ML19351F247
| ML19351F247 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/17/1980 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19351F248 | List: |
| References | |
| NUDOCS 8101120056 | |
| Download: ML19351F247 (25) | |
Text
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NUCLEAR REGULATORY COMMISSION v.,
E WASHINGTON, D. C. 20555
%[CrRf NORTHERN STATES POWER COMPANY DOCKET N0. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 44 License No. DPR-42 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Northern States Power Cogany (the licensee) dated May 6,1980, as supplemented Septenber 19 and December 2,1980, co@ lies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be corducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cogliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
81011200 %
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-42 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
44, are hereby incorporated in the license. The licensee shall operate the f acility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION be R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
l Changes to the Technical Specifications Date of Issuance: December 17, 1980 l
- p* **%,og UNITED STATES 8
s.(
NUCLEAR REGULATORY COMMISSION a
7, g jj WASHINGTON, D C. 20555 gpcW/
NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 38 License No. DPR-60 1.
The Nuclear Replatory Comission (the Comission) has found that:
A.
The application for amendment by Northern States Power Co@any (the licensee) dated May 6,1980, as supplemented Septenter 19 and December 2,1980, conplies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in conpliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been' satisfied.
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 38, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This iicense amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y
NL' R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
l Changes to the Technical Specifications Date of Issuance:
December 17, 1980 l
l l
l
ATTACHMENT TO LICENSE' AMENDMENTS NOS. 44 AND 38 FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The changed areas on the revised pages are reflected by marginal lines.
Remove Pages Insert Pages TS iv TS iv TS 3.5-2 TS 3.5-2 Table TS 3.5-1 Table TS 3.5-1 Table TS 3.5-3 Table TS 3.5-3 TS 3.10.1 TS 3.10.1 TS 3.10-1A TS 3.10-2 TS 3.10-2 TS 3.10-3 TS 3.10-3 TS 3.10-4 TS 3.10-4 TS 3.10-5 TS 3.10-5 TS 3.10-6 TS 3.10-6 TS 3.10-7 TS 3.10-7 TS 3.10-8 TS 3.10-8 TS 3.10-9 TS 3.10-9 TS 3.10-10 TS 3.10-10 TS 3.10-10a TS 3.10-11 TS 3.10-11 TS 3.10-12 TS 3.10-12 TS 3.10-13 TS 3.10-13 TS 3.10-13a TS 3.10-14 TS 3.10-15 TS 3.10-16 TS 3.10-17 Figure TS 3.10-7 Figure 3.10-7
i TS-iv APPENDIX A TECHNICAL SPECIFICATIONE LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Ef fect of Fluence and Copper Content on Shif t of RI NDT Reactor Vessel Steels Exposed to 550 Temperature 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 control Bank Insertion Limits 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 For F = 2.21 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 Normalized Exposure Dependent Function BU(E;) for Exxon Nuclear Company Fuel 3.10-8 V(Z) as a function of core height 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 4.10 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-site Operating Organization 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group Prairie Island Unit 1 - Amendment No. 35, 44 i
i Prairie Island Unit 2 - Amendment No. 29, 38
TS.3.5-2 Safety Injection The Safety Injection System is actuated automatically to provide emergency cooling and reduction of reactivity in the event of a loss-of-coolant accident or a steam line break accident.
Safety injection in response to a loss-of-coolant accident (LOCA) is provided by a high containment pressure signal backed up by the low pressurizer pressure signal.
These conditions would accompany the l
depressurization and coolant loss during a LOCA.
Safety injection in response to a steam line break is provided directly by a low steam line pressure signal, backed up by the low pressurizer pressure signal and, in case of a break within the containment, by the high containment pressure signal.
The' safety injection of highly borated water will offset the temperature-induced reactivity addition that could otherwise result from coo'idown following a steam line break.
Containment Spray Containment sprays are also actuated by a high containment pressure signal (Hi-Hi) to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment.
The containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is safety injection (10% of design).
Since spurious actuation of containment spray is to be avoided, it is initiated on coincidence of high containment pressure,
sensed by three sets of one-out-of-two containment pressure signals provided for its actuation.
Containment Isolation A containment isolation signal is initisted by any signal causing automatic initiation of safety injection or may be initiated manually.
The containment isolation system provides the means of isolating the various pipes passing through the containment walls as required to prevent the release of radioactivity to the environment in the event of a loss-of-coolant accident.
Prairie Island Unit 1 - Amendment No. 35, 44 Prairie Island Unit 2 - Amendment No. 39, 38
m,
,a n
TABLE TS.3.5-1 m a
- r :p p gp ENGINEERED SAFETY FEATURES INITIATION INSTRUNENT LIMITING SET POINTS FUNCTIONAL UNIT CilANNEL LIMITING SET POINTS
- m m I
liigh Containmer.t Pressure (lii)
Safety Injection
- f,4 psig c c 2
liigh Containmeny Pressure (Ili-lii)
- a. Containment Spray f,23 psig
- b. Steam Line Isolation
"~
~<17 psig of Both Lines 8
9 3
Pressurizer Low Pressure Safety Injection *
>1815 psig (l
4 Low Steam Line Pressure Safety Injection *
>500 psig 7, 3; Lead Time Constant
>12 seconds Lag Time Constant f,2 seconds o o
[, [,
5 liigh Steam Flow in a Steam Line
' Steam Line Isolation d/pcorrespgndingto ya ja Coincident with Safety Injection of Affected Line f,0.745 x 10 lb/hr and Low T at 1005 psig g; j;.
avg
>540 F 6
High-high Steam Flow in a Steam Line Isolation f,d/p correspgnding Steam Line Coincident with of Af fected Line to 4.5 x 10 lb/hr Safety Injection at 735 psig 7
liigh Pressure Difference Between Containment Vacuum f,0.5 poi Shield Building and Containment Breakers 8
High Temperature in Ventilation Ducts Ventilation System f,120 F Isolation Dampers e
- Initiates also containment icolation, feedwater line isolation and starting of all containment fans.
e d/p means differential pressure i
m,a 55
- p :p TABLE TS.3.5-3
_~.
INSTRUMENT OPERATING CONDITIONS FOR EMERCENCY COOLING SYSTEM EI C7 i; 10
, RR 1
2 3
4 c: c=
MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF EL El OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN
{*
FUNCTIONAL UNIT CilANNELS REDUNDANCY, CONDITIONS 1 or 2 CANNOT BE MET i
8 1.
SAFETY INJECTION
!U $I a.
Manual 2
1
$I OI Ilot shutdown **
b.
liigh Containment Pressure 2
1 4
re re Hot shutdown **
ll ll c.
Steam Generator Low Steam 2
1 primary pressure Hot shutdown **
Pressure / Loop I$I$
less than 2000 psig d.
Pressurizer Low Pressure 2
1 primary pressure llot shutdown **
$$ jI less than 2000 psig
\\
2
a.
Manual 2
Hot shutdown **
b.
Hi-Hi Containment Pres-Hot shutdown **
sure (Containment Spray)
Channel a 2
1 Channel b 2
1 Channel c 2
1 Logic 2
1 o*
- Must actuate 2 switches simultaneously, o **
- If minimum condicions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken to place the unit in cold shutdown condition.
TS,3.10-1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability i
Applies to the limits on core fission power distribution and to the limits on control rod operations.
Objectire To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during power operation, and 3) limited potential reactivity in-sertions caused by hypothetical control rod ejection.
Specification 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 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 reactor core would be 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.
l B.
Power Distribution Limits 1_
1.
At all times, except d ring lgw power physics testing, measured hot channel factors, and F.d, as defined below and in the bases, shall meet the ollowing limits:
N F x 1.03 x 1.05 1 (2.21/P) x K(Z) x BU (E )
F x 1.04 1 1.55 x (1+ 0.2(1-P)]
H where the following definitions apply:
(a) K(Z) is the axial dependence function shown in Figure TS.3.10-5.
(b) Z ie the core height location.
(c) E is the maximum pellet exposure in fuel rod j for which I
- t. e F is being measured.
l (d) BU(E ) is the normalized exposure. dependence function for 4
i Exxod Nuclear Company fuel shown in Figure TS.3.10-7.
For Westinghouse fuel, BU(g) = 1.0 (e) P is the fracti n f full power at which the core is operating.
N In the F limit determination when P 1 50, set P = 0.50.
q l
Prairie Island Unit 1 - Amendment No. 35, 44 Prairie Island Unit 2 - Amendment No. 29, 38
TS.3.10-2 (f)
F or F is defined as the measured F or F respectively, w9ththesmallestmarginorgreatestekcasshk, limit (g) 1.03 is the engingering hot channel factor, F, applied to the measured F toaccountformanufacturibgtolerance.
N (h) 1.05 is applied to the measured F to account for measurement q
uncertainty.
1.04 is applied to the measured FfH (i) t ace unt f r measure-,
ment uncertainty N
N 2.
Hot channel factors, F and FAH, shall be measured and the target n
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 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.
[ (equil) shall meet the following limit for the middle axial 80%
o9thecore:
F (equil) x V(Z) x 1.03 x 1.05 < (2.21/P) x K(Z) x BU(E )
where V(Z) is defined Figure 3.10-8 and other ternas 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 neutqon fl trip set-point by 1% for each percent that the measured Fg or excee s AH the 3.10.B.1 limit. Then follow 3.10.3.3(c).
the 3.10.B.1 limi9,(equil) exceeds the 3.10.B.2 limits but not (b)
If the measured F taxe 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 configura-tion for which Specification 3.10.B.2 is satisfied, or 2.
Reducereactorpowerandthehighneutronfluxtripsetpoint by 1%'for each percent that the measured F x 1.05 x V(Z) exceeds the (2.21/P) x K(Z) k(equil) x 1.03 BU(E ) limit.
1 l
l l
Prai:1e Island Unic 1 - Amendment No. 35, 44 P airie Island Unit 2 - Amendment No. 29,38 i
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T TS.3.10-3 (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% power for the purpose of physics testing.
Identify and correct the cause of the out of limit condition prior to increasing thermal power abgve 50 ThermalpowermaythenbeincreasedprovidedF"orF'% power.
is demonstratedthroughin-coremappingtobewickinitgHyg,1t,,
l (d)
If two successiv peak pi.a power F'q measurements indicate an increase in the l
with exposure, either of the following actionsshallbe$aken:
0 l
1.
F (equil) shall be multiplied by 1.02 x V(Z) x 1.03 x 1.05 for comparison' co the limit specified in 3.10.B.~2, or N(equil)shallbemeasuredatleastonceperseven 2.
ehfectivefullpowerdaysuntiltwosuccessivemaps indicate that the peak pin power, is a t increasing.
AH l
4.
Except during physics tests, end except as provided by Specifications 5 through 8 below, the indicated axial flux difference for at least three operable excore channels shall be maintained within a +5% band about the target flux difference.
l S.
Above 90 percent of rated thermal power:
l If the indicated axial flux difference of two operable excore channels deviates from its target band, within 15 minutes either eliminate such deviation, or reduce thermal power to less than 90 percent of rated i
thermal power.
6.
Between 50 and 90 percent of rated thermal power:
l l
a.
The indicated axial flux difference may deviate from its +5%
target band for a maximum of one* hour (cumulatire) 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 shown in Figure TS.3.10-6.
I b.
If 6.a is violated for two operable excore channels then the reactor power shall be reduced to less than 50% power and the high neutron flux setpoint reduced to less than 55%
i of rated power.
- May be extended to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> during incore/excore calibration.
Prairie Island Unit 1 - Amendment No. 35, 44
?rairie Island Unit 2 - Amendment No. 29, 38
TS.3.~10-4 A power increase to a level greater than 90 percent of rated c.
power is contingent upon the indicated axial flux difference of least three operable excore channels being within the target l
at band.
7.
Less than 50 percent of rated thermal power:
l The indicated axial flux difference may deviate from its target a.
band.
b.
A power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference of at least three operable cxcore channels not being outside the target I
band for more than one hour (cumulative) 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 In applying 6a and '7b above, penalty deviations outside the f,5%
8.
target band shall be accumulated on a time basis of:
One minute penalty deviation for each one minute of power a.
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.
9.
If alarms associated with monitoring the indicated axial flux difference deviations from the 2,5% target band are not operable, the indicated axial flux difference value for each operable excore channel shall be logged at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and half-hourly thereaf ter until the alarms are returned to an operable status.
For the purpose of applying l this specification, logged values of indicated axial flux difference must be assumed to apply during the previous interval between loggings.
C.
Quadrant Power Tilt Limits 1.
Except for physics tests, if the percentage quadrant power tilt exceeds 2% but is less than 7%, 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 steps shall be taken:
a.
Correct the tilt to less than 2%
b.
Restrict core power level so as not to exceed rated power, less 2% for every percent that the quadrant power tilt ratio exceeds 1.0, Prairie Island Unit 1 - Amendment No. 29, 44 l
Prairie Island Unit 2 - Amendment No. 23, 38
TS.3.10-5 2.
If the percentage quadrant power tilt exceeds 2% but is less than 7%
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 brought to the hot shutdown condition.
Subsequent operation below 50% of rating, for testing, shall be pennitted.
3.
Except for physics tests if the quadrant power tilt ratio exceeds 1.~07, the reactor shall be brought to the hot shutdown condition.
Subsequent operation below 50% of rating, for testing, shall be permitted.
A.
If the core is operating above 85% power with one excore nuclear chann:1 out of service, then the core quadrant power balance shall be determined daily and af ter a 10" power change using either 2 movable detectors or 4 core thermocouples per quadrant, per Specifi-cation 3.11.
D.
Rod Insertion Limits 1.
The shutdown rods shall be fully withdrawn when the reactor is critical or approaching criticality.
- 2.
When the reactor is critical er approaching criticality, the control banks shall be limited in physical insertion; insertion limits are shown in Figure TS.3.10-2, -3 and -4 for normal and abnormal operating conditions.
3.
Control bank insertion may be further restricted by specification l
3.10. A if, (1) the measured control rod worth of all rods, less the worth of the worst stuck rod, is less than 5.52% reactivity at the beginning of the first cycle or the equivalent value if measured at any other time, or (2) if a rod is inoperable (Specification 3.10.G).
4.
Insertion limits do not apply during physics tests or during periodic l
~
exercise of individual rods.
The shutdown margin shown in Figure TS.3.10-1 must be maintained except for low power physics testing.
For I
this test the reactor may be critical with all but one high worth full-length control rod inserted for a 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 run on the high worth full-length rod prior to this particular low power physics test.
Prairie Island Unit 1 - Amendment No. 32, 44 Prairie Island Unit 2 - Amendment No. 26, 38
~
TS.3.10-6 E.
Rod Misalignment Limitations 1.
If a full-length rod cluster control assembly (RCCA) is misaligned from its bank by more than 15 inches, the rod will be realigned or the core power peaking factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, i
and Specification 3.10.B applied.
If peaking factors are not determined l
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the high neutron flux trip setpoint shall be reduced to 85 percent of rating.
2.
If the misaligned RCCA is not realigned within a total of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the RCCA shall be declared inoperable.
F.
Inoperable Rod Position Indicator Channels 1.
If a rod position indicator (RPI) channel is out of service then a.
For operation between 50% and 100% of rating, the position of the RCCA shall be checked directly by core instrumentation l
(excore detector and/or thermocouples and/or movable incore detectors) every shif t or subsequent to rod motion exceeding a total of 24 steps, whichever occurs first.
b.
During operation below 50% of rating,. no special monitoring is required.
2.
The plant shall be brought to the hot shutdown condition should more than one RPI channel per group or more than two RPI channels l
per bank be found to be inoperable during power operation.
3.
If a full length rod having a rod position indicator channel out of service is found to be misaligned from 1.a. above, then apply Specification 3.10.E.
G.
Inoperable Rod Limitations 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 I
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 />.
Prairie Island Unit 1 - Amendment No. 32, 44 Prairie Island Unst 2 - Amendment No. 26, 38 l
l L
TS. 3.10-7 1
2.
The plant shall be brought to the hot shutdown condition should more than one inoperable full length rod be discovered during poi e operation.
3.
If the inoperable full-length rod is located below the 200 step level and is capable of being tripped, or if the full-length rod is located below the 30 step level whether or not it is capable of be'ing tripped, then the insertion limits in Figure TS.3.10-3 apply.
l i
4.
If the inoperable full-length rod cannot be located, or if the inoperable full-length rod is located above the 30 step level and cannot be tripped, then the insertion limits in Figure TS.3.10-4 l
apply.
5.
If reactor operation is continued with one inoperable full-lengt's rod, the potential ejected rod worth and associated tranrient power distribution peaking f actors 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 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, the plant power level shall be reduced to a level consistent with the safety analysis.
H.
Rod Drop Time At operating temperature and full flow, the drop time of each full-length RCCA shall be no greater than 1.8 seconds from loss of stationary l
gripper coil voltage to dashpot entry.
If the time is greater than 1.8 seconds, the rod shall be declared inoperable.
I.
Monitor Inoperability Requirements l
1.
If the rod bank insertion limit monitor is inoperable, or if the rod l
position deviation monitor is inoperable, individual rod positions shall be logged once per shift, after a load change greater than 10 percent of rated power, and af ter 30 inches or more of red motion.
2.
If both the rod position deviation monitor and one or both of the l
quadrant power tilt monitors are inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, I
the nuclear overpower trip shall be reset to 93% of rated power in addition to the increased surveillance requirements.
l l
Prairie Island Unit 1 - Amendment No. 44 Prairie Island Onit 2 - Amendment No. 38
- ~,,
--s--
r
i TS. 3.10-8 3.
If one or both of the quadrant power tilt monitors is inoperable, l
individual upper and lower excore detector calibrated outputs and the calculated power tilt shall 'se logged every two hours af ter a load change greater than 10% of rateo power.
J.
DN3 Parameters l
The following DNS related parameters limits shall be maintained during power operation:
a.
Reactor Coolant System Tavg <564 F b.
Pressurizer Pressure 2,2220 psia
- c.
Reactor Coolant Flow 2,178,000gpm l
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 using normal shutdown procedures.
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 />.
Compliance with c. is demonstrated by verifying that the parameter is within its limit af ter each refueling cycle.
Bases Throughout the 3.1, fechnical Specifications, the terms " rod (s)" and "RCCA(s)"
are synonomous.
Shutdown Reactivity Trip shutdown reactivity is provided consistent with plant safety analyses as sump tions. One percent shutdown is adequate except for the steam break analysis, which requires more shutdown 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.
Power Distribution Control 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
>1.30 during normal operation and in short term transients, and (b)
Timiting the fission gas release, fuel pellet temperature and cladding
- Limit not applicable during either a THERMAL POWER ramp increase in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10%) RATED THERMAL POWER.
Prairie Island Unit'l - Amendment No. J$, J9, 44 Prairie Island Unit 2 - Amendment No. Jg, JJ, 38
TS.3.10-9 mechanical properties to within assumed design criteria.
In addition, limiting the peak linear power density during Condition I events y o-vides assurance that the initial conditions assumed fog the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not axceeded.
During opegation,Nthe plant staff compares the sensured hot channel factors, F" and F, (described later) to the limit determined in the traksient N d LOCA analyses. he limiting F (Z) includes measurement, engineering, and calculational uncertainties. The kerms on the right side of the equations in section 3.10.B.1 represent the analytical limits.
Those terms on the left side represcat the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.
F (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the imum local heat flux on the surf ace of a fuel rod at core elevation Z divided by the average -fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
The maximum value of F (Z) is 2.21/P g
for the Prairie Island reactors.
This value is restricted fdrther by the K(Z) and BU(E ) functions described below. The product of these three factors is 3
F (Z).
q The K(Z) function shown in Figure TS.3.10-5 is a normalized function that limits F (Z) axially for three reasons.
The K(Z) specified for the l
9 lowest six (6) feet of the core is based on large break LOCA analyses.
Above this region the K(Z) value is based on DNBR requirements since the minimum DNBR would be expected in this region of the core, based on power, pressure, and temperature.
The K(Z) value in the uppermost region of the core is based on the small break LOCA analyses.
F (Z)
O in the uppermost region is limited to reduce the PCT expected during a small break LOCA s *nce this region of the core is expected to uncover temporarily for some small break LOCA's.
The BU(E.) function shown in Figure TS.3.10-7 is a normalized function limits F (Z) based on exposure dependent analyses for the ENC fuel.
that 0
These analyses consider pin internal pressure uncertainties, fuel swelling, rupture pressures, and flow blockage.
N F is the measured Nuclear Hot Channel Factor, defined as the l
mkximum local neutron flux in the core divided by the average neutron flux in the core.
l Vf.Z) is an axjally dependent function applied to the equilibrium measured F" to bound r's that could be measured at non-equilibrium conditions.
This function is based on power distribution control analyses that eval-9 uated the ef fect of burnable poisons, rod position, axial ef fects, and xenon worth.
I e
l F~, Engineering Heat Flux Hot Channel Factor, is defined as the I
aklowanceonheat flux required for manufacturing tolerances. The engineering f actor allows for local variations in enrichment, pellet density and disseter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined s tatistically the net ef fect is a factor of 1.03 to be applied to fuel rod surface heat flux.
l Prairie Islar' dnit 1 - Amendment No. 35, 44 l
Prairie Island Unit 2 - Amendment No. 29, 38 i
. - - - - ~
TS.3.10-10 The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those measurements to establish the assembly local power distribution.
F, Nuclear Enthalpv Rise Hot Channel Factor, is defined as the ratio ok the integral of d.near power algng the rod with the highest integrated power to the average rod power.
F" is l
is used as such in the DNB 'calculaNons. based on an integral and 1,ocal heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizational (x y) power shapes throughout the core. Bus the horizontal power' shape at the pgint of maximum heat flux is not necessarily directly related to Fg.
In the specified limit of F^N there is an 8 percent allowance for un-H certainties which means thaE normal operation of the core is expected to y
result in Fg <1.55/1.08.
De logic behind the larger uncertainty in this case is that:
l (a) abnormal perturbations in the radial power shape (e.g. rog misalignment) affect F H, in a st cases without necessarily affecting F,
q (b) the operator has a direct influence on F through movement of rods, and cag limit it tothedesiredvalue,wkilehehasnodirectcontrol over F
- and, g
(c) an error in the predictions for radial power shape, which may be dgtected during startup physics tests can be compengated for in F" by tighter axial control, but compensation for F is less rhadilyavailable.
g N
When a measur ment of F is taken, experimental error must be allowed for as d 4 perceN is the appropriate allowance for a full core I
map taken with the movable incore detector flux mapping system.
Measurements of the hot chanael f actors are required as part of startup physics tests, at least once each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot. channel factors, he incere map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.
De periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise af fect these bases.
For normal operation, it is not necessary to nessure these quantities.
Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met; these conditions are as follows:
1.
Control rods in a single tank move together with no individual rod insertion differing by more than 15 Prairie Island Unit 1 - Amendment No. 35, 44 Prairie Island Unit 2 - Amendment No. 29, 38
TS.3.10-11 l
inches from the bank demand position. An accidental misalignment limit 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 Te:hnical Specification 3.10.
3.
The control bank insertion limits are not violated.
i 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 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.
N N
The permitted relaxation in F and F allows for radial power J1H shape changes with rod insertton to tke insertion limits.
i It has been determined that provided the above conditions 1 through 4 are obsegved, l
l these hot channel f actor limits are met.
In specification 3.10, F l
arbitrarily limited for P < 0.5 (except for low power physics testk)is i
The procedures for axial power distribution control referred to above I
are designed to minimize the ef fects of 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 ( 4LI) and a reference value which corresponds to the full power equilibrium value of Axial Offset i
(Axial Offset = diI/ 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 TS.3.0(Z) upper bound envelope of 2.21/P times Figures TS.3.10-5 and the F IO-7 is not exceeded and xenon distributions are not developed 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.
i 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 l
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 power at which the core was operating is the full power value of 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 allowances for excore detector error are necessary and indicated deviation of 15 percent AI are permitted from the indicated reference value.
Figure TS.3.10-6 shows the allowed deviation from l
t e target fiv difference as the function of thermal power.
h Prairie Island Unit 1 - Amendment No. 35,44 Prairie Island Unit 2 - Amendment No. 29,38
t l
TS.3.10-12 The alarms provided are derived from the plant process computer which determines the one minute averages of the operable excore detector outputs l
to monitor indicated axial flux difference in the reactor core and alerts the operator when indicated axial flux difference alarm conditions exist.
Two types of alarm messages are output.
Above a preset (90%) power level, an alarm message is output immediately upon determining a delta flux (as determined from two operable excore channels) exceeding a preset band about a target delta flux value.
Below this preset power level, an alarm message is output if the indicated axial flux difference (as determined from two-operable excore channels) exceeded its allowable limits for a preset cum-ulative (usually 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) amount of time in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For periods during which the alarm on flux difference in inoperable, manual surveillance
(
will be utilized to provide adequate warning of significant variations in i
expected flux differences.
However, every attempt should be made to restore the alarm to an operable condition as soon as possible. Any deviations from the target band during manual logging would be treated as deviations during i
the entire preceding logging interval and appropriate actions would be taken.
This action is necessary to satisfy NRC requirements; however, more frequent readings may be logged to minimize the penalty associated with a deviation from the target band to justify continued operation at the current power.
The time that deviations from the target band occur are normally accumulated by the computer above 15% power.
Below 15% the probability of exceeding the allowable limits becomes increasingly smaller as it becomes theoretically impossible to l
deviate from the target band.
Between 15-50% power the deviations are more significant and are accumulated at 1/2 of their actual time. Above 50% the deviations are most significant and their time is accumulated on a one for one time basis.
Strict control of the flux difference (and rod position) is not as necessary during part power operation because xenon distribution control at part power is les s significant than control at full power. Allowance has been made in predicting the heat flux peaking factors for less strict t
control at part power.
Strict control of the flux difference is not possible during certain physics tests or during required, periodic, excore calibrations which require larger flux differences than permitted.
There fore,
the specifications on power distribution control are not applied during physics tests or excore calibrations; this is acceptable due to the low probability of a significant accident occurring during these operations.
I In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the targe t band when the reduced power level is reached. This does not necessarily affect the xenon distribution suf ficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band, however to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band.
This ensures that the resulting xenon distributions are not significantly different from those Prairie Island Unit 1 - Amendment No. J$,44 Prairie Island Unit 2 - Amendment No. J9,38
~
l TS.3.10-13 resulting from operation within the target band.
The consequences of being outside the +5% target band but within the Figure TS.3.10-6 limit for power levels between 50% and 90% has been evaluated and determined to result in acceptable F (Z) values.
Therefore, while the deviation q
exists the power level is limited to 90 percent or lower depending on the indicated arial flux difference.
In all cases the +5 percent target band is the Limiting Condition for Operation. Only when the target band is violated dc the limits under Figure TS.3.10-6 apply.
l If, for any reason, the indicated axial flux difference is not controlled widsin the +5 percent band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of some accidents.
As 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 I
the full length control rods to produce the required indicated flux difference.
For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 by an automatic protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.
Quadrant Power Tilt Limits Quadrant power tilt limits are based on the following considerations.
Fre-quent 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 core limits protected per Specification 3.10.E.
A quadrant tilt by some other means (x y xenon tran-sient, etc.) would not appear instantaneously, but would build up over several hours and the quadrant tilt limits are set to protect against this situation.
They also serve as a backup protection against the dropped or misaligned rod.
Operational experience shows that normal power tilts are less than 1.01.
Thus, sufficient time is available to recognize the presence of a tilt and correct the cause before a severe tilt could build up.
During start-up and power escalation, however, a large tilt could be initiated.
Therefore, the Technical Specification has been written so as to prevent escalation above 50 percent power if a large tilt is present.
(
l Prairie Island Un.it 1 - Amendment No. J8,44 I
Prairie Island Unit 2 - Amendment No. J9,38 l
TS.3.10-14 The numerical limits are set to be commensurate with design and safety limits for DNB protection and 14.near heat generation rate as described be low.
~
percentage quadrant power t!.lt of 22 at which remedial and corrective ac. ion is required has been set so as to provide DNB and linear heat gen-eration rate protection with x y power tilts.
Analyses have shown that percentage increases in the x y power peaking f actor are less than or equal to twice the increase in the indicated quadrant power tilt.
An increase in F is not likely to occur with tilts up to 3% because l
misalignedcontr81rodspr ucing such tilts do not extend to the unrodded i
l plane, where the maximum F occurs.
Therefore, a limiting power tilt of 3 percent can be tolerated.
Howeve r,
a measurement uncertainty is associated with the indicated quadrant powe r tilt.
Thus, allowing for a low measurement of power tilt, the action level of indicated tilt has been set at 2 percent. An alarm is set to alert
[
the operator to an indicated tilt of 2 percent or greater for which action is required. To avoid unnecessary power changes, the operator is allowed two hours in which to verify the actual tilt with in-core mappings or to determine l and correct the cause of the tilt.
Should this action not be taken, the margin for uncertainty in F is reinstated by reducing the power by 2 percent for each percent ok tilt above 1.0, in accordance with the relationship described above, or as required by the restriction on peaking factors.
The upper limit on the quadrant tilt at which hot shutdown is required has been set so as to provide protection against excessive linear heat generation rate.
The ratio of overgower to normal operation is approximately l l
1.15.
Since the x y component of F is bounded by the above described j
relation with indicated quadrant tikt, the overpower linear heat generation I
rate can be s. voided if the indicated tilt is restricted below 7 percent.
l l
l l
l l
l Prairie Island Unit 1 - Amendment No. Jf,44 Prairie Island Unit 2 - Amendment No.19,38
TS.3.10-15 Rod Insertion L_imits l
Rod insertion limits are used to assure adequate trip reactivity, to assure meeting power distribution limits, and to Ibnit the consequences of a hypothetical rod ejection accident.
The available control rod re-activi'_y (or excess beyond needs) decreases with decreasing boron con-centration.
The negative reactivity required to reduce the cote power'1evel from full power to zero power is largest when the boron concentration is low since the power defect increases with core burnup.
The intent of the test to measure control rod worth and shutdown margin (Specification 3.10 D.) is to measure the worth of all rods less the worth of the the most. reactive rod. The measurement would be anticipated l
a= part of the initial startup program and infrequently over the life of the plaat, to be associated primarily with determinations of special intarest such as end of life cooldown, or startup of fuel cycles which deviate from normal equilibrium conditions in terms of fuel loading patterns and anticipated control bank worths.
These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical t
basis.
l l
An evaluation has been made of anticipated transients and postulated
~accidents, assuming that they occur during the portion of this test when the reactor is critical with all but one full-length co*. trol rod i
fully inserted.
Further, the withdrawn full-length rod is 4.ssumed l
not to trip.
As a result of this evaluation, it has been oetermined that for a m line break upstream of the flow restrictor, the possibility of I
core D.
. ;ts.
However, even if core damage does result, any core fission p.oduct release would be low because of the low fission product inventory during initial startup physics testing; and further, would be contained within the reactor coolant system.
Thus, for the initial startup physics tests, this test will not.
endanger the health and safety of the public esco in the event of highly improbable accidents coupled with the failure of the withdrawn control rod to trip.
To perform this test late: in life is equally valuable, as stated above. Therefore, this spe:ification has been written to further minimize the likelihood of any hypothesized event dur' ; the performance of these tests later in life. This is accomplished by limiting to two hours per year the time the reactor can be in this type of configuration, and requiring that a rod drop test is performed on the rod to be measured prior to performance of test.
l l
Pi airie Island Unit 1 - Amendment No. 44 Prairie Island Unit 2 - Amendment No. 38 i
1 I
TS.3.10-16 Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.
Rod Misalignment Limitation i
Rod misalignment requirements are specified to ensure that power distribu-tions more severe than those assumed in the safety analyses do not occur.
Inoperable Red Position Indicator Channels The rod position indicator channel is sufficiently accurate to detect a rod
+7 inches away from its demand position. A misalignment less than 15 inches does not lead to over-limit power peaking f actors.
If the rod position indicator channel is not operable, the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or core thermocouples, and/or movable incore detectors, will be used to verify power distribution symmetry.
These indirect measurements do not have the same resolution if the bank is near either end of the core, i
because a 15-inch misalignment would have no effect on power distributions.
There ere, it is necessary to apply the indirect checks following significant 8
rod notion.
Inoperable Rod Limitations l
One inopereble control rod is acceptable provided that the power distribution limits are met, trip shutdown capability is availabi.e, and provided the l
potential hypothetical ejection of the inoper:ble rod is not worse than the cases analyzed in the safety analysis report. The rod ejection accident for an isolated fully-inserted rod will be worse if the residence time of the rod is long enough to cause significant non-uniform fuel depletion. The four-week period is short compared with the time interval required to achieve a significant non-uniform fuel depletion.
Rod Drop Time l
The required drop time to dashpot entry is consistent with the safety analysis.
I Monitor Inoperability Requirements If either the rod bank insertion limit monitor or rod position deviation monitor are inoperable, additional surveillance is required to ensure adequate shutdown me gin is maintained.
1 l
l Prairie Island Unit 1 - Amendment No.19, 44 Prairie Island Unit 2 - Amendment No.13, 38
- a d>
TS.3.10-17 If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that adequate core prctection is provided in the event that unsatis-factory conditions arise that could af fect radial power distribution.
Increased surveillance is required, if the quadrant power tilt monitors are in-operable and a load change occur e, in order to confirm satisfactory power distribution behavior. The autynatic alarm functions related to quadrant power tilt must be considered incapable of alerting the operator to unsatis '
factory power distribution conditions.
DN3 Parameters The RCS flow rate, T and Pressurizer Pressure requirements are based on transient analyses afs$m,ptions.
The 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 specification value indicated, ahutdown is required to investigate acequacy of core cooling during operation.
For fuel regions with high burnups, the depletion of fissile nuclides and build-up of fission products greatly rgduces power production capability.
These combined burnup ef fects reduce F sufficiently to cover residual rod bow AH penalties beyond a region average burnup of 40,000 MWD /MTU.
l I
l Prairie Island Unit 1 - Amendment No. 44 Prairie Island Unit 2 - Amendment No. 38
FIGURE TS.3.10-7 oukk Normalized Exp'osure Dependent Function BU(E ) for Exxon Nuclear Coriany Fuel 11 m a w
e-+
w m.
aa C C
_ g.
...g L
N ~*
1.00 1-K
- 8
' %L_
gg
.(01.00?
7 g
- 5_
3 gg
_f _w a a gg
.... _lil : T~. 1
~_ (41.85,,878) T_-
i y a M et gg nm BU(E )
J y y
._~
- -=
i U
~~~'
.L 0.0.
0 4
8 12 16 20 24 28 32 35 40.
' 54.__ _ _.
_, _ y y
Burnup (CUD /t!TU) e
-