ML20215M149
| ML20215M149 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/04/1987 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215M141 | List: |
| References | |
| NUDOCS 8705130100 | |
| Download: ML20215M149 (16) | |
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UNITED STATES l* y ,#r p, 8
NUCLEAR REGULATORY COMMISSION W ASHING TON, D. C. 20$55 5-
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OMAHA PUBLIC POWER DISTRICT l DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE ,
Amendment No.109 License No. OPR-40 i
- 1. The Nuclear Regulatory Commission (the Commission) has found that: t A. The application for amendment by the Omaha Public Power Ofstrict (the licensee) dated January 22, 1987, as supplemented February 13 and 24, 1987, complies with the standards and requirements of the .
Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 ruit.s and regulations of the !
Commission; 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 compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common 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. L t
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- l B705130100 870504 l L
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- 2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.
DPR-40 is hereby amended to read as follows:
B. Technical Specifications The Technical 5)ecifications contained in Appendix A, as revised tirough Amendment No.109, are hereby incorporated in the license. The licensee shall operata the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION v, e $' '
I lJoseA.Calvo, Director Project Directorate - IV l Division of Reactor Pro;'ects - 111, IV, V and Special Pro;ects
Attachment:
Changes to the Technical l Specifications Date of Issuance: May 4, 1987 l
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ATTACHMENT TO LICENSE AMENDMENT NO.109 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 5 5 6 6 Figure 1-3 Figure 1-3 1-9 -
1-9 2-48 2-48 2-50c 2-50c Figure 2-6 Figure 2-6 Figure 2-8 Figure 2-8 Figure 2-9 Figure 2-9 2-57a 2-57a 2-57c 2-57c 3-63a 3-63a 4-3 4-3
DEFINITIONS MISCELLANEOUS DEFINITIONS Operable - Operability ,
A system, subsystem, train, component or device shall be OPERABLE or have .
OPERABILITY when it is capable of performing its specified function (s). [
Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, nonnal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related i supportfunction(s),
f In Operation J
A system or component is in operation if it is performing its design function.
j:
i CEA's ,
All full length shutdown and regulating control rods.
Non-trippable (NT) CEA'_s CEA's which are non-trippable. '
gntainmentIntear_i_ty Containment integrity is defined to exist when all of the following are met:
- (1) All nonautomatic containment isolation valves which are not required to be open during accident conditions and blind flanges are closed.
(2) The equipment hatch is properly closed and scaled.
(3) At least one door in the personnel air lock is properly closed and se61cd.
! (4) All automatic containment isolation valves are operable or locked closed ,
I (or isolated by locked closed valves or blind flanges as permitted by limitingconditionforoperation).
(5) The uncontrolled containment leakage satisfies Specification 3.5. l I
Amendment No. 51,109 5
o a DEFIN!!!0NS g Core Alteration 1
I
' The movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of Core Alteration shall not preclude completion of movement of a component to a safe, conservative position.
Equivalent Full Power Day (EFPD)
The time interval during power operation when the heat generated by the ;
reactor is equivalent to reactor operation at 100% of rated power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
, Shutdown Margin t
' Shutdown Margin shall be the amount of reactivity by which:
(1) the reactor is suberiticall or 1
- (2) the instantaneous amount of reactivity by which the reactor would be subcritical from its present condition assuming:
3
- a. All known trippable full length control element assemblies (shut- '
I down and regulating) are fully inserted except for the single l assembly of highest reactivity worth which is assumed to be fully withdrawn, and i,
- b. No change in non-trippable control element assembly position, j Axial Shape _Index_
The external Axial Shape Index (iE ) is the power level detected by the lower I excoronuclearinstrumentdetectors(L)lessthepowerleveldetectedbythe upper excore nuclear instrument detectors (U) divided by the sum of these power Icvels. The internal Axial Shape Index (Y 1 ) used for the trip and protrip signals in the reactor protection system is the above value (YE )
! modified by the shape annealing factor, SAF, and a constant, D, to deter-
! mine the true core axial power distribution for that channel.
l YE*l'S TI = SAF x Yg+B L+U '
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Amendment No. D .H. 109 6 ,
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590
.u. 580 570 n N \ \
is G <
N \ \\ 1 l
X 2400 psia I =:!
A N N \ 2250 psia _
i 550 2075 psia g
8 N
\
u j 530 C 1750 psia-I i 520
- 60 70 80 90 100 110 120
CORE POWER (% OF RATED POWER)
- Pyg = 22 PF(8) Ai(Y)0+22.iT 3n -12703 ;
PF (8) = 1.0 02100%
= .000 0 + 1.0 50%<8<100%
= 1.4 0550% i A1 (Y) = .5Y 1 + 1.125 Y 3 s .25 ,
= .5Y 3 + .075 Y 3 > .25 j Thereal Margin / Lc'.I Pressure LSSS Osaha Putlic Power District figure ;
! 4PeroOperation FortCalhour.StationllnitNo.i 1-3 l h*nd=nt M. p.79,n.n.n.n,in . !
1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS _
1.3 Limiting Safety System Settings, Reactor Protective System (Continued)
(7) Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down simultaneously with the initiation of the safety injection system. The setting of this trip is identical to that of the containment high pressure signal which indicates safety injec-tion system operation.
(8) Axial Power Distribution - The axial power trip is provided to ensure that excessive axial peaking will not cause fuel damage.
The Axial Shape Index is determined from the axially split excore detectors. The set point functions, shown in Figure 1-2 ensure that neither a DNBR of less than 1.18 nor a maximum l linear heat rate of more than 22 kW/f t (deposited in the fuel) will exist as a consequence of axial power ma1 distributions.
Allowances have been made for instrumentation inaccuracies and uncertainties associated with the excore symmetric offset -
incore axial peaking relationship.
(9) Steam _ Generator Dif_ferential Pressure - The Asymetric Steam GeneratorTransientProtectionTripFunction(ASGTPTF) utilizes a trip on steam generator differential pressure to ensure that neither a DNBR of less than 1.18 nor a peak linear heat rate of l more than 22 kW/f t occurs as a result of the loss of load to one steam generator.
(10) Physics Testing at L H vB s less than 10"gw'Tof Power - During rated power, thephysics testing tests may at power require that the reactor be critical. For these tests only the low reactor coolant flow nd thermal margin / low pressure trips may be bypassed below 10* % of rated power. Written test arocedures which are approved by the Plant Review Comittee will se in effqct during these tests. At reactor power levels less than 10*'% of rated power the low reactor coolant flow and the thermal margin / low pressure trips are not required to prevent fuel element thermal limits being exceeded. Both of these trips are bypassed using the same bypass switch. The low steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown if a steam line break were to occur during the tests.
References USAR Section 14.1 USAR, Section 7.2.3.3 USAR, Section 7.2.3.2 USAR, Section 3.6.6 USAR, Section 14.6.2.2,14.6.4
- USAR, Section 14.7 USAR, Section 7.2.3.1 USAR, Section 3.6 USAR, Section 14.10 Amendment No. 7.)/,79,77,f//,10g I9
2.0 LIMITING C0f(DIT10NS FOR OPERATION 2.10 Reactor Core 2.10.1 Minimum conditions for Criticality App _licability Applies to the status of the reactor coolant system during reactor criticality.
Objective To prevent unanticipated power excursions of an unsafe magnitude.
Specifications (1) Except during physics tests at less than 10-l% power, the reactor shall not be made critical if the average reactor coolant temperature is below 515'F.
(2) In no case shall the reactor be made critical if the reactor coolant temperature is below NDTT +120'F.
(3) No more than one CEA at a time in a regulating or non-trippable l group shall be exercised or withdrawn until after a steam bubble and normal water level as given in operating procedures are established in the pressurizer.
(4) Reactor coolant boron concentration shall not be reduced below that required for the Hot Shutdown Condition until af ter a steam bubble and normal water level are established in the pressurizer.
Basis At the beginning of each fuel cycle, the moderator temperature co-efficient is expected to be slightly negative at operating temper-atures with all CEA's withdrawn. However, variations in cycle core loading and the uncertainty of the calculation are such that it is possible that a slightly positive coefficient could exist.
The moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature. It is, therefore, prudent to restrict the operation of the reactor when reactor coolant temperatures are less than 515'f.
Amendment No M.109 2-48
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l 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.2 feactivity Control Systems and Core Physics Parameters Limits (Continued)
Non-trippable CEA Position During Power Operatio_n_ l (5)
All non-trippable CEA's (NTCEA) shall be withdrawn to at least ,
i 114 inches (actual position). If one or more NTCEA's becomes :
misalignedfromotherNTCEA'sbymorethan12 inches (actual position) either:
I
- a. Restore the NTCEA to within the specified alignment requirements within one hour, or b'. Be in at least hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
! (6) Shutdown CEA Insertion Limit During Power Operation All shutdown CEA's shall be withdrawn to at least 114 inches as a condition for reactor criticality, or with one or more shutdown CEA's inserted to more than 114 inches withdrawn, except for surveillance testing, within one hour, either:
- a. Withdraw the CEA's to at least 114 inches, or
- b. Declare the CEA's inoperable and apply Specification 2.10. 2(4 ) .
(7) Regulating CEA Insertion Limits During Hot Standby and Power Operatlon The regulating CEA groups shall be limited to the insertion sequence and to the insertion limits shown on Figure 2-4 except during CEA exercises above 114 inches. With all CEA's operable.
CEA insertion beyond the Long Term Insertion Limits are restricted l
to:
- 1. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval,
- 3. 14 EFPD per calendar year.
r
- a. When the regulating CEA groups are inserted beyond the l Transient insertion Limits within two hours. cither:
(i) Restore the regulating CEA groups to above the Transient Insertion Limits, or (ii) Reduce reactor power to the allowed power of Figure 2 4 which pomits continued operation ;
above the Transient Insertion Limit using the existing CEA group position. . !
Amendment No M. 109 2 50c i
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! LimitingConditionforOperationfor ExcoreMonitoringofLHR FortCalhounStation-llnitNo,i 2-6 [
! Amenenent No. p.79. AZ.#3.#7.79.77.pl,109 ;
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% CORE LENGTH FROM BOTTOM OmahaPublicPowerDistrict figure FluxPeakingAugmentation fortCalhounStation-UnitNo.I 2-8 factors Amr
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FjANDF xy OmahaPublicPowerDistrict figure Fj fx!, and Core Power 2-9 Limitations FortCalhounStation-UnitNo.1
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2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)
(ii) Be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
(2) Total Integrated Radial Peaking Factor .
ThecalculatedvalueofF[definedbyF[=F R (1+Tq ) shall be limited to 1 1.80. FR is determined from a power distribu-tion map with no non-trippable CEA's inserted and with all l full length CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, T q, is the measured value of Tq at the time FR is determined.
With FR > 1.80 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:
(a) ReducepowertobringpowerandFfwithinthelimits of Figure 2-9, withdraw the full length CEA's to or beyond the Long Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's, l or (b) Be in at least hot standby.
(3) Total Planar Radial Peaking Factor The calculated value of FxyT defined as FxyT = Fxy (1+Tq) shall be limited to 1 1.85. Fxy shall be determined from a power distribution map with no non-trippable CEA's inserted and with l all full length CEA's at or above the Long Tem Steady State Insertion Limit for the existing Reactor Coolant Pump combina-tion. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. The azimuthal tilt.
Tq , is the measured value of Tq at the time Fxy is detemined.
With FxyT > 1.85 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:
(a) Reduce power to bring power and FxyT to within the limits of Figure 2-9, withdraw the full length CEA's to or beyond the Long Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's, or l (b) Be in at least hot standby.
Amendment No. 32,43,47,70,77,92, 109 2-57a
t, .
. 2.0 LIMITING CONDITIONS F0'R OPERATION 2.10 ReactorCore(Continued) 2.10.4 Pcwer Distribution Limits (Continued)
(5) DNBR Margin During Power Operation Above 15% of Rated Power (a) The following DNB related parameters shall be maintained within the limits shown:
(1) Cold Leg Temperature 1 54 5'F* [
(ii) Pressurizer Pressure 1 2075 psia *
(iii) Reactor Coolant Flow 1197,000 gpm**
(iv) Axial Shape Index, Y y 1 Figure 2-7***
(b) With any of the above parameters exceeding the 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 power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Basis Linear Heat Rate The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Excore Detec-tor Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring cf the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable limits of Figure 2-6 as adjusted by Specification 2.10.4(1)(c) for the allowed linear heat rate of Figure 2-5, RC Pump configuration, and Fx T of Figure 2-9.
In conjunction with the use of the excore monitoring system a5d in establishing the axial shape index limits, the following assumptions are made: (1) the CEA insertion limits of Specification 2.10.2(6) and long term insertion limits of Specification 2.10.2.(7) are satisfied, (2) the flux peaking augmentation factors are as shown in Figure 2-8, and (3) the total planar radial peaking factor does not exceed the limits of Specification 2.10.4(3).
- Limit not applicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step of greater than 10%
of rated thermal power.
- This number is an actual limit and corresponds to an indicated flow rate of 202,500 gpm. All other values in this listing are indicated values and include an allowance for measurement uncertainty (e.g., 545 F, indicated, allows for an actual Tc of 547 F).
Amendment No. 32,43,57,79,77, pp, 109 2-57c
O. .
3.0 SURVEILLANCE REQUIREMENTS 3.10 Reactor Core Parameters (Continued)
(2) Moderator Temperature Coefficient The MTC shall be determined at the following frequencies and power conditions during each fuel cycle:
- 1. Prior to initial operation above 5% of rated power, after each fuel loading.
- 2. At any power level within 500 MWD /T of initial operation after each refueling.
- 3. At any power level within + 14 EFPD of reaching a l rated power equilibrium boron concentration of 300 ppm.
(3) Regulating CEA Insertion Limits
- a. The position of each regulating CEA group shall be determined to be above the Transient Insertion Limits at least once per shift,
- b. The accumulated times during which the regulating CEA groups are inserted beyond the Steady State Insertion Limits but above the Transient Insertion Limits shall be detennined once per day.
(4) Linear Heat Rate Monitoring Systems
- a. The incore detector monitoring system may be used for monitoring the core power distribution provided that at least once per 31 days of accumulated power operation the incore detector alarms generated by the plant computer are verified to be valid and satisfy the requirements of the core distribution map. s
- b. The excore detector monitoring system may be used for monitoring the core power distribution by
1 1. Verifying at least once per 31 days of accumulated power operation that the axial shape index, YI , monitoring limit setpoints are maintained within the allowable limits of Figure 2-6, as adjusted by Specification 2.10.4(1).
(5) Total Integrated and Total Planar Radial Peaking Factors (FRT and FxyT) f FRT and FxyT shall be detennir:ed to be within the limits of l
Specification 2.10.4 at the following intervals:
- a. After each refueling and prior to operation above 70 percent l of rated power. ,
- b. At least once per 31 EFPD's of accumulated power operation.
Amendment No. 32,109 3-63a
o ..
i-4.0 DESIGN FEATURES 4.3 Nuclear Steam Supply System (Continued) 4.3.1 Reactor Coolant System (Continued)
The reactor coolant system is designed and constructed in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Vessels including all addenda through the winter of 1967 and the Code for Pressure Piping USAS B31.1.
The reactor coolant system is designed for a pressure of 2500 psia and a temperature of 650 F except for the pressurizer which has a design temperature of 700 F. The volume of the reactor coolant system is approximately 6,616 cubic feet.
4.3.2 Reactor Core and Control The reactor core shall approximate a right circular cylinder with an equivalent diameter of 106.5 inches and an active height of 128 inches. The reactor core shall normally consist of Zircaloy-4 clad fuel rods containing slightly enriched. uranium in the form of sintered U02 Pellets. The fuel rods shall normally be grouped into 133 assemblies.
The core excess reactivity shall be controlled by a combination of boric acid chemical shim, control element assemblies, and mechanically fixed boron rods where required. Forty-nine control element assemblies are distributed throughout the core as shown in Figure 3.4-4 of the USAR; four of the CEA's are full length non-trippable CEA s.
4.3.3 Emergency Core Cooling Emergency core cooling is provided by the Safety Injection System which' consists of various subsystems, each with internal redundancy. Included in the Safety Injection Systen are four safety injection tanks, three high-pressure and two low-pressure safety injection pumps, a safety injection and refueling water storage tank, and interconnecting piping as shown in USAR Section 6. l l
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Amendment No. E,20,50,109 4-3 l
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