ML20138Q160

Amend 92 to License DPR-40,revising Tech Specs to Support Operation of Unit at full-rated Power During Cycle 10

ML20138Q160
Person / Time
Site:	Fort Calhoun
Issue date:	11/29/1985
From:	Thadani A Office of Nuclear Reactor Regulation
To:
Shared Package
ML20138Q153	List: ML20138Q149 ML20138Q160 ML20138Q172
References
NUDOCS 8512270075
Download: ML20138Q160 (17)

Text

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UNITED STATES

~ *t, NUCLEAR REGULATORY COMMISSION g j

• WASHING TON, D. C. 20555 e

%,...../

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285, FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 92 License No. DPR-40

1. The Nuclear Regulatory Comission (the Comission) has found that:

I A. The application for amendment by the Omaha Public Power District

.I (the licensee) dated September 6,1985 as supplemented by letter

dated September 3, 1985, complies with the standards and requirements I 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; h C. There is reasonable assurance (1) that the activities authorized J 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 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.

e 8512270075 851129 i PDR ADOCK 05000285 '

P PDR 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 Specifications contained in Appendix A, as revised through Amendment No. 92 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR T}lE NUCLEAR REGULATORY COPfilSSION

.. r i Asho C. Thadani, Director

! PWR roject Directorate #8 Division of PWR Licensing-B i

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Attachment:

Changes to the Technical l Specifications Date of Issuance: November 29, 1985 L

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ATTACHMENT TO LICENSE AMENDMENT NO. 92

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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.

l Remove Pages Insert Pages

! 1-2 1-2 1-7 1-7 1-8 1-8

,' 1-9 1-9 1-10 1-10 2-2b 2-2b 2-57a 2-57a 2-57b 2-57b l 2-57c 2-57c i 3-63b 3-63b t

Figure 1-1 Figure 1-1 I Figure 1-3 Figure 1-3 l Figure 2-6 Figure 2-6 Figure 2-9 Figure 2-9 5

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1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.1 Safety Limits - Reactor Core (Continued) would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin Io-DNB. The minimum value of the DNBR during steady state opera-tion, normal operational transients, and anticipated tran-sients is limited to 1.18. A DNBR of 1.18 corresponds to a l 95% probability at a 95% confidence level that DNB will not occur, which is considered an appropriate margin to DNB for all operating conditions.(1)

The curves of Figure 1-1 represent the loci of points of re-actor thermal power (either neutron flux instruments or AT in-struments), reactor coolant system pressure, and cold leg temperature for which the DNBR is 1.18. The area of safe opera-tion is below these lines. l The reactor core safety limits are based on radia5 peaks limit-ed by the CEA insertion limits in Section 2-10 and axial shapes within the axial power distribution trip limits in Figure 1-2 and a total unrodded planar radial peak of 1.85. l The LSSS in Figure 1-3 is based on the assumption that the un-rodded integrated total radial peak (F ) is 1.80. This peak- 1 ing factor is slightly higher (more co servative) than the i maximum predicted unrodded total radial peak during core life, excluding measurement uncertainty.

Flow maldistribution effects for operation under less than full reactor coolant flow have been evaluated via model tests.(2) The flow model data established the maldistribution factors and hot channel inlet temperature for the thermal analyses that were used to establish the safe operating enve-lopes presented in Figure 1-1. The reactor protective system

• is designed to prevent any anticipated combination of tran-sient conditions for reactor coolant system temperature, pres-sure, and thermal power level that would result in a DNBR of

.i less than 1.18.(3) k-References

() USAR, Section 3.6.7

() USAR, Section 1.4.6 j () USAR, Section 3.6.2 l

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1 Amendment No. 8,32,6 ,47, 1-2 l 79,77,92

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. . l 1.0 1.3 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS  ;

Limiting Safety System Settinas, Reactor Protective Syste5F' (Continued)

During reactor operation at power levels below 19.1%

rated power, a reactor trip will occur in the event of a reactivity excursion that results in a power increase up to the lower fixed set point of the VHPT circuit of 19.1% of rated power.(3) During normal power increases be-low 19.1% reactor trip would be initiated at 19.1% of rated power unless the set point is manually adjusted.

(2) Low Reactor Coolant Flow - A react'or trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly.

Flow in each of the four coolant loops is determined from a measurement of pressure drop from inlet to out-let of the steam generators. The total flow through the reactor core is measured by summing the loop pres-sure drops across the steam generators and correlating this pressure sum with the pump calibration flow Curves.

The percent of normal core flow is as follows:(6)

~

4 Pumps 100%

During' four-pump operation, the low flow trip setting of 95% insures that the reactor cannot operate when the flow rateinstrument sidering is less thanerrors.

93% of the nominal value con-1 -

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l' Amendment No. 7,32,79,77, 92 1-7

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1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

• 1.3 Limiting Safety System Settings, Reactor Protective System (Continued) '

(3) High Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system overpressure (Specification '

2.1.6). In the event of loss of load without reactor

' rip, the temperature and pressure of the reactor t

coolant system would increase due to the reduction in the heat removed from the coolant via the steam gener-ators. The power-operated relief valves are set to operate concurrently with the high pressurizer pressure reactor trip. This setting is 100 psi below the nominal safety valve setting (2500 psia) to avoid un-necessary operation of the safety valves. This setting is consistent with the trip point assumed in the ac-cidentanalysis.(1)

(4) Thermal Margin / Low Pressure Trip - The thermal margin /

low pressure trip is provided to prevent operation when the DNBR is less than 1.18, including allowance for measurement error. The thermal and hydraulic limits j shown on Figure 1-3 define the limiting values of re-actor coolant pressure, reactor inlet temperature, axial shape index, and reactor power level which ensure that the thermal criteria (s) are not exceeded. The low set point of a 1750 psia trips the reactor in the unliFVly event of a loss-of-coolant accident. The thermal rregin/

low p'ressure trip set points shall be set according to l

the formula given on Figure 1-3. The variables in the formula are defined as:

B

' = High auctioneered thermal (AT) or nuclear power in % of rated power.

T IN = Core inlet temperature. 'F.

P Reactor pressure, psia.

Yy VAR= = Axial Shape Index, asiu Amendment No. 8,28,32, 1-8

1. 7.79,77,92

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1 l.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 containm_ent 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 21 kW/ft (deposited in the fuel) will exist as a consequence of axial power maldistributions.

Allowances have been made for instrumentation inaccuracies and uncertainties associated with the excore symmetric offset -

incore axial peaking relationship.

(9) Steam Generator Differential Pressure - The Asymetric Steam Generator Transient Protection Trip Function (ASGTPTF) utilizes a trip on steam generator differential pressure to ensure that neither a CNBR of less than 1.18 nor a peak linear heat rate of l more than 21 kW/ft occurs as a result of the loss of load to one steam generctor.

(10) Physics Testing at Low Power - During physics testing at power levels less than 10-8% of rated power, the tests may require that the reactor be critical. For these tests only the low reactor coolant flow gnd thermal margin / low pressure trips may be bypassed below 10-'% of rated power. Written test procedures

which are approved by the Plant Review Comittee will be in effget during these tests. At reactor power levels less than i 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.

c References

! . (1) USAR, Section 14.1 (2) USAR, Section 7.2.3.3 (3) USAR, Section 7.2.3.2 (4) USAR, Section 3.6.6 USAR, Section 14.6.2.2, 14.6.4 USAR, Section 14.7 USAR, Section 7.2.3.1 (8 USAR, Section 3.6 (9) USAR, Section 14.10 Amendment No. 7,M,79,77,92 1-9

TABLE 1-1 RPS LIMITING SAFETY SYSTEM SETTINGS ,

No. Reactor Trip Trip Setooints 1 High Power Level (A) 4-Pump Operation 3107.0% of Rated Power 2

Low Reactor Coolant Flow (B)(F) 4-Pump Operation >95% of 4 Pump Flow 3

Low Steam Generator Water Level 31.2% of Scale (Top of feedwater ring; 4'10" below normal water level) 4 Low Steam Generator Pressure (C) >500 psia 5 High Pressurizer Pressure <2400 psia 6

Thermal Margin / Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the re-actor coolant temper-ature as shown in Figure 1-3) 7 HighContainmentPressure(D) <5 psig 8

Axial Power Distribution (E) (Figure 1-2) 9 Steam Generator Differential Pressure <135 psid Amendment No. 7.32.47.77 92 1-10 1

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I 2.0 LIMITING CONDITIONS FOR OPERATION 1 1 2.1 Reactor Coolant System (Continued) I 2.1.1 Operable Components (Continued)  :

1 (a) A pressurizer steam space of 60% by volume or gNater l exists, or -

l (b) The steam generator secondary side temperature is less than 50*F above that of the reactor coolant system

, cold leg.

(12) Reactor Coolant System Pressure Isolation Valves i l (a) The integrity of all pressure isolation valves listed in Table 2-9 shall be demonstrated, except as specified

,~ in (b). Valve leakage shall not exceed the amounts

j indicated.

i (b) In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunc-i tional valve are in and remain in the mode corresponding to the isolated condition.*

(c) If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated and the reactor shall

be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I Basis i

i The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.18 during all l

normal operations and anticipated transients.

In the hot shutdown mode, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure i

considerations require that two loops be operable.

l In the cold shutdown mode, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be operable.

Thus, if the reactor coolant loops are not operable, this specification requires two shutdown cooling pumps to be operable.

The requirement that at least one shutdown cooling loop be in operation during

' refuelirg ensures that: (1) sufficient cooling capacity is available to .

remove decay heat and maintain the water in the reactor pressure vessel below 210*F as required during the refueling mode, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

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• Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplied deenergized.

Amendment No. 56, 9/ddf (/29/82, 79, 2-2b

///, 92

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 />.

_, i (2) Total Integrated Radial Peaking Factor ~

The calculated value of F T d (1+T)shallbelimitedt$<efinedbyFT,p 1.80. F i fromapowerdistributionmapwithn$pa$det!rmined 9

rt length I

CEAs inserted and with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, T , is the measured value of T at the time FRisdeEermined. q WithFf>1.80within6 hours: l l

i (a) Reduce power to bring power and Ff within the limits of Figure 2-9 withdraw the full i length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the PLCEAs, or (b) Be in at least hot standby.

(3) Total Planar Radial Peaking Factor

! The calculated value of F T defined as F T=F (1+T shallbelimitedtEY<1.85. F shlY1beH#-

' term 9n)edfromapowerdistributionm5%withnopart 1 I length CEAs inserted and with all full length CEAs at or above the Long Tenn Steady State Insertion

Lir,it for the existing Reactor Coolant Pump combi-nation. This determination shall be limited to core planes between 15% and 85% of full core height in-

. inclusive and shall exclude regions influenced by grid effects. The azimuthal tilt T , is the measured value of Tq at the time F,y isSetermined.

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 F T to with-inthelimitsofFigure2-9,withEfawthe full length CEAs to or beyond the Long Term -

4 Steady State Insertion Limits of Specifi-cation 2.10.2(7), and fully withdraw the PLCsAs, or (b) Be in at least hot standby.

i Amendment No. 32:43,47.79, 92 2-57a -

77,

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2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 _ Power Distribution Limits (Continued)

(4) Azimuthal Power Tilt (Tq)

=.

When operating above 70% of rated power.

(a) The azimuthal power tilt (Tq) shall not exceed 0.10 whenever Mini CECOR/BASSS is operable, the CEA's are at or above the Long Term Insertion Limit and Mini CECOR/BASSS is'being utilized to monitor F T and FT*

xy R (b) The azimuthal power tilt (Tq) shall not exceed 0.03 whenever the provisions of 2.10.4(4)(a) do NOT allow

' and M{ni F

t0. With JECOR/BASSS the indicated azimuthal to bepo'ver utilized tilt to monitor M*5et I

be >0.03 but <0.10, correct the power tilt within i

two hours or determine within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at

' least once per subsequent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, yhat the total integrated radial peaking factor, F is within the limit of Specification 2.10.4(2) an0,that the total planar radial peaking factor F T of 2.10.4(3), or reduce power t3Y less. is within than 70%the limit of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of confiming Tg >0.03.

(c) With the indicated power tilt detemined to be >.10, p wer operation may proceed up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided F and F T do not exceed the power limits of Figure i

2 9, or 55 in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

j Subsequent operation for the purpose of measurement to identify the cause of the tilt is allowable provided

! the power level is restricted to 20% of the maximum i allowable thermal power level for the existing reactor i coolant pump combination.

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Amendment No. 32,#3,(7,92 2-57b

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6 2.0 LIMITING CONDITION'S FOR OPERATION 2.10 ReactorCore(Continued) 2.10.4 Power 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 1 540*F* l (11) Pressurizer Pressure 1 2075 psia

• i *

(iii) Reactor Coolant Flow 1 197,000 gpm**

(iv) Axial Shape Index, Y g i Figure 2-7*** I i (b) With any of the above parameters exceeding the limit, restore j 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 />.

f Basis l Linear Heat Rate l

The ifmitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.

i Either of the two core power distribution monitoring systerrs, the Excore Detec-

! tor Monitoring System, or the Incore Detector Monitoring System, provide

. [ adequate monitoring of the core power distribution and are capable of verifying

' ( that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System perfoms this function by continuously monitoring the axial

shape index with the operable quadrant symetric excore neutron flux detectors and verifying that the axial shape index is maintained within the -11mble 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 F T of Figure 2-9.

3 In coajunction with the use of the excore monitoring system Ed in establishing  :

the axial shape index limits, the followin assumptions are made: (1) the CEA l insertion limits of Specification 2.10.2 and long term insertion limits of 1

! Specification 2.10.2.(7) are satisfied, the flux peaking augmentation i

r factors are as shown in Figure 2-8, and the total planar radial peaking I factor does not exceed the limits of Specification 2.10.4(3).

• Limit not applicable during either a themal pcwer ramp in excess of 5% of l rated themal power per minute or a thermal power step of greater than 10%

l of rated thermal power.

i i **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., 540*F, indicated, allows for an actual Te of542'F).

• "*The AXIAL SHAPE INDEX. Core power shall be maintained within the limits established by the Better Axial Shape Selection System (BASSS) for CEA insertions of the lead bank of < 65% when BASSS is operable, or,within the limits of Figure 2-7.

Amendment No. 32,#3,57,79,77, 92 2-57c y v- w, w ,,.--.......--v-,---,,-.,w.e,---.w.e---,3- , , ,, - ..- , .,--e.---&,--,-..-y._-.--,..,e-.,*,.- ,,... , -m,--.----.-,w-ge .-- m- a

3.0 SURVEILLANCE REQUIREMENTS 3.10 Reactor Core Parameters (Continued)

(6) Azimuthal Power Tilt (Tq) _

Whenever the core power is above 70% of rated power, the azimuthal power tilt shall be determined to be within its limits by calculating the tilt at least once every day using either:

a. The excore detectors with at least four safety channels operable, or I

b. The incore detectors with at least two strings of three rhodium detectors per full core height quadrant operable.

(7) DNB Parameters

a. The cold leg temperature, pressurizer pressure, and axial shape index shall be verified to be within the limits of Section 2.10.4(5) at least once per shift.

b. The reactor vessel coolant total flow rate shall be deter-mined to be within its limit by measurement at least once per month.

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Amendment No. E , 75, 92 3-63b

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70 80 90 100 110 120 CORE POWER (% OF RATED POWER) l l

ThermalMargin/LowPressureSafety OmahaPublicPowerDistrict figure limits 4PumpOperation FortCalhounStation-UnitNo.1 1-1 Amemsment no. 61, pp, thr,92

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B0 70 80 90 100 110 120 CORE POWER (% OF RATED POWER) l Pyg = 22 PF(B) Ai(Y)8+22.iTg -12674 PF (B) = 1.0 B2100% .

=

.008 8 + 1.8 50%<B<iOO%

= 1.4 BS50% .

Ai (Y) = .5Y1 + 1.125 , Y'y s .25

= .5Y3 + .875 Yy > .25 ThermalHargin/LowPressureLSSS 4PumpOperation DeahaPublicPowerDistrict figure FortCalhounStation-UnitNo.i 1-3 k ne ent W . 3, 3 , A7, 7Dr, W.92

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__._ ExcoreMonitoringofLHR OmahaPublicPowerDistrict figure FortCalhounStation-UnitNo.I 2-6 Aaendment No. 8. 29. 32, J3, p 4 ff, p y --~-

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