required as part of startup
<THYSICS TESTS,:at least once each effective full power month of operation, j
-and whenever; abnormal power distribution conditinns require a reduction of l
, core power-to a level based on measured hot channel factors. The incore JI map 'takenL following initial ~ loading provides confirmation of the basic {
nuclear design' bases: including proper fuel loading patterns. The periodic. 1 monthly incore mapping provides' additional assurance that the nuclear- ' ..'c ; design bases remain-inviolete and identify operational anomalies which would Lotherwise affect-these bases, [' For normal-operation,-.it-is not'necessary to measure these quantities.
- Instead:it ,has been de'terained that, provided certain conditions 'are
~
d observed. . the hot channel' factor limits will be met; these conditions are , 4 as follows: ; r 1.' Control rods in a single bank move together with no
~
' individual rod insertion differing by more than'15 ;
inches from the bank demand position. An accidental
^
g .
. misalignment *11mit of 13 steps precludes a rod misalign- ;
ment greater.than 15. inches with' consideration of maximum .! instrumentation error. . 2 ~. Control rod banks are sequenced with overlapping banks
.as described.in Technical Specification 3.10. .
- 3. The . control bank insertion liraits specified in the CORE OPERATING '
LIMITS REPORT are not violated, j
- 4. Axial power distribution control' procedures, which are given in terms of flux difference control and control bank insertion .
limits are observed. Flux difference refers to the difference in 31gnals-.between the top and bottom halves of two-section excore neutron' detectors. The flux difference is a measure of the axial offset which is defined as th, difference in normalized power between the -top and bottom halves of the core,
.i k
J
TV b,q-.; - < K x en b h :x . O' . B.3.10a4-3.10~ CONTROL ROD'AND POWER DISTRIBUTION LIMITS Bases ' continued
' B. : , Power Distribution Control- (continued)
Thepermitted-relaxationinF$HandFhallowsforl radial-powershape k L : changes'wich rod insertion-to the insertion limits. It has been, fdetermined that provided the.above conditions 1 through 4 are obse ;e d , these hot channel factor limits are net. . In specification 3.10,. is-
- arbitrarily limited for P less than or equal to 0.5.(except for lo power PHYSICS TESTS).
The. procedures'for axial power distribution' control referred to above are' designed to minimize the effectsJof xenon redistribution _on the axial power distribution during load-follow maneuvers. Basically control'of flux difference is required to: limit the difference between-the current value of Flux Difference ( AI)' and a reference .value which ! corresponds to the fullEpower equilibrium value of Axial Offset (Axial l Offset - AI/ fractional power). The reference value of flux difference -! varies with power level and burnup but expressed as axial offset it-varies ~only with burnup. The technical specifications on power, distribution control assure that
-the Fq111mit_is not exceeded and xenon, distributions are-not developed !
i which at a later time,' would cause greater local. power peaking even though . .; the flux difference is then within the limits:specified by-the procedure. H 4 The: target (or reference) value of flux difference'is determined as
~follows: At any time that equilibrium-xenon conditions have been established, the indicated-flux difference is noted with the' full length rod control rod bank more than 190 steps withdrawn (1.e., normal full power operating-position appropriate for the time in life, usually y ; withdrawn farther as burnup' proceeds). This value, divided by the -;
L -fraction of full power at which the core was operating is the-full i power value of the target flux difference. Values for all other core ; power levels are obtained by multiplying the full power value1by the ;
; fractional power. Since the indicated equilibrium was noted, no _'
i_ allowances for excore detector error are necessary and indicated j l< deviation from-the indicated reference value but within the target band '
~
is permitted. The allowed deviation from the target flux difference as
- l. a function of THERMAL POWER is specified in the CORE OPERATING LIMITS-l u REPORT.
I-l$ g l. I 1 1
; su ,
1 -f ,,;;f.,y'y g , # . t ;W B.3.10 6 M / ( ,; 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS , Bases continue'dl n s- , [3
~
JB. ' Power Distribution Control (continued) a
- ;InLsome instances'of' rapid plant power reduction, automatic.-rod ,
eW ' motion will'cause the flux difference to deviate from~the target . 4 zbandLwhen the reduced power level is reached. This does not J M' ;necessarily affect the xenonLdistribution sufficiently-_to change- ;
-the1 envelope of peaking factors'which can be reached on a subse--
C quent return to full power within the target band,l however. to ; simplify the specification, a limitation of one hour in.any period g .of 24 hours is placed on operation outside'the band. This' ensures F '
- that the resulting xenon distributions are not significantly .
.different'from those resulting from operation within the target band. '
t p The consequences lof being'outside the target band but within the limits
- y. , specified in the. CORE OPERATING LIMITS REPORT for power levels between-50% and 904:has been evaluated and determined to result in acceptable peaking factors. Therefore, while the-deviation exists the power = level
.is limited to.90. percent or<1ower depending on the indicated axial flux.
> difference. In all cases-the target band is the Limiting Condition for .
Operation. Only when the target band is violated do the limits
~
specified in the. CORE OPERATING LIMITS REPORT-apply. If, for any reason, the-indicated axial flux difference is not control- 'I led'within: the target band for as;1ong a' period .as one hour, then xenon l
- di~stributions may be significantly changed and operation at.or below 50 percent:is' required to protect against potentially more severe-i consequences.of-some accidents.
Asidiscussedlabove, the essence of the procedure'is to maintain the xenon distribution in the core'as close to the equilibrium full power , condition as possible. This is accomplished by using the-boron system - to position the full length control rods,to produce the required ,
- 6. f '
indicatied' flux difference. For Condition-II events the core is protected from overpower and a- ' minimum DNBR of 1.30.for Exxon fuel and 1.'17 for Westinghouse fuel by' an1 automatic protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, i operator errorLand equipment malfunctions are separately assumed to lead to the cause of the transients considered. C. QUADRANT POWER TILT RATIO QUADRANT POWER TILT RATIO limits are based on the following considera-tions. . Frequent power tilts are not anticipated during normal operation since this phenomenon is caused by some asymmetric perturbation, e.g. rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F, and
', (," ',!
Ja ?;; p .
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t . f- '! us: , / , : P14.IL 3;f[ ' '
.4[ ,
. B.3.10-10 J
j, 3.10-CONTROL ROD AND EQKER DISTRIBUTION LIMITS f
, Bases continued , .
}
~
.H. - Rod Drop Tine; j The required drop time'to dashpot entry is~ consistent Oith the safety: I analysis.
4 $ I, fMonitor Inoperability, Requirements ; l' ,If either the, rod bank insertion limit monitor'or rod position devia- 'f
-tion monitorcarefinoperable,Ladditional surveillance is. required to, "
h ensure-adequate, shutdown margin is maintained-i ;
, .. . . , -f If the' rod. position deviation monitor and quadrant power tilt monitor (s) y .are'inoparable,,the ovorpower' reactor trip setpoint is reduced-(and also .;
power)ito ensure that adequate. core protection is provided.in the event-ithat unsatisfactory? conditions arise that could affect. radial power
, distribution, s , .
,, ' Increased surveillance is required, if the quadrant power tilt monitors
'EC '
, are inoperable and a load change occurs, in order to confirm satisfac- '
n s' tory power distribution behavior. 1Rie automatic alarm functions related '
'V .to QUADRANT POWER TIL,T must be considered incapable of-alerting the -l
, 7; operator to unsatisfactory power distribution conditions. ,
J J. ' DNB' Parameters v ~
- w. .
and Pressurizer Pressure requirements are' based The;RCSflow' rate,T,y$s,sumptions.The on transient analyses 2 flow rate shall be verified by calorimetric flow data and/or elbow taps. Elbow; taps are used'in the (
. reactor coolant system as'an instrument device:that-indicates the status
.. :of the reactor coolant flow. The basic function of this device is to -
provide information as to whether or not a reduction in-flow rate has occurred. If a reduction in flow rate is indicated'below the value - specified in the CORE OPERATING LIMITS' REPORT, shutdown is required to ir investigate adequacy of core cooling during operation. 1 ii . ' 1 4
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~~ Prairie Island Nuclear. Generating Plant?
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.a +, .- n iLicense-Amendment' Request' Dated November 17,- 1989 pe
, SAMPLE CORE OPERATING LIMITS REPORT i: z
' UNIT.1-- CYCLE 13 .;
m.; UNIT 2 - CYCLE 13- m e,
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CORE OPERATING LIMITS REPORT 4 d Unit'l - Cycle 13 Unit 2.- Cycle 13 Revision 0 -; w i 1 . I
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-- CORE OPERATING LIMITS REPORT y ,
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,1 l , . Unit l1 - Cycle _13 1
. . - Unit 2L- Cycle:13: :d 1 jf y '
+
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- q. .g Revision 0-O
,s i b jThisreportprovidesthe.valuesofthelimits--for it l' Cycle 13 and Unit 2-Cycle 13.as; required /by Te.chnical Specification etion 6.7.A.6. These values c , -have been established using.NRC! approved metho logy and are established-such
, that allEapplicablei .licits of..the plant safety a lys's are met.
>y ,-
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~-Heat Flux Hot ChaDDfl Egctor'Lirgit;3,
'l ,,
- i ' RTP.- > 'N
% ,[1 J Fq V2.50 : K(Z)Ivalses;are provided.in F ure 1 lM 3(Z) values are provided in Fig 2. Reference Technical-Speci c on ec ons: 3.10.B.1 and 3.10'.B.2 y s
- Nuclear Enthalov Ris Hot anne Factor Limits' :3.;
x i
. " RTF S LFon - 1.7 v .
> PFDH - 0.3 ;;j n_'.
. If'the. nuclear en Ipy rise hot channel factor exceeds its limit in Technical j 1 -Specification 3.10.B.1, reduce the high neutron flux trip setpoint.by 3.33% 1
-for each percent that the measured nuclear enthn1py rise hot channel factor ')
E exceeds the 3.10.B.1 limit. <
- r Reference Technical Specification Sections: 3.10.B.1, 3.10.B.2 and 3.10.B.3 .lu 1
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.., Mnear Heat Ceneration Rat;,g K ,
lThe 956 probability level ECCS analysis calculation utilized a peak linear llj he.at generation rate of 14.2 kw/ft.
. The' Appendix-K ECCf analysis calculation utilized a peak linear heat
. gen 3 ration rate of 15.8 kw/f t foz the Fq limit of 2.5.
t
-Reference Technical Specification Section: 3.10.5 ; Axial Flux Difference 13mf ts The axin1' flux ' difference limits are provided in F1 re 3. l ; The Axial Flux Difference target band is i$4. ;
7
;- Reference Technical Cpecification Sections: 3.10.B. through .10.B.9 jhg down Rod Insertion Limits s The shutdown rods shd11 be fully ithdra .
l. L ,, i - -~3eference Technical Specificat n Se t on: 3.10.D i 1: ' a t'ontrol Rod ' Insertion limen The control" rod ba s shall we ed in physical insertion as shown in L ; Figures 4. $ and.6.
' Reference Tea cal S e [ica ou action: 3.10.D Reactor Cno low mt >,
, The' reactor coolant stem flow shall be k 178,000 gpm.
' Reference Technical Specification Section: 3.10.J
~
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- sting Li'its Report 7.
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' !j . , t ?!k , Pr+ pared By:
Eugene Eckholt Date t Reviewed By: Leeve Schaefer Date Superintendent of Nuclear Engineering Prairie Island Rkviewed By: Roger Anderson Date Manager NucIsar Analysis Approved By: \ Thomas M Parker Date
; p Manager Nuclear Support Se ces
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Core Operating Llatts Report
'o' . Unit 1, Cycle 13 cnd Unit 2, Cycle 13
, Revision 0
' Page 4 of 9 1.2 I
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Figure 1 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE d
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Core Operating L121ts Report l oy Unlt 1, Cycle 13 cnd Unit 2. Cyc1e 13 - Revision 0 Pope $ of 9 ) t i i 1.20 , 1.18
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Figure 2 V(Z) AS A FUNCTION OF CORE HEIGHT e -l 1
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CIre operating Lbits Report l Unit i, Cyct a 13 and Unit 2, Cyclo 13 Revision 0 -
- Page 6 of 9 F
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