ML20086M999

Proposed Tech Specs 1.1,1.3,2.1.1,2.1.2,2.3,2.10.2 & Tables 1-1,2-2,3-1 & 3-5,eliminating Requirement to Inspect Fuel Elements Removed from Vessel Each Refueling

ML20086M999
Person / Time
Site:	Fort Calhoun
Issue date:	02/09/1984
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20086M946	List: ML20086M943 ML20086M973 ML20086M999
References
NUDOCS 8402170176
Download: ML20086M999 (140)

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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 to DNB. The minimum value of the DNBR during steady state oper-ation, normal operational transients, and anticipated tran-A DNBR of 1.22 corresponds to a l sients is limited to 1.22.

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 of AT in-struments), reactor coolant system pressure, and cold leg The area of safe oper- l temperature for which the DNBR is 1.22.

ation is below these lines.

The reactor core safety limits are based on radial 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 1.78.

The LSSS in Figure 1-3 is based on the assumption that the un-rodded integrated total radial peak (FRT) is 1.73. This peak- l ing factor is slightly higher (more conservative) than the maximum predicted unrodded total radial peak during core life, excluding measurement unce rtain ty.

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 tran-system is designed to prevent any anticipated combination of sient conditions for reactor coolant system temperature, pres-sure, and thermal that would result in a DNBR of less than 1.22.(3) power level References (1) USAR, Section 3.6.7 (2) USAR, Section 1.4.6 (3) USAR, Section 3.6.2 l

Amendment No. 8, $ 2, 4 7, 17, 70 1-2 ATTACHMENT A 8402170176 840214 p DR ADOCK 05000285 PDR

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1750 psia 530 70 80 90 100 110 120 CORE POWER (% OF RATED POWER) 1 ThermalMargin/LOWPressureSafety OmahaPublicPowerDistrict Figure Limits 4PumpOperation FortCalhounStation-UnitNo.i 1-1 Amendment No. 47, 70

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P = 13.5 PF (B) B + 18.6 T g -9574 PF (B) = 1.0 82100%

= .008 8 + 1.B 50%sBs100%

= 1.4 8550%

ThermalMargin/LowPressureLSSS OmahaPublicP0werDistrict Figure 4PumpOperation F0rtCalh0unStati0n-UnitNo.i 1 Amendment No. 8, 20, A7, 70 .

l.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limiting Safety System Settings, Reactor Protective System (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. 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 reactor trip is provided to protect the core against DNE should the coolant flow suddenly decrease significantly. Provisions are made in the reactor protective system to permit operation of uns reactor at reduced power if one or two coolant pumps are taken out of service. These low-flow and high-flux settings have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNB ratio above 1.22 under normal operation (4) and expected transients.(5) For reactor operation with one or two coolant pumps inoper-ative, the low-flow trip points, the overpower trip points, and the thermal margin / low pressure trip points and the axial power distribution trip points are simul-taneously changed when the pump condition selector switches (one per safety channel for a total of four switches) are set to the desired 2 or 3 pump posi-tion.(2)

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 shown in the follow-ing table:(6) 4 Pumps 100%

3 Pumps 75.3%

2 Pumps (each on a different steam generator) 49.6%

2 Pumps (both on same steam generator) 48.8%

During four-pump operation, the low flow trip setting of 95% insures that the reactor cannot operate when the flow rate is less than 93% of the nominal value con-sidering instrument errors. The high-power level trip, the thermal margin / low pressure trip, the low reactor coolant flow trip, and the axial power distribution l trip are reduced to compensate for the corresponding l core flow reduction experienced with fewer than-four L pumps in operation. The limits of trip points are shown in Table 1-1.(7) l Amendment No. 7, 32, 70 1-7 l

l __

I l

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

is consistent with the trip point assumed in the ac-cident ana' lysis.(1)

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

low pressure trip is provided to prevent operation when the DNBR is less than 1.22, including allowance for l measurement error. The thermal and hydraulic limits shown on Figure 1-3 define the limiting values of re-actor coolant pressure, reactor inlet temperature, and reactor power level which ensure that the thermal cri-teria(8) are not exceeded. The low set point of of 1750 a loss-psia trips the reactorThein the unlikely thermal event margin / low pressure of-coolant accident.

trip set points shall be set according to 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.

= Core inlet temperature, *F.

TIN PVAR = Reactor pressure, psia.

Amendment No. 8, 7d, 34, 47, 70 1-8

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 contain-ment high pressure is provided to assure that the re-actor 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 pres-sure signal which indicates safety injection system operation.

(8) Axial Power Distribution - The axial power trip is pro-vided 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 l

DNBR of less than 1.22 nor a maximum 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 inaccura-cies and uncertainties associated with the excore sym-metric offset - incore axial peaking relationship.

(9) Steam Generator Differential Pressure - The Asymmetric Steam Generator Transient Protection Trip Function (ASGTPTF) utilizes a trip on steam generator differ-ential pressure to ensure that neither a DNBR of less than 1.22 nor a peak linear heat rate of more than 21 kW/ft occur as a result of the loss of load to one steam generator.

I (10) Physics Testing at Low Power - During physics testing at power levels less than 10-l% of rated power, the tests may require that the reactor be critical. For these tests only the low reactor coolant flow and thermal margin / low pressure trips may be bypassed below 10-l% of rated power. Written test procedures which are approved by the Plant Review Committee will be in effect during these tests. At reactor power levels less than 10-l% 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 (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 (5) USAR, Section 14.6.2.2, 14.6.4 (6) USAR, Section 14.7 (7) USAR, Section 7.2.3.1 (8) USAR, Section 3.6 (9) USAR, Section 14.10 Amendment No. 7, 32, 70 1-9

TABLE l-1 )

RPS LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Trip Setpoints No.

1 High Power Level (A) 4-Pump Operation 1107.0% of Rated Power 3-Pump Operation 145% of Rated Power 2-Pump Operation 130% of Rated Power 2 Low Reactor Coolant Flow (B)(F) > 95% of 4 Pump Flow 4-Pump Operation > 71% of 4 Pump Flow 3-Pump Operation _

> 46% of 4 Pump Flow 2-Pump Operation _

3 Low Steam Generator Water Level 31.2% of Scale (Top of feedwater ring- 4'10" below normal water level)

Low Steam Generator Pressure (C) >500 psia 4 _

i 5 High Pressurizer Pressure i2400 psia Thermal Margin / Low Pressure (B)(F) 1750 psia to 2400 psia 6

(depending on the re-actor coolant temper-ature as shown in Figure 1-3) 7 High Containment Pressure (D) 15 psig Axial Power Distribution (E) (Figure 1-2) 8 9 Steam Generator Differential Pressure 1135 paid Amendment No. 7, $7, 47 1-10

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable components (Continued)

(a) A pressurizer steam space of 60% by volume or greater exists, or (b) The steam generator secondary side temper-4 ature is less than 50*F above that of the re-actor coolant system cold leg.

(12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure isolation valves listed in Table 2-9 shall be demon-strated, except as specified in (b). Valve leakage shall not exceed the amounts indi-cated.

(b) In the event that the integrity of any pres-sure 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 non-functional valve are in and remain in the mode corresponding to the isolated condi-tion.*

(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 />.

Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.22 during all normal operations and l anticipated transients.

In the hot shutdown mode, a single reactor coolant loop pro-vides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be operable.

In the cold shutdown mode, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal cap-ability for removing decay heat, but single failure consider-ations require that at least two loops be operable. Thus, if the reactor coolant loops are not operable, this specifi-cation requires two shutdown cooling pumps to be operable.

The requirement that at least one shutdown cooling loop be j in operation during refueling ensures that: (1) sufficient cooling capacity is available to remove decay heat and main-tain the water in the reactor pressure vessel below 210*F as required during the refueling mode, and (2) sufficient cool-ant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and pre-vent boron stratification.

• ttanual valves shall be locked in the closed position; motor operated valves -

shall be placed in the closed position and power supplies deenergized.

Amendment No. 56, 9/ddf/#didd/4/20/SJ, 70 2-2b

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1 l

2.0 LIMITING CONDITIONS FOR OPERATION

2.1 Reactor Coolant System (Continued) )

i i 2.1.2 Heatup and Cooldown Rate (Continued)

The curve in Figure 2-3 shall be used to pre-4 (a) dict the increase in transition temperature based on integrated fast neutron flux. If measurements on the irradiation specimens indi-Cdte a deviation from this curve, a new curve i

shall be constructed.

(b) The limit line on the figures shall be updated for a new integrated power period as follows:

the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure ( E> l MeV ) . For this 4

plant, based upon surveillance materials tests and the reduced vessel fluence rate provided 4 by core load designs beginning with fuel Cycle 8, the predicted surface fluence at the re-

- actor vessel belt-line weld material for 40 years at 1500 Mut and an 80% load factor is 3.6x1019 n/cm2 The predicted transition tem- l perature shift to the end of the new period

' shall then be obtained from Figure 2-3.

(c) The limit lines in Figures 2-1A and 2-1B shall

' be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the l

transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 82*F as it is set by the NDTT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall re-main at 162*F because components related to this temperature are also not subject to fast neutron flux.

(d) The Technical Specification 2.3(3) shall be re-l vised each time the curves of Figures 2-1A and i

2-1B are revised.

Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes.(1) These cyclic loads.are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown, the rates of temperature l

and pressure changes are limited. .The design number of cycles for heatup and cooldown is based upon a rate of 100*F in any one hour' period and for cyclic operation.

Amendment No. 22,47,6#,74 2-4 M [,_

2.0 LIMITING CONDITIONS FOR OPERATidN 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

The maximum allowable reactor coolant system pressure at any temperature is based upon the stress limitations for brittle These limitations are derived by fracture considerations.

using the rules contained in Section III(2) of the ASME Code including Appendix G, Protection Against Nonductile Failure, and the rules contained in 10 CFR 50, Appendix G, Fracture  :

l Toughness Requirements. This ASME Code assumes that a crack 10-11/16 inches long and 1-25/32 inches deep exists on the Furthermore, operating limits inner surface of the vessel.

on pressure and temperature assure that the crack does not grow during heatups and cooldowns.

The reactor Thevessel belt-line nilductility material transition consists (TNDT) temperature of six Of plates.

each plate was established by drop weight tests. Charpy tests were then performed to determine at what temperature the plates exhibited 50 ft-lbs. absorbed energy andNRC 35 mils lateral expansion for the longitudinal direction.

technical position MTEB 5-2 was used to establish a refer-ence temperature for transverse direction (RTNDT) of -12*F.

The mean RTNDT value for the Fort Calhoun submerged arc ves-sel weldments was determined to be -56*F with a standard de-viation of 17'F. In accordance with the methods identified in "NRC Staff Evaluation of Pressurized Thermal Shock", SECY 82-465, Appendix E, a weld material reference temperature (RTNDT) was established at -22*F based on a mean value plus twc standard deviations.

Similar testing was not performed on all remaining material in the reactor coolant system. However, sufficient impact testing was performed to meet appropriate design code re-quirements(3i and a conservative RTNDT of 50*F has been es-tablished.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT with operation.

The techniques used to predict the integrated fast 'autron

( E >l MeV)

Section fluxes 3.4.6 of USAR, of the the reactor vessel except thatare thedescribed integratedinfast neutron flux (E>l MeV) is 3.6x1019 n/cm2, including toler- l ance at the vessel inside surface, over the 40 year design life of the vessel.(5)

Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in The maximum exposure of the re-calculated flux magnitude.

actor vessel will be obtained from the measured sample ex-posure by application of the calibrated azimuthal neutron Amendment fio. 22, 47 H, 74 2-5 l

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) flux variation. The maximum integrated fast neutron (E>l -

MeV) exposure of the reactor vessel including tolerance is computed to be 3.6x1019 n/cm2 at the vessel inside surfaceThe l for 40 years operation at 1500 MWt and 80% load f actor.

exposure at the 1/4t depth from The the inner surface predicted TNDTis com-shift for puted to be 2.2x1019 n/cm2,tS) 19 an integrated fast neutron (E>l MeV) exposure of 2.2x10 n/cm 2 is 329'F, the value obtained from the curve shown in Figure 2-3. The actual shift in TNDT will be re-established periodically during the plant operation by testing of re-actor vessel material samples which are irradiated cumu-latively by securing them near the inside wall of the re-actor vessel as described in Section 4.5.3 and Pigure 4.5-1 of the USAR. To compensate for any increase in the TNDT caused by irradiation, limits on the pressure-temperature re-lationship are periodically changed to stayAnalysis within theof the stress limits during heatup and cooldown.

seccad removed irradiated reactor vessel surveillance speci- l men, combined with a new core loading design for Cycle 8, indicates that the fluence at the end of 8.5 Effective Full Power Years (EFPY) at 1500 MWt will be 1.2x1019 n/cm2 on 2 at the ir. side surface of the reactor vessel and 6.9x10 18 n/cm the 1/4t depth.(5) This results in a total shift of the RTNDT of 264'F for the area of greatest sensitivity (weld l metal) at the 1/4t location as determined from Figure 2-3.

Operation through fuel Cycle 9 will result in less than 8.5 EFPY.

The limit lines in Figures 2-18 and 2-1B are based on the following:

A. Heatup and Cooldown Curves - From Section III of the ASME Code, Appendix G-2215.

KIR = 2Kg+I KIT KIR = Allowance stress intensity factor at temper-(ASME III Figure atures related to RTNDT G-2110.1).

KIM = Stress intensity factor for membrane stress (pressure). The 2 repcesents a safety factor of 2 on pressure.

KIT = Stress intensity factor radial thermal gradient.

The above equation is applied to the reacto; vessel belt-line. For plant heatup the thermal stress is op-posite in sign from the pressure stress and consider-ation of a heatup rate would allow for a higher pres-sure. For heatup it is therefore conservative to con-sider an isothermal heatup or KIT = 0.

Amendment No. 22, #7, H , 74 2-6

RCSPRESS-TEMPLIMITSHEATUP 8.5EFPY REACTORNOTCRITICAL 1500NWt maImaEss esIA) 3200 t

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Amendment No. 74 ,

RCS PRESS-TEMP LIMITS C00LDOWN8.5 EFPY REACTORNOTCRITICAL 1500MWt mis e PSIA) 3200

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TECHNICALSPECIFICATIONS 2-1B "

Amendment No. 74 ,

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FORT CALHOUN FIGURE TECHNICAL 2-3 SPECIFICATIONS Amendment flo. 74

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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency core cooling System (Continued)

(3) Protection Against Low Temperature Overpressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPS pumps need not be required if the reactor ves-sel head, a pressurizer safety valve, or a PORV is re-moved.

Whenever the reactor coolant system cold leg temper-ature is below 330*F, at least one (1) HPSI pump shall l be disabled.

Whenever the reactor coolant system cold leg temper-ature is below 320*F, at least two (2) HPSI pumps shall l be disabled.

Whenever the reactor coolant system cold leg te aipe r-

_ature is below 282*F, all three (3) HPSI pumps shall be disabled.

In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection to the core.

Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor cool-ant. With this mode of start-up, the energy stored in the re-actor coolant during the approach to criticality is sub-stantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not re-quired.

The SIRW tank contains a minimum of 283,000 gallons of usable water containing at least 1700 ppm boron.(1) This is suffi-cient boron concentration to provide a shutdown margin of 5%,

including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 60*F.(2)

The limits for the safety injection tank pressure and volume assure the required amou.it of water injection during an ac-cident and are based on values used for the accident analyses.

The minimum 116.2 inch level corresponds to a volume of 825 ft3 and the maximum 128.1 inch level corresponds to a volume of 895.5 ft3, Amendment No. 17, 39, 43, S , 74 2-22

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)

Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct ,

alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be prcparly aligned prior to start-up of the reactor.

j The uperable status of the various systems and components is

< to be demonstrated by periodic tests. A large fraction of these tests will be performed while the reactor is operating in the power range.

If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. For a single com-ponent to be inoperable does not negate the ability of the system to perform' its function. If it develops that the in-operable component is not repaired within the specified allow-able time period, or a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of cooling requirements after a postulated loss-of-coolant ac-cident. This will also permit improved access for repairs in some cases. After a limited time in hot shutdown, if the mal-function (s) is not corrected, the reactor will be placed in the cold shutdown condition utilizing normal shutdown and cool-down procedures. In the cold shutdown condition, release of fission products or damage of the fuel elements is not con-sidered possible.

The plant operating procedures will require immediate action to effect repairs of an inoperable component and therefore in  !

most cases repairs will be completed in less than the speci-fied allowable repair times. The limiting times to repair are intended to assure that operability of the component will be restored promptly and yet allow sufficient time to effect re-pairs ucing safe and prcper procedures.

The requirement for core cooling in case of postulated loss-of-coolant accident while in the hot shutdown condition is significantly reduced below the requirements for a postulated loss-of-coolant accident during power operation. Putting the reactor in the hot shutdown condition reduces the consequences of a loss-of-coolant accident and also allows more free access to some of the engineered safeguards components in order to effect repairs.

Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indicative of a require-ment for major maintenance and, therefore, in such a case, the reactor is to be put into the cold shutdown condition.

With respect to the core cooling function, there is functional redundancy over most of the range of break sizes.(3)(4)

Amendment No. 12, 39, 47, 49, 74 2-23

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency core cooling system (continued)

The LOCA analysis confirms adequate core cooling for the break spectrum up to and including the 32 inch double-ended break as-suming the safety injection capability which most adversely af-fects accident consequences and are defined as follows. The entire contents of all four safety injection tanks are assumed to be available for emergency core cooling, but the contents of one of the tanks is assumed to be lost through the reactor coolant system. In addition, of the three high-pressure safe-ty injection pumps and the two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one low pressure operate while only one of each type is assumed to operate in the small break analysis (51; and also that 25% of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sizes considered are shown on USAR Figure 1-9 (Amendment No. 34).

Inadvertent actuation of three (3) HPSI pumps and three (3) charging pumps, coincident with the opening of one of the two PORV's, would result in a peak primary system pressure of 1190 psia. 1190 psia corresponds with a minimum permissible temper-ature of 330*F on Figure 2-1B. Thus, at least one HPSI pump is disabled at 330* F.

Inadvertent actuation of two (2) HPSI pumps and three (3) charging pumps, coincident with the opening of one of the two PORV's, would result in a peak primary system pressure of 1040 psia. 1040 psia corresponds with a minimum permissible temper-ature of 320*F on Figure 2-1B. Thus, at least two HPSI pumps will be disabled at 320*F.

Inadvertent actuation of one (1) HPSI and three (3) charging pumps, coincident with opening of one of the two PORV's, would j result in a peak primary system pressure of 685 psia. 685 psia corresponds with a minimum allowable temperature of 282*F l on Figure 2-1B. Thus, all three HPSI pumps will be disabled at 282*F. l Inadvertent actuation of three (3) charging pumps, coincident with the opening of one of the two PORV's, would result in a peak primary system pressure of 160 paia. 160 psia would cor-respond with a minimum allowable temperature that is less than the 82*F boltup temperature limit on Figure 2-18. Therefore,.

operation of the charging pumps need not be restricted.

l Removal of the reactor vessel head, one pressurizer safety valve, or one PORV provides sufficient expansion volume to limit any of the design basia pressure transients. Thus, no additional relief capacity is required.

Technical Specification 2.2(1) specifies that, when fuel is in the reactor, at least one flow path shall be provided for boric acid injection to the core. Should boric acid injection become necessary, and no charging pumps are operable, oper-ation of a single HPSI pump would provide the required flow path.

Amendment No. M , D , H , 74 2-23a

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.2 Reactivity Control Systems and Core Physics Parameters Limits Applicability Applies to operation of control element assemblies and moni-toring of selected core parameters whenever the reactor is in cold or hot shutdown, hot standby, or power operation conditions.

Objective To ensure (1) adequate shutdown margin following a reactor trip, (2) the MTC is within the limits of the safety analy-sis, and (3) control element assembly operation is within the limits of the setpoint and safety analysis.

Specification (1) Shutdown Margin With Tcold >210

• F Whenever the reactor is in hot shutdown, hot standby or power operation conditions, the shutdown margin shall be >4.0% Ak/k. With the shutdown margin <4. 0 %

ak/k, initiate and continue boration until the re-quired shutdown margin is achieved.

(2) Shutdown Margin With Tcold 2210*F Whenever the reactor is in cold shutdown conditions, the shutdown margin shall be >3.0% Ak/k. With the shutdown margin <3 . 0 % ak/k, initiate and continue boration until the required shutdown margin is achieved.

(3) Moderator Temperature Coefficient The moderator temperature coefficient (MTC) shall be:

a. Less positive than +0.2x10-4 ap/*F including uncertainties for power levels at or above 80%

of rated power.

b. Less positive than +0.5x10-4 ap/*F including uncertaintins for power levels below 80% of.

rated power.

c. More positive than -2.7x10-4 Ap/'F including l uncertainties at rated power.

With the moderator temperature coefficient confirmed outside any one of the above limits, change reacti-vity control parameters to bring the extrapolated MTC value within the above limits within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Amendment No. 8, 32, A3, A7, 62, 70 2-50

17 i i i i i i P

i 16 -

Y<

UNACCEPTABLE OPERATION i- .

I 15.22KW/FT E 15 -

E ACCEPTABLE OPERATION U ,

4 E +

"i g 14 -

E

-i 13 0 2000 4000 6000 8000 10000 12000 14000 CYCLE AVERAGE 31, NUP- (MWD /MTU)

AllowablePeakLinearHeatRate OmahaPublicPowerDistrict Figure VS.Burnup FortCalhounStation-UnitNo.1 2-5 Amendment No. 8,20,72,47

i10 i i i i i 100 -

g 90 -

(-0.04,86.0) (0.08, 86.0)

E g 80 -

(-0.E 76.0) ,(0.E 78.0)

W 70 - _

. S -

d 60 g 50 - -

5 40 - -

E 30 - -

to E

o 20 - -

10 - -

0 ' I I

-0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 AXIAL SHAPE INDEX Y1 i

1 LimitingCanditionforOperationfor OmahaPublicPowerDistrict Figure  !

ExcareManitoringofLHR FortCalhounStation-UnitN0.1 2  !

Amendment No. 8,20,22,43,A7,70

l l

110 i i i i i 100 - (-0.057, 100.5) (0.098, 100.5) -

90 - -

IE E 80 - -

E (-0.2, 75.0) (0.2,75.0)

.S 70 - -

tc a

60 - -

Es E 50 - -

x W 40 - -

2 g 30 - -

8 20 - -

10 - -

0 I ' I

-0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 AXIAL SHAPE INDEX Y y I.initingConditionforOperation OmahaPublicPowerDistrict Figure forDNBN0nitoring FortCalhounStation-UnitNo.i 2-7 Amendment No. 8, 20, 32, 47, 70

1 I

110 i i i i i i

100 -

FT LIMIT -

n 2

E l E 90 Q

y 8

l e  !

o 80 -

l -

0 FjLIMIT l 5 l 5

a. 70 - l N

a i o l t

60 -

lI -

i I I 0 I I i 1.70 1.75 1.80 1.85 1.90 1.95 2.00 FT AND FT n n i

i F j,F[and Core Power OmahaPublicPowerDistrict Figure Elmitations FortCalhounStation-UnitNo.1 2-9 Amendment No. D, 20, 32, 43, 47, 70

2.0 LIMITING CONDITIONS FOR OPERATION )

2.10 Reactor Core (Continued) l l

2.10.4 Power Distribution Limits Applicability Applies to power operation conditions.

Objective To ensure that peak linear heat rates, DNB margins, and radial peaking factors are maintained within acceptable limits during power operation.

Specification (1) Linear Heat Rate The linear heat rate shall not exceed the limits shown on Figure 2-5 when the following factors are appropriately included:

1. Flux peaking augmentation factors are shown in Figure 2-8,

2. A measurement-calculational uncertainty factor of 1.062,

3. An engineering uncertainty factor of 1.03,

4. A linear heat rate uncertainty factor of 1.002 due to axial fuel densification and thermal expansion, and

5. A power measurement uncertainty factor of 1.02.

(a) When the linear heat rate is continuously monitored b-1 the incore detectors, and the linear heat rate is exceeding its limits as indicated by four or more valid coincident incore detector alarms, either:

(i) Restore the linear heat rate to with-in its limits within one hour, or (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 />.

l l

j Amendment No. 5, 29, if, 3k, 4}, 47 l

2-56 l

l t

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

(ii)

Be in at least hot standby withir. 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 The calculated value of FRT defined by F RT = F R (1+T FR is determined l q ) shall be limited to <l.73.

from a power distribution map with no part length CEAs inserted and with all full length CEAs at or for above the Long Term Steady State Insertion Limit The the existing Reactor Coolant Pump combination. at azimuthal tilt, T g, is the measured value of Tg the time FR is determined.

l With F'T R >1.73 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 FR T within the limits of Figure 2-9, withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification or 2.10.2(7), and fully withdraw the PLCEAs, (b) Be in at least hot standby.

(3) Total Planar Radial Peaking Factor The calculated value of F T definedF as shall Fxy T=F xy l (1+Tq ) shall be' limited toxy<l.78. xy be de- I termined from a power distribution map with no part length CEAs inserted and with all full length CEAs at or above the Long Term Steady State Insertion Limit 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-clusive and shall exclude regions influenced by grid effects. The azimuthal tilt, To, is the measured value of T q at the time F xy is determined.

With Fxy >l.78 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 Fxy T to with-in the limits of Figure 2-9, withdraw the full length CEAs to or beyond the Long Term Steady State Insertion: Limits of Specifi-cation 2.10.2(7), and fully withdraw the PLCEAs, or (b) Be in at least hot standby.

Amendment No. 8A, 44, 47, 70 2-57a

~_ M~

.-- - ~- -- . .

r

! 2.0 LIMITING CONDITIONS FOR OPERATION ~ ,

2.10 Reactor Core (Continued) l 2.10.4 Power Distribution Limits (Continued)

DNBR Margin During Power Operation Above 15% of

~

l- (5)

' Rated Power i.

(a) The following DNB related parameters shall be 1

i maintained within the limits shown:

Cold Leg Temperature <545'F*~

! (i)

Pressurizer Pressure >2075 psia *

> (ii) >202,500 gpm*

• l (iii) Reactor Coolant ~ Flow l-

' (iv)~ Axial Shape Index, YI .jFigure 2-7 i

(b) With any of the above parameters exceeding the limit, restore the parameter .tcr 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

~I Linear Heat Rate

)

The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding j

will not exceed 2200*F.

Either of the two core power distribution monitoring 2

! systems, the Excore Detector Monitoring System, or.the Incore Detector Monitoring System, provide adequate moni-toring of the core power distribution and~are capable of t

verifying that the linear heat rate does not' exceed its.

limits. The Excore Detector Monitoring System performs j this function by continuously monitoring the axial shape

~

index.with the operable-quadrant symmetric excore neutron t-flux detectors and verifying.that the axial shape index is maintained within the allowable limits.of Figure 2-6 as ad-justed by Specification-2.10.4.(1).(c) for the allowed and-l' linear heat rate of Figure 2-5,'RC Pump' configuration, F xy T.of FigureL2-9. In conjunction with the.use of the.ex-i core monitoring system;and in establishing the axial shape index limits, the following: assumptions _are made: '(1) the

~

7 CEA insertion limits of' Specification 2.10.2.(6) and long-term insertion limits'of specification 2.10.2.(7.) are satis-

' fied, (2) the flux peaking augmentation. f actors: are 'as shown in Figure 2-8, (3) the' azimuthal _ power tilt re-strictions of Specification 2.10.4.(4)lare satisfied, and

~

(4) the total planar' radial peaking factor does'not exceed 4,

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 g

greater than-10% of rated thermal power. I

• • All values in this listingfare-indicated values _and include anfal-lowance for measurement 1 uncertainty (e.g., 545'F,: indicated,; allows i for-an actual T e of f 547' F) '..

> FORT CALHOUfL 2-57c Amendment No.~- 32,13,57,.70- ]

i

- . , . _ ~ - - . , , . . - - _ _ _._._._.,._h -,.,,_ .-, , . _ - . , , ,

TABLE 2-2 Instrument Operating Requiremente for Reactor Protective System Minimum Minimum Permissible Operable Degree of Bypass

-Channels Redundancy Condition N o ._ Functional Unit None Manual (Trip Buttons) 1 None 1

2(bc) 1(c) Thermal Power 2 High Power Level Input Bypass-ed Below 10-4% of Rated Power (a)(d) 2(b) 1 Below 10-4% of 3 Thermal Margin / Low Rated Power (a)(d)

Pressurizer Pressure 2(b) 1 None 4 High Pressurizer Pres,sure 2(b) 1 Below 10-4% of 5 Low R.C. Flow -

Rated Power (a)(d) 1/ Steam None 6 Low Steam Generator 2/ Steam Gen (b) Gen Water Level 1/ Steam Below 550 psia (a)(d'.

7 Lew Steam Generator 2/ Steam Gen (b) Gen Pressure 2(b) 1 During Leak Test 8 Containment High Pressure 2(bc) 1(c) During 15% of 9 Axial Power Distribution Rated Power (e) 2 1 Below 10-4% and 10 High Rate Trip-Wide Above 15% of Range Long Channels Rated Power (a)(e) 2(b) 1 Below 15% of 11 Loss of Load Rated Power (e) 2(b) 1 None 12 Steam Generator Differential Pressure a Bypass automatically removed. l b One of the inoperable channels must be in the tripped condition.

c If two channels are inoperable, load shull be reduced to 70% or less of rated power.

d For low power physics testing this trip may be bypassed up to 10-lg of rated power.

o For each channel, the same bistable automatically activates the Loss of Load and Axial Power Distribution (APD) trips and auto-Only matically bypasses the high rate trip at 15% of rated power.

the APD trip is a Limiting Safety System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

~ Amendment fio. 60 2-67

TABLE 3-1 (Continued)

MINIMUM FREOUENCIES FOR CHECKS, CALIBRATIONS AND TESTING OF REACTOR PROTECTIVE SYSTEM Check S a. Comparison of four level indications

6. Steam Generator Level a.

per generator.

b. Calibrate R b. Known differential pressure applied to sensors.

c. Test M(2) c. Distable trip tester.Il)

Check S a. Comparison of four pressure indi-

7. Steam Generator Pres- a.

cations per generator.

sure

b. Calibrate R b. Known pressure applied to sensors.

c. Test M(2) c. Bistable trip tester.(1)

a. Calibrate R a. Known pressure applied to sensors,

8. Containment Pressure w

b. Test M(2) b. Simulate pressure switch action.

di Test P a. Manually trip 2/4 turbine main steam

9. Loss of Load a.

stop valves.

Test F a. Manually test both circuits.

Manual Trips a.

10.

a. Check S a. Comparison of four differential pres-

11. Steam Generator sure indications between the two steam Differential Pressure generators.

R b. Known differential pressure applied to

b. Calibrate sensors,

c. Test M(2) c. Bistable trip test.(1)

M(2) a. Internal test circuits check logic Reactor Protection a. Test

12. networks and clutch power contacters.

System Logic Units

- ~

2 k

? TABLE 3-5 g (Continued)

• USAR Section Frequency Reference

• Test

,=__

4. Automatic and/or manual initia- At least once per plant operating gg0c. (Continued) cycl e .

• tion of the system stiall be de-toonstrated.

"as 1 year, 2 years, 5 years, and 9.10

11. Containment Cool- 1. Demonstrate damper action.

every 5 years thereaf ter ing and Iodine Removal Fuseable 2. Test a spare fuseable link. .

Linked Dampers During each refueling outage 8.4.3

12. Diesel Generator Calibrate.

Under-Vol tage

,w Relays Verify the contactor pickup value During each refueling outage

$13. Motor Operated Safety Injection at <8a% of 460 V.

Loop Valve Motor Starters (HCV-311, 314, 317,

-320, 327, 329, 331, 333, 312, 315,'318,321)

Verify control circuits operation During each refueling outage

.14. Pressurizer Heaters for post-accident heater use.

Test neutron poison samples for Intervals of 1, 2, 4,

> 15. Spent Fuel' Pool j Region 1 Racks dimensional change, hardness change, 7, 11, 15, 20, and 25 m and neutron attenuation change. years after installation.

E.

E!

w.

j JUST1FICATION, DISCUSSION, AND 'l SIGNIFICANT HAZARDS CONSIDERATIONS FOR CYCLE 9 RELOAD The Port Calhoun Technical Specifications are being amended toTable re-flect changes which are a result of the Cycle 9 core reload.

B-1 presents a summary of the Justification Technical Specification changes for the changes and the is con-explanation for the changes.

tained in the attached Fort Calhoun Cycle 9 Core Reload Evaluation. l' i-Significant Hazards Considerations:

It has been determined, based on the analytical information supplied in the Cycle 9 Core Reload Evaluation, that this amendment request This con-does not involve a significant hazards consideration.

clusion was derived by applying the Commission's guidance for imple-mentation of 10 CFR 50.92. The Commission provided this guidance con-corning the application of these standards through certain examples in the Federal Register, Volume 48, Number 87, Wednesday, April 6, 1983, Rules and Regulations. Example iii of actions involving no t significant hazards coasiderations, on page 14870 of the Federal Register, is quoted below:

"For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly dif-ferent from those found previously acceptable to the NRC This for a previous core at the facility in question are involved.

assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable."

As described in the Cycle 9 Core Reload Evaluation, no. fuel as-semblies to be loaded into the Cycle 9 core will be of new or dif-ferent design than those used previously and found to be acceptable to the NRC. No proposed changes to the Technical Specifications for Cycle 9 involve acceptance criteria which are significantly The minimumdifferent ac-from those previously found acceptable to the NRC.

ceptable DNBR limit has been increased to 1.22 from 1.19 (using the CE-1 correlation) to be consistent with the employment of the Sta-tistical Combination of Uncertainties methodology described in report 1983.

CEN-257(0)-P which was submitted to the Commission in November, The analytical methods used to demonstrate conformance with the Technical Specifications and regulations are. consistent with previou's OPPD-NA-NRC approvals (and documented in reports OPPD-NA-8301-P, 8302-P, and OPPD-NA-8303-P which were submitted to the Commission in September, 1983) and involve no significant changes.

include It is concluded that the proposed license amendment does not significant hazards considerations in that:

ATTACHMENT B

f

1. The probability or consequences All events of /accidents accidents previously evalu-not enveloped by l ated are not increased.

Cycle 8 parameters were evaluated and shown to have acceptable l

consequences, with violation of no safety limits.

2. The Cycle 9 core reload does not create the possibility of a new or different kind of accident from any previously evalu-ated. The core loading utilizes fuel management techniques which have previously been proven acceptable.

3. The Cycle ) core reload does not result in a reduction in the which margin of safety because the demonstrates Cycle 9 reloadthat evaluation, the margin of uses approved methodologies, safety is maintained in the revised Technical Specifications limits.

1 4

5 4

1 i

e 5

4

JUSTIFICATION, DISCUSSION, AND SIGNIFICANT HAZARDb CONSIDERATIONS FOR DELETION OF SPENT FUEL INSPECTION REOUIREMENTS The District has recently completed a comprehensive fuel assembly demonstration project in conjunction with the Department of Energy (DOE). As a part of this project, comprehensive fuel inspections were performed on a high burnup demonstration assembly which achieved an assembly average buraup of 52 GUD/MTU and fuel rods which had These undergone a simulated IN-OUT-IN fuel management duty cycle.

examinations included both poolside and hot cell inspections. i to de-The goals of the fuel assembly demonstration project were fuel for monstrate acceptable fuel performance for standard designTo ac-high burnup and IN-OUT-IN fuel management applications.

complish these goals, a fuel assembly was irradiated to a burnup for future reload cores and greater selected than the burnup anticipatedfuel reds were taken through a simulated IN-OUT-IN f management duty cycle. Comprehensive fuel inspections were conducted during the prbgram. The poolside inspections included visual examinations, assembly length measurements, shoulder gap measure-ments, fuel rod length measurements, fuel rod bow measurements, eddy current testing to determine fuel rod integrity, oxide film thickness measurements, and fuel rod profilometry. The hot cell examinations included fission gas release measurements, gamma scans, profilometry, burnup analysis, metalographic examinations, fuelThe pellet density results of measurements, and fuel rod cladding burst tests.

these examinations have been or will be published by DOS.

The District has concluded that the lead demonstration assembly suc-cessfully completed six cycles of irradiation having lead rod burnups of approximately 56 GWD/MTU with no rod perforations and no anomalies, which would indicate deterioration of performance cap-abilities. Excellent performance was also observed in the fuel rods which underwent a simulated IN-OUT-IN fuel management duty cycle.

The District has completed a program which demonstrated excellent of that currently planned for fuel performance for a burnup in excess Fort Calhoun reload fuel and for a fuel duty cycle more severe than that currently anticipated. Therefore, the District concludes that routine fuel element inspections are no longer required.

In accordance with the criteria contained in 10 CFR 50.92, this change involves no significant hazards considerations because:

(1) There is no significant increase in the probability or con- i sequence of an accident previously evaluated because no ac-cident evaluations are affected by this change in surveillance l requirements.

ATTACHMENT B

l (2) The change does not create the possibility of. a new or dif-ferent kind of accident from any accident previously evalu-ated because the deleted surveillance requirement was only applicable to fuel discharged from the core.

(3) No significant reduction in a margin of safety is involved because the completed inspection program demonstrates the adequacy of the current fuel assembly design.

i i

J 1

4 9

+% --

..t_ay ,c,. - . _ px,,m._ _ _ , , , , _ , , . _

l l

i TABLE B-1 .' l; Explanation for Cycle 9 Technical Specification Changes i

Tech. Spec. Number Changes Reasons Change i

! 1.1,1.3(2),1.3(4), Change minimum DNBR The minimum DNBR 1

1.3(8),2.1.1 value from 1.19 to has been raised Page 1-2, 1-7, 1-8, 1.22 to 1.22 to be 1-9, 2-2b consistent with statistical com-bination of un-certainties 1.1 Change total unrod- The unrodded 2

Page 1-2 ded planar radial planar radial peak from 1.67 to peaking factor

~

1.78 is being raised for Cycle 9 Change FR T from 1.62 The integrated 3 1.1 radial peaking Page 1-2 to 1.73 factor is being raised for Cycle 9 4 Figure 1-1 Replace Figure 1-1 The TM/LP safe-with enclosed Figure ty limits have 1 ~-1 been changed to reflect changes in peaking fac-tors and imple-mentation of statistical com-bination of un-certainties Figure 1-3 Replace Figure 1-3 The TM/LP trip 5

with enclosed Figure LSSS equation 1-3 has been adjust-ed to reflect the statistical combination of

' uncertalaties add a new section for New protective 6 1.3(9) iogic added to Page 1-9 'the asymmetric steam generator transient further reduce protection trip func- effect of a pos-tion sible asymmetric steam generator transient

~

l I

l TABLE B-1 Explanation for Cycle 9 Technical Specification Changes Change Tech. Spec. Number Changes Reasons 7 1.3(10) Change 9 to 10 for Maintain correct Page 1-9 physics tes m at numbering scheme low power Table 1.1 Add steam generc See Change 6 8

Page 1-10 differential pressur, setting 9 2.10.2(3) Change more positive Physics calcula-Page 2-50 than -2.5 x 10-4 tions predict a Ap/ F to more posi- more negative tive than -2.7 x 10-4 MTC at E0C Ap/ F 10 Figure 2-5 Replace Figure 2-5 Cycle 9 will be with enclosed Figure 18 month cycle 2-5 and the curve needed to be ex-tended to . allow for the higher burnups assoc 1-ated with long-er cycles 11 Figure 2-6 Replace Figure 2-6 The LHR excore with enclosed Figure LC0 has been 2-6 changed to re-flect higher radial peaking factors and the implementation of the statisti-cal combination of uncertainties program 12 Figure 2-7 Replace Figure 2-7 The DNB excore with enclosed Figure LC0 has been 2-7 changed to re-flect higher radial peaking factors and implementation of the statisti-cal combination of uncertainties program i

TABLE B-1 Exploriation for Cycle 9 Technical Soecification Changes Changes Reasons Change Tech. Spec. Number _

Replace Figure 2-9 The FxyT and 13 Figure 2-9 with enclosed Figure FR T limits have been changed to 2-9 reflect higher radial peaking factors in con-junction with the statistical combination of uncertainties program Change 1.07 to 1.062 Changed to reflect 14 2.10.4(1)(2) - CECOR accuracy Page 2-50 in measuring Fg (CENPD-153-P, ,

R ev . 1-P-A ,

INCA /CECOR Power Peaking ,

Uncertainty)

Change limited to The FRT changes 15 2.10.4(2) have been made Page 2-57a f 1.62 to limited to reflect pro-to < 1.73 and with FR T > 1.62 to with posed changes FRT > 1.73 in Tech. Spec.

1.1 Change limited to The FxyT changes -

16 2.10.4(3) have been made Page 2-57a f 1.67 to limited to reflect pro-to i 1.78 and with posed changes FR T > 1.67 to with in Tech. Spec.

FRT > 1.78 1.1 Change The RCS flow val-17 2.10.4(5)(a)(iii) >197,000 GPM** to ue was changed to Page 2-57c make the applica-

)

._202,500 GPM**

tion of uncer-tainties consis-tent with the other primary system parame-ters and to apply the un-certainty in a j manner consis- l tent with the statistical combination of uncertainties )

program l

! ~

i TABLE B-1 Explanation for Cycle 9 Technical Specification Changes Changes Reasons Change Tech. Spec. Number Change ** this num- See Change 17 18 2.10.4(5)(a)iii**

Page 2-57c ber is an actual limit (not including uncertainties) . All other values in this listing to all values in this listing Add steam generator See Change 6 19 Table 2-2 Page 2-67 differential pressure i

Add steam generator See Change 6 20 Table 3-1 Page 3-5 differential pressure as Item 11' Table 3-1 Change reactor pro- Maintain consis-21 tent numbering Page 3-5 tective system logic units from Item 11 scheme to Item 12 t

i l

~ ~

t JUSTIFICATION, DISCUSSION, AND f

SIGNIFICANT HAZARDS CONSIDERATIONS FOR-HEATUP AND COOLDOWN CURVES 4

' This amendment application is required to allow for safe operation of the Fort Calhoun reactor vessel and associated primary coolantThis system beyond 7.0 Equivalent Full Power Years (EFPY) of operation.

a application requests continued operation through 8.5 EFPY.

In determining the Limiting conditions for Operation at this in-creas(3 EFPY, the impact of the initial nil-ductility transition reference temperature (RTNDT) and the associated RTNDT shift must be 7

accounted for due to the-effect of Results neutron of'the fluence on welds of evaluation in the sur-

reactor vessel belt-line region.

veillance capsule W-265, which was removed after 5.92 EFPY (Cycle 7),

were extrapolated to predict an end-of-life fluence to the reactor i vessel of 4.8x1019 n/cm . However, beginning in Cycle 8, a new low 2

leakage core. loading pattern was implemented and is conservatively 4, estimated to reduce the rate of fluence to the reactor vessel by 30%.

Using this information in conjunction with the methodology of Regu-

! latory Guide 1.99, Revision 1, yields the predicted RTNDT shift're-flected in the proposed 8.5 ETPY heatup and,cooldown: limit curves.

These curves will ensure adequate fracture toughness is maintained through all conditions of normal-operation, including anticipated-l operational transients and system hydrostatic tests.

The proposed 8.5 EFPY heatup and cooldown curves are required for Cycle 9 operation. Commission approval of the' proposed Technical Specifications is, therefore, requested prior to Cycle 9'heatup to-ensure adequate curves are available to begin this cycle.

t Significant Hazards Considerations:

This Technical Specification amendment will not increase.the probabi-lity of occurrences or the consequences of an accident ,or malf unction 3

of equipment important to safety previously evaluated in the Safety.

Analysis Report'because the chatge only places conservative re-4 strictions on pressure-temperature limits for the reactor vessel.

The probability of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not c

be created because this application only revises'the'heatup and cool-

' down curves which.are bounded by the' existing Safety Analysis: Report.

The margin of ~ safe ty as : defined in .the , basis for the Technical Speci -

fications will not be reduced because'the methodology.of Regulatory-l'

~

. Guide. l.99,~ Revision 1, has again been used to determine the.value of

,, the RTNDT shift.

4 h

l

, . ,-v-, .

. .- - m --.- ~ - -- - m _ - .. . - - . - ~ , . - . . . . . . _ . -- ,

l l

l FORT CALHOUN UNIT 1 CYCLE 9 RELOAD EVALUATION

j  ;

Fort Calhoun l Cycle 9 License Application j

CONTENTS f

i;

1. INTRODUCTION AND

SUMMARY

2. OPERATING HISTORY OF THE REFERENCE CYCLE 1

3. GENT.RAL DESCRIPTION

4. FUEL SYSTEM DESIGN d

4 S. NUCLEAR DESIGN i

6. THERMAL-HYORAULIC DESIGN

7. TRANSIENT ANALYSIS

8. ECCS PERFORMANCE ANALYSIS

9. STARTUP TESTING

10. RPS ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION DESCRIPTION

11. REFERENCES 4

9 m

, . . - . . - . . , . , . , ,, --.e -, . . . - , . , -, .,, -

1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation of design and perfomance for the operation of Fort (.alhoun Station Unit 1 during its ninth fuel cycle at full rated power of 1500 MWt. All planned operating conditions remain the same as those for Cycle 8.

l The core will consist of 81 presently operating H, I and J assemblies, 8 fresh Batch J assemblies, 36 fresh Batch K assemblies and 8 G and I assemblies discharged from previous cycles.

The Cycle 9 analysis is based on a Cycle 8 termination point between 9,500 MWD /T and 10,000 MWD /T. In perfoming analyses of design basis events, detemining limiting safety settings and establishing limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 9 conditions would be enveloped, provided the Cycle 8 temination point falls within the above burnup range. The analysis presented herein will accommodate a Cycle 9 length of up to 13 500 3 MWD /T.

The eyaluation of the reload core characteristics have been conducted with respect to the Fort Calhoun Unit No.1 Cycle 8 safety analysis described in the 1983 update of the USAR, hereafter referred to as the

" reference cycle" in this report unless otherwise noted.

Specific core differences have been accounted for in the present anal-ysis. In all cases, it has been concluded that either the reference cycle analyses envelope the new conditions or the revised analyses pre-sented herein conticue to show acceptable results. Where dictated by variations from the previous cycle, proposed modifications to the plant Technical Specifications have been provided.

The Cycle 9 core has been designed to reduce fluence to critical reac-tor pressure vessel welds to minimize the RTNDT shift of these welds.

This will delay the time when the reactor vessel welds reach the Pres-surized Thermal Shock RTNDT screening criteria discussed in SECY 82-465. The use of a low radial leakage core design has resulted in in-creased radial pecking factors. The increased peakin; factors have been accommodated in the safety analysis through the use of a Statis-tical Combination of Uncertainties (SCU) Analysis. The SCU analysis was submitted in Reference 1.

The analysis presented in this report was perfomed utilizing the meth-odology documented in the District's reload analysis methodology re-ports (References 2, 3 and 4). These methodologies were previously transmitted in Reference 5.

The Cycle 9 analysis includes credit for an Asymmetric Steam Generator Transient Protection Trip Function. This RPS module is described in Section 10 and will be installed in the Fort Calhoun RPS during the up-coming outage.

2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE Fort Calhoun Station is presently operating in its eighth fuel cycle J utilizing Batch G, H, I and J fuel assemblies. Fort Calhoun Cycle 8 operation began on April 2,1983, and reached full power on May 7, 1983. The reactor has operated up to the present time with the core reactivity, power distributions and peaking factors having closely followed the calculated predictions.

It is estimated that Cycle 8 will be terminated on or about March 3, 1984. The Cycle 8 termination point can vary between 9,500 MWD /T and 10,000 MWD /T and still be within the assumptions of the Cycle 9 anal-yses. As of January 15, 1984, the Cycle 8 burnup had reached 8,461 MWD /T.

I l

e

3.0 GENERAL DESCRIPTION The Cycle 9 core will consist of the number and type of assemblies and fuel batches shown in Table 3-1. The primary change to the core in Cycle 9 is the removal of 29 Batch G assemblies and 19 Batch H assem-blies. These assemblies will be replaced by 20 fresh shimmed Batch K assemblies (3.50 w/o enrichnent),12 fresh unshimmed Batch K assem-blies (3.50 w/o enrichment), 8 fresh unshimmed Batch J assemblies (3.50 w/o enrichment), 4 Batch I assemblies (3.50 w/o initial enrich-ment) discharged from Cycle 7 and 4 Batch G assemblies (3.03 w/o ini-tial enrichment) discharged from Cycle 7.

Figure 3-1 shcws the fuel management pattern to be employed in Cycle

9. Figure 3-2 shows the locations of the poison pins within the lat-tice of fresh Batch K assemblies and the fuel rod locations in un-shimmed assemblies.

Figure 3-3 shows the beginning of Cycle 9 assembly burnup distribution for a Cycle 8 termination burnup of 10,000 MWD /T. The initial enrich-ment of the fuel assemblies is also shown in Figure 3-3. Figure 3-4 shows the end of Cycle 9 assembly burnup distribution. The end of Cycle 9 core average exposure is approximately 27,000 MWD /T and the average discharge exposure is approxinately 34,900 MWD /T.

i 1

l 1

==

w , - - . - - - -

Table 3-1 Fort Calhoun Cycle 9 Core Loading Initial BOC Batch EOC Batch Poison Poison Average Burnup (MWD /T) Average Burnup (MWD /T) Rods per Loading Assembly Number of

[E0C 8 = 10,000 MWD /T] [E0C 9 = 13,000 MWD /T] Assembly gm Bi o/ inch

-Designation Assemblies 27,751 39,527 0 0 GIII 4 31,691 38,715 0 0 H- -21 21,840 33,997 0 0 I 36 6,792 27,365 0 0 I(2) 4 9,351 24,491 0 0 J 28 0 14,407 0 0 J(3) -8 0 12,429 0 0 K -12 0 17,643 8 0.0238 K/ 20 l

TOTAL 133 (1) Assemblies Discharged From Cycle 7 i

(2) Assemblies Discharged Af ter One Cycle's Exposure From Cycle 7 (3) Assemblies Delivered in 1982 But Not Loaded Into Cycle 8

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - + - - - +

Fiqure 3-1 FORT CALHOUN CYCLE 9 CORE LOADING PATTERN 01 ,_ 02 03 04 05 OG 07

.< JM C' </

08 09 10 11 12 13

< J  :: </ -

14 15 16 17 18 19

< J  :: </  ::  ::*

20 21 22 23 24 25 26 _ _

J*  :: </ G*- J  ::

27 28 29 30 31 32 33 C' </  :: J  :: C' 34 35 36 37 38 39

</ -

M  :: J -

Figure 3-2 FORT CALHOUN CYCLE 9 ASSEf1BLY FUEL AND POISON UNSHIMMED ASSEMBLY R0D LOCATIONS I I I

l I

1 l

!1 l l l \

( '

l l

l' )

i i l

K-8 POISON ROD ASSEMBLY I I I l I l

N '

X j l \ \

l l

j l

i > X l l

\

l I X X l X X

__ FUEL ROD LOCATION E POISON ROD LOCATION

f Figure 3-3 FORT CALHOUN CYCLE 9 l B0C9 ASSEMBLY AVG. EXPOSURE AND INITIAL ENRICHMENT AA -QUARIG CORE LOCATION 01 02 j I H l B.88 -INITIAL ENRICHMENT,W/0 U-235 3.50 3.50 l CCCCC -ASSEMBLY AVG. EXPOSURE, NWD/T 22390 34836 l 03 04 05 06 07 H K JM J K/-

3.50 3.50 3.50 3.50 3.50 32638 0 0 6355 0 08 09 10 11 12 13 H K J I K/ H 3.50 3.50 3.50 3.50 3.50 3.50 32627 0 9576 20540 0 24893 14 15 16 17 18 19 K J I K/ I In 3.50 3.50 3.50 3.50 3.50 3.50 0 9556 21853 0 24175 7901 3 20 21 22 23 24 25 JM I K/ GN J I 26 3.50 3.50 3.50 3.50 3.50 3.50 I O 20505 0 27751 11417 16762 i 3.50 27 28 29 30 31 32 22297 J K/ I J I J 33 3.50 3.50 3.50 3.50 3.50 3.50 H 6263 0 24187 11350 23024 11890 3.50 34 35 36 37 38 39 I

34783 K/ H IN I J H 3.50 3.50 5.50 3.50 3.50 3.50 0 25438 5683 18414 9986 25317 NOTE: EOC 8 CORE AVERAGE BURNUP = 10000 MWD /T

Figure 3-4 FORT CALHOUN CYCLE 9 ASSEMBLY AVERAGE BURNUP AT E0C9 (MWD /T) 4 AA -GUARTER CORE LOCATION 01 02 BBBBB -ASSEMBLY AVG. EXPOSURE,NWD/T 27452 39873 03 04 05 06 07 37487 11742 14385 20953 15407 08 09 10 11 12 13 37481 13779 24833 34945 18412 38778 ,

i 14 15 16 17 18 19 11767 24854 35952 17913 38162 24679 20 21 22 23 24 25 14429 34967 17962 39527 26833 31694 26 27385 27 28 29 30 31 32 20927 18498 38268 26785 36434 27043 33 39842 34 35 36 37 38 39 15453 39317 23035 33128 25462 -38089 s . - - . .. , -- , ,- -,- ,.

i 4.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch K reload fuel is identical to that of the Batch J fuel described in the 1983 update of the USAR and the Cycle 8 reload submittal . The fuel system design and analysis for ENC fuel in the Fort Calhoun reactor is described in Reference 6.

J i

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f i

n

- ,, , r - - a ~- , - - , , - m

_. _ __ . -. _.- _ _ __ _ _ . _ ~ _ . _ _ _ -. _ __

f 5.0 NUCLEAR DESIGN 5.1 PHYSICAL CHARACTERISTICS l 5.1.1 Fuel Management l

The Cycle 9 fuel management uses a low-radial leakage design, with twice and thrice burned assemblies predom-

! inately loaded on the periphery of the core. Alow-radial leakage fuel pattern is utilized to minimize the flux to the pressure vessel welds. While this type of fuel management results in reduced pressure vessel flux over a standard out-in-in pattern, the radial peaking

, factors are increased.

I

! As described in Section 3.0, the Cycle 9 loading pat-

tern incorporates 32 fresh Batch K assemblies (20

shimmed, K/ and 12 unshimmed, K).with an enrichment of 3.50 w/o. In addition, 8 fresh Batch J assemblies (3.50 w/o U-235) which were not used in Cycle 8 are also inserted. Four once burned Batch I assemblies, I which were removed from the core at E0C7, and 4 thrice burned G assemblies,.which were removed at E0C7, are i

combined with 28 once burned Batch J assemblies, 36 .

twice burned Batch I assemblies, and 21 thrice burned Batch H assemblies to produce a Cycle 9 pattern with a cycle energy of 13,000 t 300 MWD /T. The Cycle 9 core ,

' characteristics have been examined for Cycle 8 tennina- '

tions between 9,500 MWD /T and 10,000 MWD /T and limiting i values established for the safety analysis. The load-ing pattern is valid for any Cycle 8 endpoint between these values.

i Physics characteristics including reactivity coeffi-4 cients for Cycle 9 are listed in Table 5-1 along with the corresponding values from the reference cycle. -

It should be noted that the values of parameters actu-ally employed in safety analyses are different from those displayed in Table 5-1 and are typical'iy chosen to conservatively bound predicted values with acconmo--

dation for appropriate uncertainties and allowances.

~

I

- Table 5-2 presents'a summary of CEA shutdown worths and i

reactivity allowances for the end of Cycle 9 Hot Zero Power Steam Line Break ' accident. The E0C HZP SLB is

, the most limiting accident of those used in the deter-mining of the~ required. shutdown margin. The Cycle 9 values calculated for mininum ' scram worth exceed those for Cycle 8 and thus, tensure an adequate margin to the Technical-Specification limit, on required shutdown mar-gi n.

.5.1.2 Power Distribution .

l. Figures 5-1 through 5-3. illustrate the all rods out

.(ARO) planar radial 'powe'r distributions at B0C9, MOC9 l

1 >

x - 'i ,,,,%. , e, . -u ., ,mm.m, m . . , o

5.0 NUCLEAR DESIGN (Continu:d) 5.1 PHYSICAL CHARACTERISTICS (Continued)

S.1.2 Power Distribution (Continued) and E0C9, respectively, that are characteristic of the high burnup end of the Cycle 8 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel height. The high burnup end of the Cycle 8 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 9. The planar radial power distribu-tions for the above region with Bank 4 fully inserted at beginning and end of Cycle 9 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 8 shutdown window.

The radial power distributions described in this sec-tion are calculated data without uncertainties or f her allowances. However, the single rod power peaking val-ues do include the increased peaking that is character-istic of fuel rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded con-figurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 9. These conservative values, which are used in Section 7 of this document, establish the allowable lim-its for power peaking to be observed during operation.

Figures 3-3 and 3-4 show the integrated assembly bcrnup values at 0 and 13,000 MWD /T, respectively, based on an E0C8 burnup of 10,000 MWD /T.

The range of allowable axial peaking is defined by the limiting conditions for operation covering the axial shape index (ASI). Within these ASI limits, the neces-sary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-di-mensional or total peaking factor anticipated in Cycle 9 during nonnal base load, all rods out operation at full power is 1.84, not including uncertainty allow-ances.

5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data The maximum reactivity worth and planar pow-er peaking factors associated with an eject-ed CEA event are shown in Table 5-3 for both beginning and end of Cycle 9. These values encompass the worst conditions anti-cipated during Cycle 9 for any expected Cycle 8 tennination point. The values shown for Cycle 9 are calculated in accord-ance with Reference 4. In addition, Table

l 5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICALCHARACTERISTICS(Continued) 5.1.3 SafetyRelatedData(Continued) 5.1.3.1 Ejected CEA Data (Continued) 5-4 lists those values used in the Reference Analysis (Cycle 6) for compa rison, 5.1.3.2 Dropped CEA Data i The Cycle 9 safety related data for the dropped CEA analysis were calculated iden-tically to that used in Cycle 8. The data is reported in the dropped CEA analysis.

5.2 ANALYTICAL INPUT TO IN-CORE MEASUREMENTS In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the same manner as for Cycle 8.

5.3 NUCLEAR DESIGN METHODOLOGY Analyses have been performed in the manner and with the method-ologies documented in References 2 and 3.

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties which are applied to Cycle 9 are the same as those presented in Reference 2.

~ . -.

TABLE 5-1 FORT CAI HOUN CYCLE 9 NOMINAL PHYSICS CHARACTERISTICS Reference Units Cycle

• Cycle 9 Critical Baron Concentration Hot Full Power, AR0, 1108 Equilibrium Xenon, B0C PPM 840 Inverse Boron Worth 100 110 Hot Full Power, B0C PPM /%Ap 88 Hot Full Power, E0C PPM /%Ap 85 Reactivity Coefficients (CEAs Withdrawn)

Moderator Temperature Coefficients i

10-4Ap/*F +0.16 40.86 Beginning of Cycle, HZP End of Cycle, HFP 10-4ap/* F -2.4 -2.5 Doppler Coefficient Hot Zero Power, BOC 10-5ap/ F -1.93 -1.78 10-5ap/*F -1.41 -1.40 Hot Full Power, B0C Hot Full Power, E0C 10-53pfer _1,46 -1.58 Total Delayed fleutron Fraction, s eff 0.00592 0.00640 BOC 0.00528 0.00545 E0C Neutron Generation Time, t*

10-6 sec .23.7 23.1 ,

B0C i

10-6 sec 27.9 30.7 l E0C -

l

• Cycle 8

i TABLE 5-2 l

FORT CALHOUN UNIT 1 CYCLE 9 LIMITING VALUES OF l REACTIVITY WORTHS AND ALLOWANCES FOR HOT ZER0 POWER STEAM LINE BREAK, %ap END-0F-CYCLE (E0C)

Reference Cycle (Cycle 8) Cycle 9 9.12 9.95

1. Worth of all CEA's Inserted 2.06 1.88

2. Stuck CEA Allowance

3. Worth of all CEA's Less Worth 7.06 8.07 of Most Reactive CEA Stuck Out

4. Power Dependent Insertion 1.11 1.21 Limit CEA Worth 5.85 6.96

5. Calculated Scram Worth 0.59 0.69

6. Physics Uncertainty plus Bias 5.26 6.27

7. Het Available Scram Worth

8. Technical Specification 4.00 Shutdown Margin 4.00

9. Margin in Excess of Technical 2.27 Specification Shutdown Margin 1.26 4

4 E

5

N TABLE 5-3 FORT CALHOUN UNIT 1 CYCLE 9 CEA EJECTION DATA Cycle 6 Value 80C9 Value E0C9 Value I

Maximum Radial Power Peaking Factor Full Power with Bank 4 inserted; worst CEA 6.00 3.24 3.19 i ejected Zero power with Banks 4+3 inserted; worst CEA ejected - 13.00* 4.04 3.72 Maximum Ejected CEA Worth (%Ap)

Full power with Bank 4 inserted; worst CEA ejected 0.30 0.23 0.18.

Zero Power with Banks 4+3 inserted; worst CEA ejected 0.90* 0.30 0.22

• Banks 4+3+2 inserted l

-. - , , , , , , - , r

Figure 5-1 FORT CALHOUN CYCLE 9

! ASSEME'.Y

. RELATIVE POWER DENSITY 0 MWD /T, HFP, EQ. XENON l

AA -GUARTER CORE LOCATION 01 02 8.8888 -RPO 0.3393 0.3107 03 04 05 06 07 0.3339 0.8896 1.1015 1.0957 1.0823 08 09 10 11 12 13

~

0.3335 1.0950 1.2066 1.1045 1.3649 1.0388 14 15 16 17 18 19 0.8971 1.2158 1.0748 1.3267 1.0697 1.3541 20 21 22 23 24 25 1.1149 1.1179 1.3365 0.8990 1.2388 1.2276 26 0.3444 27 28 29 30 31 32 1.1128 1.3858 1.0867 1.2434 1.0849 1.2549 33 0.3159 34 35 36 37 38 39 1.0981 1.0488 1.4229 1.2118 1.2885 1.0493 X

X = HAXINUM i-PIN PFAK = 1.6192 m m

Figure 5-2 FORT CALHOUN CYCLE 9 MSSEMBLY RELATIVE POWER DENSITY 6000 MWD /T, llFP, EQ. XENON AA -00ARTER CORE LOCATION 01 02 B.BBBB -RPO 0.3945 0.3762 03 04 05 06 07 0.3571 0.9042 1.1226 1.1624 1.2247 08 09 10 11 12 13 0.3578 1.0793 1.1756 1.1066 1.4519 1.0690 14 15 16 17 18 19 0.9059 1.1792 1.0695 1.3879 1.0487 1.2827 20 21 22 23 24 25 1.1253 1.1108 1.3915 0.8916 1.1431 1.1019 26 0.3961 27 28 29 30 31 32

. 1.1670 1.4579 1.0541 1.1435 0.9666 1.0969 33 X

0.3781 34 35 36 37 38 39 1.2281 J.0670 1.3299 1.0821 1.1243 0.9153 V = HAXIMUM i-PIN PEAK = 1.6190 l

I l

Figure 5-3 FORT CALHOUN CYCLE 9 ASSEMBLY RELATIVE POWER DENSITY 13000 MWD /T, HFP, EQ. XENON I

AA -GUARTER CORE LOCATION 01 02 B.8888 -RPD 0.4530 0.4445 03 04 05 06 07 0.3918 0.9267 1.1237 1.1870 1.3266 08 09 10 11 12 13 0.3920 1.0711 1.1360 1.0887 1.4640 1.0739 X

14 15 16 17 18 19 0.9260 1.1369 1.0542 1.4078 1.0293 1.2137 20 21 22 23 24 25 i.1221 1.0885 1.4076 0.9017 1.0892 1.035f:

28 0.4529 27 28 29 30 31 32 1.1860 1.4618 1.0296 1.0884 0.9221 1.0322 33 0.4445 34 35 36 37 38 39 1.3235 1.0675 1.2493 1.0169 1.0565 0.8760 X = HAXIMUM i-PIN PEAK = 1.6401

Figure 5-4 FORT CALHOUN CYCLE 9 RPD WITH BANT 4 INSERTED 0 MWD /T, HFP, EQ. XENON AA -GUARTER CORE LOCATION 01 02 8.8888 -RPD 0.3674 0.3481 03 04 05 06 07 0.2180 0.7759 1.1279 1.2051 1.2149 08 0 10 11 12 13 0.2198 1.0113 1.1236 1.4948 1.1603 14 15 16 17 18 19 0.7849 1.0215 1.0060 1.3774 1.1648 i.4933 20 21 22 23 24 25 1.1446 1.1397 1.3890 0.9428 1.3158 1.2981 26 0.3739 27 28 29 30 31 32 1.2267 1.5209 1.1855 1.3220 1.0766 1.1574 33 0.3546 34 35 36 37 38 ,

1.2353 1.1739 1.5724 1.2831 1.1883 6592 X

X = MAXIMUM i-PIN PEAK = 1.8021

[ CEA BANK 4 LOCATION l

l i

, , - - - - -w - - , , .

Figure 5-5 FORT CALHOUN CYCLE 9 RPD WITH BANK 4 INSERTED 13000 MWD /T, HFP, EQ. XENON AA -00ARTER CORE LOCATION 01 02 B.8888 -RPD 0.4954 0.5027 ,

03 04 05 06 07 0.2495 0.8027 1.1587 1.3166 1.5005 08 .

10 11 12 13 0.2496 .5018 0.9460 1.1124 1.6096 1.?034 X

14 15 16 17 18 19 0.8016 0.9466 0.9881 1.4611 1.1180 1.3357 20 21 22 23 24 25 1.1562 1.1116 1.4604 0.9512 1.1488 1.0865 26 0.4949 27 28 29 30 31 32 1.3146 1.6062 1.1182 1.1477 0.9018 0.9293 33 0.5024 34 35 36 37 38 39 f.4959 1.1954 1.3745- 1.0662 0.9505 0.5062 X = HAXIMUM i-PIN PEAK = 1.8337 CEA BANK'4 LOCATION i

. , .y

6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 9 at the rated power of 1500 MWt have been performed using the TORC computer code described in Reference 1; the CE-1 critical heat flux correlation des-cribed in Reference 2, and the CETOP-D conputer code described in Reference 3. This combination was used in the reference cycle (Cycle 8) Fort Calhoun reload analysis (Reference 4) and the reload methodology can be found in Reference 5.

Table 6-1 contains a list of pertinent thermal-hydraulic para-meters used in both safety analyses and for generating reactor protective system setpoint information. Also note that the cal-culational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed on Table 6-1 have been conbined statistically with other uncertainty factors at the 95/95 confidence /probabil-ity level (Reference 6) to define a new design limit on CE-1 minimum DNBR.

6.2 FUEL R00 BOWING The fuel rod bow penalty accounts for the adverse impact on MDNBR of random variations in spacing between fuel rods. The penalty at 40,000 MWD /MTU burnup is 0.5% in MDNBR. This penalty was applied to the new design limit in the stdtistical combina-tion of uncertainties (Reference 6).

l l

l l

i l .

TABLE 6-1 Fort Calhoun Unit 1 Thennal-Hydraulic Parameters at Full Power Unit Cycle 9*

Total Heat Output (Core Only) MWt 1500 106 BTV/hr 5119 Fraction of Heat Generated in Fuel Rod .975 Primary System Pressure Nominal psia 2100 Minimum In Steady State psia 2075 Maximum In Steady State psia 2150 Inlet Temperature . *F 545 Total Reactor Coolant Flow gpm 202,500 (Steady State) 106 lbm/hr 76.45*

(Through the Core) 106 lbm/hr 73.04*

Hydraulic Diameter ft .044 (Nominal Channel)

Average Mass Velocity lbm/hr-ft2 2.24 Core Average Heat Flux BTU /hr-ft2 182,180**

(Accounts for Heat Generated in the Fuel Rod)

Total Heat Transfer Surface Area ft2 28,101**

Average Core Enthalpy Rise BTU /lbm 70.1 Average Linear Heat Rate kw/ft 6.04**

Engineering Heat Flux Factor 1.03***

Engineering Factor on Hot Channel Heat Input 1.03***

Rod Pitch and Bow 1.065***

Fuel Densification Factor (Axial) 1.01***

• Design inlet temperature and nominal primary system pressure were used to calculate these paracters.

• • Based on a Cycle 9 specific value of 160 shims.

• • • These factors were combined statistically (Reference 6) with other uncertainty factors at 95/95 confidence / probability level to define a design limit on CE-1 minimum DNBR.

7.0 TRANSIENT ANALYSIS This section presents the results of the Omaha Public Power District Fort Calhoun Station Unit 1, Cycle 9 Non-LOCA safety analysis at 1500 MWt.

The Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7-1. These events were categorized in the following aroups:

1. Anticipated Operational Occurrences (A00s) for which the inter-vention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable limits.

2. A00s for which the intervention of the RPS trips and/or initial steady state thennal margin, maintained by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding accept-able limits.

3. Postulated Accidents The Design Basis Events (DBEs) considered in the Cycle 9 safety anal-yses are listed in Table 7-1. Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design limits are presented in Table 7-2.

As indicated in Table 7-1, no reanalysis was perfonned for the DBEs for which key transient input parameters are within the bounds (con-servative with respect to) of the reference cycle values (Fort Calhoun Updated Safety Analysis Report including Cycle 8, Reference 1). For these DBEs the results and conclusions quoted in the reference cycle analysis are valid for Cycle 9.

For the events reanalyzed, Table 7-3 shows the reason for the reanaly-sis, the acceptance criterion to be used in judging the results and a sunnary of the results obtained. Detailed presentations of the re-sults of the reanalyses are provided in Sections 7.1 through 7.3.

TABLE 7-1 FORT CALHOUN UNIT 1, CYCLE 9 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS Analysis Status 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits:

7.1.1 Boron Dilution Reanalyzed 7.1.2 Startup of an Inactive Reactor Coolant Pumpl Not Reanalyzed 7.1.3 Loss of Load Not Reanalyzed 7.1.4 Excess Load Reanalyzed 7.1.5 Loss of Feedwater Flow Not Reanalyzed 7.1.6 Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed 7.1.7 Reactor Coolant System Depressurization Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits:

7.2.1 Sequential CEA Group Withdrawal 2 Reanalyzed 7.2.2 Loss of Coolant Flow 3 Reanalyzed 7.2.3 Full Length CEA Drop Reanalyzed 7.2.4 Part length CEA Dropl Not Reanalyzed 7.2.5 Transients Resulting from the Malfunction of One Steam Generator 4 Reanalyzed 7.3 Postulated Accidents 7.3.1 CEA Ejection Not Reanalyzed 7.3.2 Steam Line Break Not Reanalyzed 7.3.3 Steam Generator Tube Rupture Not Reanalyzed 7.3.4 Seized Rotor 3 Reanalyzed 1 Technical Specifications preclude this event during operation.

2 Requires High Power and Variable High Power Trip 3 Requires Low Flow Trip 4Requires trip on high differe7ttial steam generator pressure

TABLE 7-2 FORT CALHOUN UNIT 1, CYCLE 9 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle (Cycle 8)

Physics Parameters Units Values Cycle 9 Values Radial Peaking Factors For DNB Margin Analyses (FRT)

Unrodded Region 1.65 1.75+>*

Bank 4 Inserted 1.69 1.79+,*

For Planar Radial Component

! (F T) of 3-D Peak (C Limit Analyses) ,

Unrodded Region 1.72 1.78+,*

Bank 4 Inserted 1.81 1.93+>*

Maximum Augmentation Factor- 1.057 1.057 Moderator Temperature Coefficient 10-4ap/*F -2.5 to +0.5 -2.7 to +0.5 Shutdown Margin (Value Assumed in Limiting E0C Zero Power SLB)- %ap -4 . 0 -4.0 Tilt Allowance  % 3.0 3.0

• For DNBR and CTM calculations, effects of. uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to.DNB and CTM limits are detailed in References 2a, 2b, 2c.

+The values. assumed are conservative with respect to the Technical Specifica-tion limits.

TABLE 7-2 (Continued)

Units Cycle 8 Values Cycle 9 Values Safety Parameters MWt 1530 1500*

Power Level Maximum Steady State 545*

• F 547 Temperature Minimum Steady State 2075*

psia 2053 Pressurizer Pressure gpm 197,000 208,280*

Reactor Coolant Flow Negative Axial Shape LC0 Extreme Assumed -0.18 at Full Power (Ex-Cores) Ip -0.20 Maxinum CEA Insertion  % Insertion 25 of Bank 4 25 at Full Power Maximum Initial Linear Heat Rate for Transient 15.22 Other than LOCA KW/ft 15.22 Steady State Linear Heat Rate for Fuel CTM Assumed in the Safety KW/ft 21.0 21.0 Analysis CEA Drop Time to 100%

Including Holding Coil 3.1 Delay sec 3.1 1.19 1.22*

Minimum DNBR (CE-1)

• For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. The procedures ~used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c.

TABLE 7-3 DESIGN BASIS EVENT REANALYZED FOR FORT CALHOUN CYCLE 9 Acceptance Summa ry Reason for Reanalysis Criterion of Results Event .

(changes relative '

to reference cycle) l Dilution to critical Acceptance criteria Boron Dilution Increased critical boron met. See Section concentrations and shut- time limits of 30 minutes down margin changes fran for refueling and 15 7.1.1 for details.

Cycle 8 incorparated. minutes for all other subcritical modes must be met.

4 Change in TM/LP (Pvar) - Pbias = 64.9 psia Excess Load which is more limit-trip equation and MTC ing (as in Cycle 8)

Tech. Spec. limit. than the RCS Depres-Reevaluate Pbias term, surization.

Change in Tech. Spec. - Ph ias = 27.5 psia RCS Depressurization wh'ch is less limiting negative limit on MTC.

than that of Excess Load event.

Change in classification  !!inimum DNBR greater MDNBR = 1.27 Sequential CEA Group Withdrawal than 1.22 using CE-1 PLHGR < 21 kw/ft.

of event from "A00 for which intervention of correlation. Transient RPS is necessary to pre- PLHGR < 21 kw/ft.

vent exceeding acceptable limits" to "A00 for which -

sufficient steady state thermal margin must be maintained by LCOs", and changes in Tech. Spec.

limits on radial peaking

• factors.

TABLE 7-3 (Continued)

DESIGN BASIS EVENT REANALYZED FOR FORT CALHOUN CYCLE 9 Reason for Acceptance Summa ry Evert Reanalysis Criterion of Results (changes relative to reference cycle)

Loss of Coolant Flow Increased Tech. Spec. Minimum DNBR greater Minimum DNBR = 1.48 limits on radial peak- than 1.22 using CE-1 ing factors. correlation.

Full Length CEA Drop Increased Tech. Spec. Minimum DNBR' greater Minimum DNBR = 1.44 limits on radial peaking than 1.22 using CE-1 factors and more negative correlation.

MTC allowed by Tech. Spec.

Asymmetric Steam Generator Event Increased Tech. Spec. limits Minimum DNBR greater Minimum DNBR = 1.53 (Loss of Load to One Steam on radial peaking factors, than 1.22 using CE-1 Generator) more negative MTC Tech, correlation.

Spec. limit, and incorpora-tion of steam generator differential pressure trip into RPS.

Seized Rotor Increased Tech. Spec. limits Site boundary dose within Site boundary dose on radial peaking factors. 10CFR100 limits, specific- acceptable. Less ally less than 1% failed than 1% failed fuel.

fuel.

7.0 TRANSIENT ANALYSIS 7.1 (Continued) 7.1.1 Boron Dilution Event The Boron Dilution event was reanalyzed for Cycle 9 to determine if sufficient time is available for an opera-tor to identify the cause and to terminate an approach to criticality for all subcritical modes of operation.

It was also analyzed to verify corresponding shutdown margin requirements for modes 2 through 5 as they are defined by the Technical Specifications. The event was analyzed using the methods of Reference 3.

Table 7.1.1-1 compares the values of the key transient parameters assumed in each mode of operation for Cycle 9 and the reference cycle (Cycle 8).

Table 7.1.1-2 compares the results of the analysis for Cycle 9 with those for Cycle 8. The key results are the minimum times required to lose prescribed negative reactivity in each operational mode. As seen from Table 7.1.1-2, sufficient time exists for the operator to initiate appropriate action to mitigate the conse-quences of this event.

TABLE 7.1.1-1 FORT CALHOUN CYCLE 9 KEY PAR #tETERS ASSUMED Ill THE B0RON DILUTI0tl ANALYSIS Cycle 8 Cycic 9 Parameter Critical Boron Concentration, PPM ( All Rods Out, Zero Xenon)_

Mode 1330 1560 Hot Standby 1330 1560 Hot Shutdown 1340 1360 Cold Shutdown - Normal RCS Volume 1145 1190 Cold Shutdown - Minimum RCS Volume

• 1260 1290 Refueling a

Inverse Boron Worth, PPM /%Ao Mode 90 90 Hot Standby 55 55 Hot Shutdown 55 55 Cold Shutdown - Normal RCS Volume 55 55 Cold Shutdown - Minimum RCS Volume 55 55 Refueling Minimum Shutdown Margin Assumed, %Ap

-Mode .

-3.0 -4.0 Hot Standby

-3.0 -4.0 Hot Shutdown

-3.0 -3.0 Cold Shutdowa - Normal RCS Volume

.-3.0 -3.0 Cold Shutdown - Minimum RCS Volume Refueling

• Shutdown Groups A and B out, all Regulating Groups inserted except most reactive rod stuck out.

• • 1700 ppm initially

_ _ ~

TABLE 7.1.1-2 FORT CALHOUN CYCLE 9 RESULTS OF THE BORON DILUTION EVENT l

Criterion For Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown Margin (Min) Margin (Min)

Mode Cycle 8 Cycle 9 82.5 92.7 15 Hot Standby 40.1 45.2 15 Hot Shutdown Cold Shutdown - Normal 15 RCS Volume 39.8 39 .3 Cold Shutdcwn - Minimum 15 RCS Voluce 17.0 16.4 38.0 35.0 30 Refueli n9

7.0 TRANSIENT ANALYSIS (Continutd) 7.1 (Continued) 7.1.4 Excess Load Event

The Excess Load event was reanalyzed for Cycle 9 to l detennine the pressure bias term for the TM/LP trip setpoint.

l The Excess Load event is one of the DBEs analyzed to determine the maximum pressure bias term input to the TM/LP trip. The methodology used for Cycle 9 is des-cribed in References 3 and 4. The pressure bias term accounts for margin degradation attributable to mea-surement and trip system processing delay times.

Changes in core power, inlet temperature and RCS pres-sure during the transient are monitored by the TM/LP trip directly. Consequently, with TM/LP trip setpoints and the bias term determined in this analysis, adequate protection will be provided for the Excess Load event to prevent the acceptable DNBR design limit from being exceeded.

The assumptions used in the analysis to maximize the pressure bias term are consistent with those described in Reference 3 and include:

(1) The event is assumed to occur due to the inadver-tent opening of the steam dump and bypass valves due to a failure of the steam dump control inter-l ock. This results in a decreasing core inlet temperature which produces an increase in core power due to the a sumption of the most negative moderator and fuel temperature coefficients dur-ing the cycle.

(2) The pressurizer control systems are assumed to be inoperative thus maximizing the rate of pressure decrease and the rate of approach to the DNBR limit.

(3) The initial axial power shape and the correspond- l ing scram worth versus insertion used in the anal-  !

I ysis is a bottom peaked shape. This power distri-bution maximizes the time required 'to terminate the decrease in DNBR following a trip.

The analysis of this event shows that a pressure bias tenn of 64.9 psia is required. This is greater than that input from the RCS Depressuriza-tion event, the other event for which a pressure bias tenn is calculated. Hence, the use of the '

pressure bias factor detennined by this event in conjunction with the TM/LP trip, will provent ex-ceeding the DNBR design limit for A00's which re-quire TM/LP trip protection.

7.0 TRANSIENT ANALYSI_S_ (Continued) 7.1 (Continued) 7.1.7 RCS Depressurization Event The RCS Depressurization event was reanalyzed for Cycle 9 to detennine the pressure bias tenn for the TM/LP set-point.

The RCS Depressurization event is one of the DBEs anal-yzed to determine the maximum pressure bias tenn input to the TM/LP trip. The methodology used for Cycle 9 is the same as that used for Cycle 8 and is described in References 3 and 4. The pressure bias term accounts for margin degradation attributable to measurement and trip system processing delt:" times. Changes in core power, inlet temperature, and RCS pressure during the transient are monitored by the TM/LP trip directly.

Consequently, with TM/LP trip setpoints and the bias tenn determined in this analysis, adequate protection will be provided for the RCS Depressurization event to prevent the acceptable DNBR design limit from being ex-ceeded.

The assumptions used to maximize the rate of pressure decrease and, consequently, the fastest approach to DNBR limits are consistent with those described in Ref-erence 3 and include:

(1) The event is assumed to occur due to an inadvert-ent opening of both pressurizer relief valves while operating at rated thennal power. This re-sults in a rapid drop in the RCS pressure and, consequently, a rapid decrease in DNBR.

(2) The charging pumps, the pressurizer heaters, and the pressurizer backup heaters are assumed to be inoperable. This maximizes the rate of pressure decrease and, consequently, maximizes .the rate of approach to the DNBR limit.

(3) The initial axial power shape and the correspond-ing scran worth versus insertion used in the anal-ysis is a bottom peaked shape. This power distri-bution maximizes the time required to terminate the decrease in DNBR following a trip.

The analysis of this event shows that a pressure bias tenn of 27.5 psia is required. This is less than that input from the Excess Load event, the other event for which a pressure bias tenn is calculated. Hence, the use of the Excess Load pressure bias term in conjunc-tion with the TM/LP trip, will provide adequate DNBR margin for this and other A00's which require TM/LP trip protection.

i 7.0 TRANSIENT ANALYSIS (Continued) 7.2 (Continued) 7.2.1 CEA Withdrawal Event.

The CEA Withdrawal event was reanalyzed for Cycle 9 to determine the initial margins that must be maintained by the LCOs such that the DNBR and fuel centerline to melt (CTM) design limits will not be exceeded in con-junction with the RPS (Variable High Power Trip or Ax-ial Power Distribution Trip).

As stated in Reference 3, the methodology of CEN-121(B)

-P, (Reference 5) is now being employed reclassifying the CEA Withdrawal event as one for which the accept-able DNBR &nd centerline to melt limits are not vio-lated by virtue of sufficient initial steady state ther-mal margin provided by the DNBR and Linear Heat Rate (LHR) related Limiting Conditions for Operations (LCOs). Depending on the initial conditions and the reactivity insertion rate associated with the CEA With-drawal, the Variable High Power Trip in conjunction with the initial steady state LCOs, prevents DNBR lim-its from being exceeded. An approach to the CTM limit is terminated by either the Variable High Power Trip or the Axial Power Distribution Trip. The analysis took credit for only the Variable High Power Trip (utilizing input from the excore detectors) in both the detemina-i tion of the required initial overpower margin for DNBR using CETOP/CE-1 and the peak linear heat generation rate for the CTM SAFDL.

I For the HFP CEAW DNBR an61ysis, an MTC identical to that utilized in Reference 5 and the gap thermal con-ductivity consistent with the assumption of Reference 3 were used in conjunction with a variable reactivity in-sertion rate. The results of the analysis were consis-tent with those obtained in Reference 5 and the assump-tions in Reference 3. The range of reactivity inser-tion rates examined is given in Table 7.2.1-1.

4 For the HFP CEAW CTM analysis, the maximum reactivity insertion rate and the most positive MTC were assumed.

The zero power case was analyzed to demonstrate that acceptable DNBR and centerline melt limits are not ex-ceeded. For the'zero power case, a reactor trip, initi-ated by the Variable High Power Trip at 29.1% (19.1%

plus 10*. uncertainty) of rated themal power, was assumed in the analysis.

The zero power case initiated at the limiting condi-tions of operation results in a minimum CE-1 DNBR of 7.06. Also, the analysis shows that the fuel-center-line temperatures are well below.those corresponding to the acceptable fuel centerline melt limit. The se-quence of events for the- zero power case is presented

7.0 TRANSIENT ANALYSIS _ (Continued) 7.2 (Continued) 7.2.1 CEA Withdrawal Event (Continued) in Table 7.2.1-2. Figures 7.2.1-1 to 7.2.1-4 present the transient behavior of core power, core average heat flux, RCS coolant temperatures, and the RCS pressure for the zero power c,ase.

Protection against exceeding the DNBR limit for a CEA Withdrawal at full power is provided by the initial steady state thermal margin which is maintained by adhering to the Technical Specification LCOs on DNBR margin and by the response of the RPS which provides an automatic reactor trip on high power level. The mini-mum DNBR for this event, when initiated from the ex-tremes of the LCOs, is 1.27.

The HFP maximum reactivity insertion rate analysis shows that the fuel ~ centerline temperatures are well below those cor responding to the acceptable CTM limit.

The sequence of events for the full power case with the l maximum reactivity insertion rate is presented in Table 7.2.1-3. Figures 7.2.1-5 to 7.2.1-8 present the trans-ient behavior of core power, core average heat flux, RCS coolant temperatures, and the RCS pressure for this full power case.

It may be concluded that the CEA withdrawal event when initiated from the Tech. Spec. LCOs (in conjunction with the Variable High Power Trip if reouf red) will not lead to a DNBR or fuel temperature which exceed the DNBR and centerline to melt design limits.

l l

a N ,

TABLE 7.2.1-1~

FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE CEA WITFDRAWAL ANALYSIS Units HZP HFP ,

Parameter.

MWt i 102% of 1500*

Initial Core Power Level Core Inlet Coolant 532* 547*

Temperature *F psia 2053* 2053*

Pressurizer Pressure Moderator Temperature +0.5**

Coefficient x10-4Ap/*F +0.5 Doppler Coefficient -

0.85 Multiplier 0.85 CEA Worth at Trip 10-2Ap -5.25 -6.66 Reactivity Insertion 0 to 1.0 Rate Range x10-4Ap/sec 0 to 1.0 CEA Group Withdrawal in/ min 46 46 Rate Holding Coil Delay sec 0.5 0.5 Time

• For the DNBR analysis the effects of uncertainties on these parameters were accounted for statistically as detailed in References 2a, 2b, and 2c.

• • DNBR analysis assumes MTC consistent with Reference 5.

E r

L TABLE 7.2.1-2

' FORT CALHOUN CYCLE 9 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM ZERO POWER ..

Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 35.0 Variable High Power Trip Signal 29.1% of 1500 MWt 4

Generated 35.4 Reactor Trip Breakers Open ---

35.9 CEAs Begin to Drop Into Core ---

36.34 Maximum Core Power 40.2% of 1500 MWt 37.21

~

Maximum Heat Flux 27.9% of 1500 MWt 37.21 Minimum CE-1 DNBR 7.06 Maximdm RCS Pressure, psia 2?32 40.1 i

i

)

1 9

g , rm, , -, .. .. - e .,.

1 4

TABLE-7.2.1-3 FORT CALHOUN CYCLE 9 SEQUENCE OF EVENTS FOR

' CEA WITHDRAWAL FROM FULL POWER (MAXIMUM REACTIVITY INSERTION RATE)

Event Fetpoint or Value Time (sec) 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion High Power Trip Signal Generated 112% of 1500 MWt

4.88 5.28 Reactor Trip Breters Open 5.78 CEAs Begin to Drop Into Core Ma.< lmum Core Power 144.48% of 1500 MWt 5.92 Maximum Heat Flux 108.99% of 1500 MWt 6.18 Maximum RCS Pressure, psia 2096 6.59

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7.0 TRANSIENT ANALYSIS (Continued) 7.2 (Continued) 7.2.2 Loss of Coolant Flow Event The Loss of Coolant flow event was reanalyzed for Cycle 9 to detennine the minimum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) such that in conjunction with the RPS (low flow trip), the DNBR limit will not be exceeded.

The event was analyzed parametrically in initial axial shape and rod configuration using the methods described in Reference 3 (which utilizes the statistical combina-tion of uncertainties in the DNBR analysis as described in Appendix C of Reference 2b).

The 4-Pump loss of Coolant Flow produces a rapid ap-proach to the DNBR limit due to the rapid decrease in the core coolant flow. Protection against exceeding the DNBR limit for this transient is provided by the .

4 initial steady state thermai margin which is maintained by adhering to the Technical Specifications' LCOs on DNBR margin and by the response of the RPS which pro-vides an automatic reactor trip on low reactor coolant flow as measured by the steam generator differential pressure transmitters.

The flow coastdown is generated by CESEC-III (Refer-ences 5 and 6) which utilizes implicit modeling of the reactor coolant pumps. This coastdown is shown in Fig-ure 7.2.2-1. Table 7.2.2-1 lists the key transient par-ameters used in the Cycle 9 analysis.

Table 7.2.2-2 presents the NSSS and RPS responses dur-ing a four pump loss of flow initiated at an axial shape index of -0.182 which bounds the DNBR related ax-ial shape index LCO. The low flow trip setpoint is reached at 1.90 seconds and the scram rods start drop-ping into the core 1.15 seconds later. A minimum CE-1 DNBR of 1.48 is reached at 3.63 seconds. Figures 7.2.2-2 to 7.2.2-5 present the core power, heat flux, core coolant temperatures , and RCS pressure as a func-tion of time.

It may be concluded that the Loss of Flow event when initiated from the Tech. Spec. LCOs in conjunction with the Low Flow Trip will not exceed the design DNBR lim-it.

i

~ . . . _ . _ _ - _ _ _ . _

TABLE 7.2.2-1 FORT.CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE LOSS OF C0OLANT FLOW ANALYSIS Parameter Units Cycle 9 Initial Core Power Lceel MWt 1500*

Initial Core Inlet Coolant Temperature 'F 545*

Initial RCS Flow Rate gpm 208,280*

Pressurizer Pressure psia 2075*

Moderator Temperature Coef ficient 10-4Ap/'F +.5 Doppler Coefficient Multiplier -- 0.85 LFT Analysis Setpoint  % of initial . flow 0.93 LFT Response Time sec 0.65 4-Pump RCS Flow Coastdown Figure 7.2.2-1 CEA Holding Coil Delay sec 0.5 CEA Time to 100% Insertion sec 3.1 (Including Holding Coil Delay)

CEA Worth at Trip (all rods out) 10-2Ap -6.87 Total Unrodded Radial Peaking 1.75-Factor (FRT)

I *The uncertainties on these parameters have been statistically combined and included in the DNBR analysis.

t t

, - -- - .. , . . _ , ..-, . - , . . ~ . _ , -

s TABLE 7.2.2-2 FORT CALHOUN CYCLE 9 SEQUENCE OF EVENTS FOR LOSS OF FLOW Time (Sec) Event Setpoint or Value 0.0 Loss of Power to all Four Reactor - ---

Coolant Pumps -

4

+

j 1.90- Low Flow Trip Signal Generated 93% of 4-Pump Flow

. 2.55 Trip Breakers Open 2

3.05 Shutdo$n, CEAs Being to Drop - ---

into Core

<9 4

3.63 Minimun CE-1 DNBR 1.48 4.14 Maximum RCS Pressure, psia 2090 I

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7.0 TRANSIENT ANALYSIS '(Continued) 7.2 (Continued) 7.2.3 Full Length CEA Drop Event The Full Length CEA Drop event was reanalyzed for Cycle 9 to detemine the initial thennal margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNSR and fuel centerline to melt design limits will not be exceeded.

This event was analyzed parametrically in initial axial shape and rod configuration using methods described in Reference 3.

Table 7.2.3-1 lists the key input parameters used for Cycle 9 and compares them to the reference Cycle 8 val-ues. Conservative assumptions used in the analysis are consistent with those discussed in Reference 3 and in-clude:

1) The most negative moderator and fuel temperature coefficients of reactivity (including uncertain-ties), because these coefficients produce the min-imum RCS coolant temperature decrease upon return to power and lead to the minimum DNBR.

') Charging pumps and proportional heater systems 1

are assumed to be inoperable during the trans-ient. This maximizes the pressure drop during the event.

3) All other systems are assumed to be in the manual mode of operation and have no impact on this event.

Table 7.2.3-2 presents the sequence of events for the Full Length CEA Drop event producing the minimum DNBR.

This event was initiated at an ASI of -0.182 with Group 4 at the PDIL and with the other cor.ditions described in Table 7.2.3-1. The transient behavior of key NSSS parameters are presented 'in Figures 7.2.3-1 to 7.2.3-4.

The transient was conservatively analyzed at full power with an ASI of -0.182, which is outside of the LC0 limit of -0.06. This results in a minimum CE-1 DNBR of 1.44. A maximum allowable . initial linear heat genera-tion rate of 17.3 KW/ft could exist as an initial con-dition without exceeding the acceptable fuel centerline.

to melt ' limit of 21 KW/ft during this transient.- This l amount of margin is assured by setting the Linear-Heat )

Rate related LCOs based on the more limiting allowable '

i linear heat rate for LOCA.  !

l

(

. It can be concluded that the CEA Drop event when initi- {

ated from the Tech.: Spec. LCOs will not exceed the DNBR and centerline to melt design limits.

TABLE 7.2.3-1 FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Units Cycle 8 Cycle 9 Parameter MWt 102% of 1500 100% of 1500*

Initial Core Power Level 547 545*

Core Inlet Temperature F psia 2053 2075*

Pressurizer Pressure gpm 197,000 208,280*

Core Mass Flow Rate Moderator Temperature Coefficient x10-4ap/*F -2.7 -2.7 1.15 1.15 Doppler Coefficient Multiplier --

% Insertion of 25 25 CEA Inserti6n at Maximem Allowed Power Bank 4 Dropped CEA Worth %Ap unrodded -0.28 -0.2261 PDIL -0.28 -0.2238 tiaximum Allowed Power Axial Shape Index at Negative Extreme of LC0 Band -0.20 -0.18 Radial Peaking Distortion Factor i

Integrated Radial Peaking Unrodded Region 1.1579 1.1585 Bank 4 1.1696 1.1557 l

Inserted Region Planar Radial Peaking Unradded Region 1.25 1.213 Bank 4 1.24 1.205 Inserted Region

• The uncertainties on these parameters were cambined statistically and included in the DNBR analysis.

'l n

)

I i

l

TABLE 7.2.3-2 FORT CALHOUN CYCLE 9-SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP Event Setpoint or Value l Time (Sec) 0.0 CEA Begins to Drop into Core 1.0 CEA Reaches Full Inserted Position 100% Inserted 1.2 Core Power Level Reaches Minimum and 71.1% of 1500 Begins to Return to Power due to Reactivity Feedbacks 95.3 Core Inlet Temperature Reaches a 540.0*F Minimum Value Reactor Coolant System Pressure Reaches 2014 180.1

_ a Nininum Value 200.0 Core Power Returns to its Maximum Value 96.13% of 1500 MWt -

200.0 Minimum DNBR is Reached 1.44 1

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FullLengthCEADrop. OmahaPublicPowerDistrict Figure.

RCSPressurevs; Time FortCalhounStation-UnitNo.1 7.2.3-4

7.0 TRANSIENT ANALYSIS (Continued)

'7.2 (Conti nued) 7.2.5 A00s Resulting from the Malfunction of One Steam Generator The transients resulting from the malfunction of one steam generator were analyzed for Cycle 9 to determine the initial margins that must be maintained by the LCOs such that in conjunction with the RPS (asymmetric steam generator protective trip) the DNBR and fuel centerline to melt design limits are not exceeded. The methods used to analyze these events are consistent with those reported in References 3 and 4. This analysis credits the addition of the Asymmetric Steam Generator Trans-ient Protection Trip Function (ASGTPTF) which provides a trip on steam generator differential pressure. The ASGTPTF is described in Section 10.

The four events which affect a single generator are:

1) Loss of Load to One Steam Generator

2) Excess Load to One Steam Generator

3) Loss of Feedwater to One Steam Generator

4) Excess Feedwater to One Steam Generator Of the four events described above, it has been deter-mined that the loss of Load to One Steam Generator (LL/1SG) transient is the limiting asymmetric event.

Hence, only the results of this transient are reported.

The event is initiated by the inadvertent closure of a single main steam isolation valve. Upon the loss of load to the single steam generator, its pressure and temperature increase to the opening pressure of the secondary safety valves. The intact steam generator

" picks up" the lost load, which causes its temperature and pressure to decrease, thus causing the core average inlet temperature to decrease and enhancing the asymme-try in the reactor inlet temperature. In the presence of a negative moderator temperature coefficient this causes an increase in core power and radial peaking.

Thus, the most negative value of this coefficient is used in the analysis. With this assum d sequence of events, the LL/1SG event results in the greatest asym-metry in core inlet temperature distribution and the i most limiting DNBR for the transients resulting from i the malfunction of one steam generator.

The LL/1SG was conservatively assumed to be initiated ,

at the initial conditions given in Table 7.2.5-1 with an axial shape index of -0.182 which bounds the DNBR

i 7.0 TRANSIENT ANALYSIS (Continued) 7.2 (Conti nued) 7.2.5 A00s Resulting from the Malfunction of One Steam ,

Generator (Continued) related axial shape index LCO. A reactor trip is gen-erated by the Asymmetric Steam Generator Trip at 3.0 seconds based on high differential pressure between the steam generators.

Table 7.2.5-2 presents the sequence of events for the Loss of Load to One Steam Generator. The transient behavior of key NSSS parameters are presented in Fig-ures 7.2.5-1 to 7.2.5-5. The minimum transient DNBR calculated for the LL/1SG event is 1.53, as compared to the acceptable CE-1 correlation DNBR limit of 1.22.

A maximum allowable initial linear heat generation rate which can exist as an initial condition without exceed-ing the acceptable fuel to centerline melt of 21 KW/ft exceeds the LOCA Linear Heat Rate LC0 and is thus less limiting for this event.

It may be concluded that the LL/1SG event when initiat-ed from the extremes of the LCO in conjunction with the ASGTPTF will not lead to DNBR or centerline fuel temper-atures which exceed the DNBR and centerline to melt de-sign limits.

{

TABLE 7.2.5-1 FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE ANALYSIS OF t.0SS OF LOAD TO ONE STEAM GENERATOR Parameter Units Cycle 4 Initial Core Power MWt 102% of 1500*

Initial Core Inlet Temperature 'F 547*

Initial Pressurizer Pressure psia 2053*

Moderator Temperature Coefficient ap/*F -2.7 x 10-4 Doppler Coefficient Multiplier 1.15

• For the DNBR at lysis the effects of uncertainties on these parameters were accounted for s, atistically as detailed in References 2a, 2b, and 2c.

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_ TABLE 7.2.5-2 ,

FORT CALHOUN CYCLE 9 SEQUENCE OF EVENTS FOR LOSS OF LOAD TO ONE STEAM GENERATOR Time (Sec) Event Setpoint or Value

! 0.0 Spurious closure of a single main -- '

steam isolation valve 0.0 Steam flow from unaffected steam --

generator increases ta maintain turbine power 3.0 ASGTPTF* Trip Signal Generated 175 psid 3.8 Safety valves open on isolated steam 1015 psia generator 3.9 Trip Breakers open 4.4 CEAs begin to drop into core --

4.7 tiinimum DNBR occurs 1.53 6.1 Maximum steam generator pressure 1051 psia ASGTPTF - Asymmetric Steam Generator Transient Protection Trip Function 1

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7.0 TRANSIENT ANALYSIS (Continu

d) 4 5

7.3 '(Continued) 1 7.3.1 CEA Ejection j' Input parameters to the CEA Ejection accident were exam-ined and found to be bounded by the previous analysis of Cycle 6. Therefore, under the guidelines of 10CFR 50.59 no reanalysis for Cycle 9 was perfomed.

7.3.2 Steam Line Break Accident This accident was evaluated using the methodology dis-cussed in References 8 and 3. The Steam Line Break accident was previously analyzed in the Fort Calhoun FSAR and satisfactory results were reported therein.

The Steam Line Break accidents at both HZP and HFP were examined in the reference cycle (Cycle 8) safety evalu-ation with acceptable results. Both the FSAR and refer-ence cycle evaluations are reported in the 1983 update- ,

i of the Fort Calhoun USAR.

~ '

The Full Power Steam Line Break accident was evgluated for a more negative effective MTC of -2.7

• 10 ap/*F than the -2.5

• 10-4ap/*F value that was used in the Cycle 8 analysis. The cooldown curves for Cycles 1, 8

and 9 are shown in Figure 7.3.2-1. This figure shows I that the reactivity insertion for the Cycle 9 core with an MTC of -2.7

• 10-4Ap/*F due to a Steam Line Break

' accident at full power is substantially less than the value used in the Cycle 8 analysis. (This snailer reac-tivity insertion is due to the use of the DIl cross-sec-

< tions which are valid for a range of moderator. tempera-tures from room temperature to 800*K while the previous l

analyses were perfomed with cooldown curves derived by

conservatively extrapolating CEPAK cross-section values -

to low temperatures.) The fuel temperature coefficient

]

used in the Cycle 8 analysis is conservative' with res-pect to the fuel temperature. coefficient calculated for.

the Cycle 9 core including uncertainties. The Cycle 9 minimum available shutdown worth is 6.91%Ap compared to a Cycle 8 value of 6.68%ap. The reduced reactivity in-sertion due to the moderator cooldown conbined with the increased minimum available shutdown worth assure that the overall reactivity' insertion for a Cycle 9 Steam i Line Break would be less than that of the reference cycle analysis. Therefore, the return to power is less ,

than that of the reference cycle and Cycle 1 FSAR anal- '

ysis.

y 1 A similar evaluation was perfomed for.the Zero Power. -1

- Steam Line Break accident. . Again the Cycle 9 cooldown ~

for an MTC- of -2.7 '* 10-4Ap/*F shows a substantially  ;

smaller reactivity insertion than was used in the Cycle ~  !

8 analysis. Since the minimum available shutdown mar ..  !

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l 7.0 TRANSIENT ANALYSIS (Continued) 7.3 (Continued) 7.3.2~ Steam Line Break Accident (Continued) gin for Cycle 9 remains unchanged from the reference cycle value (4%ap), the overall reactivity insertion for the Cycle 9 Steam Line Break accident will be i

substantially less than that of the reference cycle.

1. Therefore, the consequences of a zero power Steam Line Break accident for Cycle 9 will be less severe than that reported for the reference cycle and the FSAR (Cycle 1) cases, i '

i 4

Based on the evaluation presented above, it is conclud-si that the consequences of a Steam Line Break accident initiated at either zero or full power are less severe

.' than the reference cycle and FSAR-(Cycle 1) cases.

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I 7.0 TRANSIENT ANALYSIS (Continued) 7.3 (Conti nued) 7.3.4 Seized Rotor Event The Seized Rotor event was reanalyzed for Cycle 9 to demonstrate that only a small fraction of fuel pins are predicted to fail during this event.

The methods used to analyze this event are consistent with OPPD Fort Calhoun Unit 1 transient methodology as described in Reference 3. A DNBR limit of 1.22 was used to be consistent with the Statistical Combination of Uncertainties Program (Reference 2b).

The single reactor coolant pump shaft seizure is postu-lated to occur as a consequence of a mechanical fail-ure. The single reactor coolant pump shaft seizure re-sults in an immediate reduction in the reactor coolant

._ flow to the three pump value. This rapid reduction in core flow will initiate a reactor trip on low flow with-in the first few seconds of the transient.

The total number of pins predicted to fail was less than 1% of all of the fuel pins in the core. Based on this result, the resultant site boundary dose would be well within the lin its of 10CFR100.

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l 8.0 ECCS PERFORMANCE ANALYSIS The Loss of Coolant accident evaluation was performed using the method-ology discussed in Reference 1. The District has verified that the physics input assumptions and the maximun rod burnup are within the bounds assumed in the large break analysis for the reference cycle (Cycle 8) as reported in the 1983 update of the USAR. Therefore,.

under the guidelines of 10CFR50.59 no reanalysis for Cycle 9 was performed.

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l 9.0 STARTUP TESTING l

The startup testing program proposed for Cycle 9 is identical to the

! program outlined in the Cycle 6 Reload Application.

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10.0 RPS ASYMMETRIC STEAM GENERATOR TRANSfENT PROTECTION TRIP FUNCTION -

LICENSING DESCRIPTION 10.1 Introduction This document describes the Reactor Protection System (RPS) Asym-metric Steam Generator Transient Protection Trip Function (ASGTPTF) and its design bases. The ASGTPTF is designed to pro-tect against Anticipated Operational Occurrences (A00s) associ-ated with secondary system malfunctions which result in asymme-tric primary loop temperatures. The most limiting event is the Loss of Load to One Steam Generator (LL/1SG) caused by a single Main Steam Isolation Valve (MSIV) closure.

The Fort Calhoun RPS presently employs an analog themal margin trip calculator as part of the Themal Margin / Low Pressure (TM/

LP) trip function. To provide a reactor trip for asymmetric de-sign basis events, pressure in each of the two steam generators I will be monitored and these signals input to the ASGTPTF module housed in the thermal margin calculator. Secondary pressure im-balances between the two generators will be calculated to gener-ate a trip signal.

Protection against exceeding the DNBR and maximum Kw/Ft Specifisi Acceptable Fuel Design Limits (SAFDLs) during the LL/1SG event is presently provided by the High Pressurizer Pressure reactor trip in conjunction with sufficient initial margin maintained by the Limiting Conditions for Operation (LCOs). With installation of the ASGTPTF credit for the steam generator delta-P trip will be taken instead of the high pressurizer pressure trip.

The ASGTPTF will result in a reactor trip sooner than the High Pressurizer Pressure trip and, hence, will produce a smaller mar-gin degradaticn during this event. The additional margin gain associated with the addition of this trip function provides assur-ance that the asymmetric transients will never be limiting A00s for establishing the LCOs.

10.2 System Description The ASGTPTF cor.sists of:

1. Existing steam generator pressure sensors (one for each steam generator per channel) and the associated process equipment.

2. Existing AT Power /Themal Margin / Low Pressure Trip Calcula-tors modified to house the ASGTPTF modules. (The TM/LP trip function remains unaffected by the addition of ASGTPTF).

3. Existing RPS Trip logic and Reactor Trip Switchgear.

4. ASGTPTF

a. The existing system provides a separate bistable trip unit for monitoring each Steam Generator Pressure Sig-nal (Figure 10-1). A single module is now added to the TM/LP Calculator (Figure 10-2) to select the low-

10.0 RPS ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION -

LICENSING DESCRIPTION (Continued) 10.2 System Description (Continued) i

4. ASGTPTF (Continued)

~

a. (Continued) est of the two steam generator pressures. This pres-sure value is input to the existing bistable trip unit fonnerly used to monitor Stean Generator 1 pres-sure (Figure 10-1). In Figure 10-2 " MAX" corresponds to the highest voltage which correlates to the lowest i steam gevierator pressure.

b. 'The Steam Generator 1 and 2 pressure signals are in-put to the AT Power /TM/LP Calculator Assembly (Figure 10-2) where the absolute value of the difference be-

' tween the two pressures is calculated. This pressure value is input to the existing bistable trip unit for-merly used to monitor Steam Geneator 2 Pressure (Fig-ure 10-1). If the difference in Steam Generator pres-sures exceeds a predetermined setpoint value, a reac-tor trip occurs.

c. The ASGTPTF function does not require an operating bypass.

10.3 Design Bases

1. Design Basis Events The ASGTPTF is designed to provide a reactor trip for those design basis events associated with secondary system mal-functions which result in asymmetric primary loop coolant temperatures. The most limiting event is the LL/1SG caused by a single Main Steam Isolation Valve (MSIV) closure.

l 2. Design Criteria The ASGTPTF is designed to the following criteria to ensure adequate performance of its trip function:

a. The trip function is designed in compliance with the applicable criteria of the General Design Criteria for Nuclear Power Plants, Appendix A of 10 CFR 50, July 15,1971.

b. Instrumentation, function and operation of the trip logic confom to the requirements of IEEE Standard 279-1968, Criteria for Protective Systens for 51uclear Power Plants.

c. The trip function is designed consistent with the re-commendations of Regulatory Guide 1.53, Application .

10.0 RPS ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION -

J.ICENSING DESCRIPTION (Continued) 10.3 DesignBases(Continued)

2. Design Criteria (Continued)

c. (Continued) of the Single-Failure Criterion to Nuclear Power Plant Protective Systems, and Regulatory Guide 1.22, Periodic Testing of Protective System Actuation Func-tions.

1

d. Four independent measurement channels are provided.

e. The protective system AC power is supplied from our separate vital instrument buses,

f. The ASGTPTF can be tested with the reactor in opera-tion or shut down.

g. Trip signal is preceded by a pretrip alarm to alert the operator of undesirable operating conditions in cases where operator action can correct the abnormal condition and avoid a reactor trip.

h. The ASGTPTF, which is of the type that has successful-ly been incorporated into Baltimore Gas and Elec-tric's Calvert Cliffs Units I and 2 and Florida Power and Light's St. Lucie Units 1 and 2 reactor protec-tive systems, will meet the same industry standard as applied to the original RPS (i.e., IEEE-279, August 1968).

i. The trip function is designed so that protective ac-tion will not be initiated due to normal operation of the generating station.

J. All equipment will be designed in accordance with the QA0M. Vendor quality control will be in accordance with CE Procedure WQC 11.1, Revision 0.

k. Modification to the TM/LP Calculator for the ASGTPTF will not jeopardize previous qualification of this equi pment.

3. Perfomance Requirements The selection of a trip setpoint is suen that adequate pro-tection is provided when all sensor and processing time de-lays and inaccuracies are taken into account.. Final deter-mination of an equipment setpoint f s based on equipment 1 characteristics, operating environrnent, NSSS perfonnance and safety analysis. The setpoint, uncertainties and res-ponse time are provided in Table 10-1.

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TABLE 10-1 ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION CHARACTERISTICS I

System Accuracy 140 psi Analysis Setpoint +175 psid Technical Specification Setpoint +135 psid Nominal Pretrip Setpoint +100 psid .

._.9 seconds Systen Response Time 6

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FIGURE 10-1 Existing Trip Hardware SG 1 Pressure _

B/S TU-6 Process Loop SG 1 Low Press.

Fixed S.P. I To 2/4

' Trip Logic SG 2 Pressure -

B/S TU-7 Process Loop ,SG 2 Low Press.

Y Fixed S.P.

FIGURE 10-2 Modified Trip Hardware to Include ASGTPTF i

l AT-Power /TM/LP CALCULATOR SG 1 Pressure Process Loop A. \

MAX.f =

B/S TU-6 SG Low Press.

/

SG 2 Pressure Fixed i l

3,p, Process Loop '

~~ \dp~ To 2/4 Trip Logic n $'

ABSOLUTE -

AP B/S TU '

ASGTPTF (SG 1 Press.-SG 2 Fress.) Fixed S.P.

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11.0 REFERENCES

References (Chapters 1-5)

' 1. Letter from W. C. Jones to J. R. Miller, LIC-83-283, dated November 11, 1983.

2. OPPD-liA-8301-P, " Reload Core Analysis Overview", September 1983.

3. OPPD-NA-8302-P, "Neutronics Design Methods and Verification",

l September 1983.

4. OPPD-NA-8303-P, " Transient and Accident Methods and.Verifica-tion", September 1983.

5. Letter from W. C. Jones to J. R. Miller, LIC-83-246, dated September 26, 1983.

6. " Generic Mechardcal Design Report for Exxon Nuclear Fort Calhoun 14 x 14 Reload Fuel Assembly," XN-NF-79-70-P, September 1979.

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3

11.0 REFERENCES

(Continued)

References (Chapter 6) ,

1. CENPD-161-P, " TORC Code, A Computer Code for Detennining the Themal liargin of a Reactor Core", July 1975.

2. CENPD-162-PA (Proprietary) " Critical Heat Flux Correlation For CE-Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," April 1975.

3. CEN-191(B)-F "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 a 2, December 1981.

4. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 70 to Facility Operating License No.

DPR-40 for the Omaha Public Power District, Fort Calhoun Station, Unit No.1 Docket No. 50-285, March 15,1983.

5. OPPD-NA-8301-P, " Omaha Public Power District Nuclear Analysis Reload Core Analysis Methodology Overview", September 1983.

6. CEN-257(0)-P " Statistical Combination of Uncertair. ties, Part 2", November 1983.

11.0 REFERENCES

(Continued)

References (Chapter 7)

1. " Amendment to Operating License DPR-40, Cycle 8 License Applica-tion", Docket No. 50-285, November 22, 1982.

2a. " Statistical Combination of Uncertainties tiethodology, Part 1:

Axial Power Distribution and Thermal Margin / Low Pressure LSSS for Fort Calhoun", CEN-257(0)-P, November 1983.

2b. " Statistical Combination of Uncertainties Methodology, Part 2:

Combination of System Parameter Uncertainties in Thennal Margin Analysis for Fort Calhoun Unit 1", CEN-257(0)-P, November 1983.

2c. " Statistical Combination of Uncertainties Methodology, Part 3:

Departure from Nucleate Boiling and Linear Heat Rate Limiting l

Conditions for Operation for Fort Calhoun", CEN-257(0)-P, November, 1983.

3. _ "Omaha Public Power District Reload Core Analysis Methodology -

Transient and Accident Methods and Verification", OPPD-NA-8303-l P, September 1983.

4. "CE Setpoint Methodology", CENPD-199-P, Rev.1-P, March 1982.

5. "CEA Withdrawal Methodology", CEN-121(B)-P, November 1979.  !

6. "CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System", Enclosure 1-P to LD-82-001, January 6, 1982.

7. " Response to Questions on CESEC", Louisiana Power and Light Company, Waterford Unit 3, Docket 50-382, CEN-234(C)-P, 1 December 1982. ]

8. "0maha Public Power District Reload Core Analysis Methodology -

Neutronics Design Methods and Verification", OPPD-NA-8302-P, i September 1983. l t

. - . . .. ... . . .- - - - - . . _ . - - - - -. . . . _ . _ - =- -.. - .

t 4-

11.0 REFERENCES

(Continued)

References (Chapter 8)

1. " Omaha Public Power District Reload Core Analysis Methodology -

l Transient and Accident Methods and Verification", OPPD-NA-8303-P, September 1983.

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JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be Class III, within the mean-

ing of 10 CFR 170.22, in that it involves a single safety concern i

and has been deemed not to involve significant hazards consider-ations.

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ATTACHMENT C

....._ __ . _ _ - _ ~ , _ . _ . . . . _ . _ . _ . - . . . _ . . . . . . . . . . . _ .