ML20135A947

Proposed Tech Specs Re Cycle 10 Operation

ML20135A947
Person / Time
Site:	Fort Calhoun
Issue date:	09/06/1985
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
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References
NUDOCS 8509100326
Download: ML20135A947 (100)

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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 opera-tion, normal operational transients, and anticipated tran-sients is limited to 1.18. A DNBR of 1.18 corresponds to a l

95% probability at a 95% confidence level that DNB will not occur, which is considered an appropriate margin to DNB for all operating conditions.(1)

The curves of Figure 1-1 represent the loci of points of re-actor thermal power (either neutron flux instruments or AT in-struments), reactor coolant system pressure, and cold leg temperature for which the DNBR is 1.18.

The area of safe opera-l tion 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 1.85.

l The LSSS in Figure 1-3 is based on the assumption that the un-rodded integrated total radial peak (F ) is 1.80.

This peak-1 ing factor is slightly higher (more co servative) than the maximum predicted unrodded total radial peak during core life, excluding measurement uncertainty.

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

The reactor protective system is designed to prevent any anticipated combination of tran-sient conditions for reactor coolant system temperature, pres-sure, and thermal power level that would result in a DNBR of less than 1.18.(3) k_

References-1 (2))

USAR, Section 3.6.7

(

USAR, Section 1.4.6 (3) USAR, Section 3.6.2 Amendment No. 8, 32, #3, 47, 70, 1-2 77 ATTACHMENT A m

285 PDR p

PDR

_ _ _ _ _ _ _ - _ - _ - - - _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ _ - - _ _ - _ _ _ = - - - _ _ _ _

590 I

l 3

g i

580 uI E

570 6

F-2400 psia 560 M

2250 psia 2075 psia o

550 g

E 8

540 3

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

590 i

i i

i i

,u.

580 ui$

i Q

90 E-W 580 t-2400 psia H

550 g

2250 psia 58 540 2075 psia -

E O

530 1750 psia 520 60 70 80 90 100 110 120 1

CORE POWER (% OF RATED POWER)

Pyg = 22 PF(B) Ai(Y)B+22.iT

-12674 m

PF (B) = 1.0 B>iOO%

=

.000 8 + i.B 50%<B<100%

= 1.4 8550%

A1 (Y) =

.5Y1 + 1.125.

1 Y

1.25

=.5Y 1 +.875 Y1 >.25 ThercalHargin/LowPressureLSSS DeahaPublicPowerDistrict figure 4PumpOperation FortCalhounStation-UnitNo.i 1-3 Acundment No. 3. N, 47, 70', 77 L.

./

1.0 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%

1 rated power, a reactor trip will occur in the event of a reactivity excursion that results in a ' power increase t

up to the lower fixed set point of the VHPT circuit of 19.1% of rated power.(3) During normal power increases be-low 19.1% reactor trip would be initiated at 19.1% of rated power unless the set point is manually adjusted.

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

Flow in each of the four coolant loops is determined from a measurement of pressure drop from inlet to out-

.let of the steam generators. The total flow through the reactor core is measured by summing the loop pres-sure drops across the steam generators and correlating this pres ure sum with the pump calibration flow 1

Curves.

1 The percent of normal core flow as follows:(6) 4 Pumps 100%

J 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.

5 i

1

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Amendment No. 7, 32, 79, 77 1-7

'l i

1

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-ators. The power-operated relief valves are set to operate concurrently with the high pressurizer pressure reactor trip. This setting is 100 psi below the nominal safety valve setting (2500 psia) to avoid un-necessary operation of the safety valves. This setting is consistent with the trip point assumed in the ac-cident analysis.(1)

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

low pressure trip is provided to prevent operation when the DNBR is less than 1.18, including allowance for j

measurement error. The thermal and hydraulic limits shown on Figure 1-3 define the limiting values of re-actor coolant pressure, reactor inlet temperature, axial l

shape index, and reactor power level which ensure that the thermal criteria (8) are not exceeded. The low set point of a 1750 psia trips the reactor in the unlikely event of a loss-of-coolant accident. The thermal margin /

low pressure 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.

T

= Core inlet temperature. *F.

IN

= Reactor pressure, psia.

P VAR = Axial Shape Index, asiu Y y Amendment No. E, 29, 32, #7, 79, 1-8 77 i

d

1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limiting Safety System Settinos, 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 DNBR of less than 1.18 nor a maximum linear heat rate l

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 1.18 nor a peak linear heat rate of more than 21 kW/ft occurs as a result of the loss of load to one l

steam generator.

(10) Physics Testing at low Power - During physics testing at power levels less than 10~1% 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 Ccemittee will be in effect during these tests. At reactor power levels less than 101% 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 (6) USAR, Section 14.7 (2) USAR, Section 7.2.3.3 (7) USAR, Section 7.2.3.1 (3) USAR, Section 7.2.3.2 (8) USAR, Section 3.6 (4) USAR, Section 3.6.6 (9) USAR, Section 14.10 (5) USAR, Section 14.6.2.2, 14.6.4 Amendment No. 7, R 70, 77 1-9 J

o TABLE 1-1 RPS LIMITING SAFETY SYSTEM SETTINGS No.

_ Reactor Trip

_ Trip Setpoints 1

High Power Level (A) 4-Pump Operation

<107.0% of Rated Power 2

Low Reactor Coolant Flow (B)(F) 4-Pump Operation

>95% of 4 Pump Flow 3

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

-4 Low Steam Generator Pressure (C)

>500 psia 5

High Pressurizer Pressure 52400 psia 6

Thermal Margin / Low Pressure (B)(F) 1750 psia to 2400 psia

,j (depending on the re-actor coolant temper-ature as shown in Figure 1-3) 7 High Containment Pressure (D)

<5 psig i

8 Axial Power Distribution (E)

(Figure 1-2) 9 Steam Generator Differential Pressure

$135 psid Amendment No. 7, 32 A7, 77 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 temperature is less than 50 F above that of the reactor coolant system cold leg.

l l

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

(b)

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

(c) If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

normal operations and anticipated transients.

In the hot shutdown mode, a single reactor coolant loop provides sufficient l

heat removal capability for removing decay heat; however, single failure considerations require that two loops be operable.

l In the cold shutdown mode, a single reactor coolant loop or shutdown cooling l

loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be operable.

Thus, if the reactor coolant loops are not operable, this specification i

requires two shutdown cooling pumps to be operable.

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

(1) sufficient cooling capacity is available to l

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

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

Amendment No. 56, 9/#f (/29/81, 79, 2-2b 77 L

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90 g

- (-0.04,83.0)

(0.ce,83.0) i g

r. 80 E

3 70 x

3

(-0.2,65.0)

(0.2,65.0) d 60 u_

o a

50 e

?

40 e

E 30 t

8 20 4

10 0

-0.3

-0 2

-0.1 0.0 0.1 0.2 0

AXIAL SHAPE INDEX YI 1

)

LimitingConditionforOperationfor OmahaPublicPowerDistrict Figure ExcoreMonitoringofLHR FortCalhounStation-UnitNo.I 2-6 Amendment No. 8, 29, 32, (), A7, 7g, 77

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FiLIMIT 100 FTLIMIT n

5 E

80

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1 60 l

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1.75 1.80 f.85 1.90 1.95 2.00 2.05 FTAND F i n

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F,ifjandCarePower Omaha Public Pol'er District figure j'

rtCalhounStation-UnitNo.1 2-9 Amendment tto g, gg, pg, g3, 97, pp, y7

?

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

(ii) Be in at least' hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(2) Total Integrated Radial Peaking Factor T

The calculated value of F d T,p (1+T)shallbelimitedtb<efinedbyFi0detbrmined l

1.80.

F 9

fromapowerdistributionmapwithn0partlength

]

CEAs inserted and with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, T, is the measured value of T at the time F isde9 ermined.

q R

I With Fp >1.80 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

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

(3) Total Planar Radial Peaking Factor The calculated value of F T defined as F T=F (1+T shallbelimitedtdY<1.85.

F shlY1beEl-termin)edfromapowerdistributionm5%withnopart l

4 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-inclusive and shall exclude regions influenced by grid effects. The azimuthal tilt, T, is the measured value of T at the time F is 8etermined.

q xy With F T >1.85 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

xy (a) Reduce power to bring power and F T to with-in the Ifmits of Figure 2-9 withdfaw 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 i

(b) Be in at least hot standby.

Amendment No. 32, A3, A7, 79, 77 2-57a

,, e a

i 2.0 LIMITING CONDITIONS FOR OPERATION

-2.10 Reactor Core (Continueol 2.10.4 Power Distribution Limits' (Continued)

I i.

(4) Azimuthal Power Tilt (Tq)'

i When operating above 70% of rated power, (a) The azimuthal power tilt (Tq) shall not exceed 0.10 whenever Mini CECOR/BASSS is operable,.the CEA's are at or above the Long Term Insertion Limit and Mini T

CECOR/BASSS is-being utilized to monitor,F T and F '

xy R

(b) The azimuthal power tilt (Tq) shall not exceed 0.03 whenever the provisions of 2.10.4(4)(a) do NOT allow M{ni CECOR/BASSS to be utilized to monitor F*determit T and F

I t0. With the indicated azimuthal power tilt be >0.03 but <0.10, correct the power tilt within two hours or determine within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per subsequent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, } hat the total integrated radial peaking factor, F is within the limit of Specification 2.10.4(2) an0,that the total-planar radial peaking factor, F T is within the limit of 2.10.4(3), or reduce power t3Y,ss than 70% of rated le power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of confirming Tg >0.03.

(c) With the indicated power tilt' determined to be >.10, p wer operation may proceed up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> proviHed F and F T do not exceed the power limits of Figure 2 9, or EE in at least hot standby within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Subsequent operationgfor the purpose of measurement to identify the cause of the tilt is allowable provided the power level is' restricted to 20% of the maximum allowable. thermal power level for the existing reactor coolant pump combination.

i 1

f f

! Amendment No.'J2, #3, 47 2-57b a l/fy y

i s.

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E m

E 2.0 LIMITING CONDITIONS FOR OPERATION f

2.10 Reactor Core (Continued)

'd 2.10.4 Power Distribution Limits (Continued)

=

g (5) DNBR Marcin Durino Power Operation Above 15% of Rated Power

(

(a) The following DNB related parameters shall be maintained within the limits shown:

g (i)

Cold Leg Temperature

< 540 F*

r (ii) pressurizer Temperature

> 2075 psia

• L (iii) Reactor Coolant Flow

> 197,000 gpm**

l (iv)

Axial Shape Index Y iFigure2-7***

g (b) With any of the above parameters exceeding the limit, restore the t

parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to lest 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 />.

a*

Basis i

E.

Linear Heat Rate 4"

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

E Either of the two core power distribution monitoring systems, the Excore i

Detector Monitoring System, or the Incore Detector Monitoring System, provide 4

adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector f-Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable limits of Figure 2-6 as adjusted by Specification 2.10.4(1).(c) for the allowed

?

linear heat rate of Figure 2-5, RC Pump configuration, and F T of Figure 2-9.

[

InconjunctionwiththeuseoftheexcoremonitoringsystemEXdinestablishing b

the axial shape index limits, the following assumptions are made:

(1) the CEA insertion limits of Specification 2.10.2(6) and long term insertion limits of Specification 2.10.2(7) are satisfied, (2) the flux peaking augmentation L

factors are as shown in Figure 2-8, (3) the total planar radial peaking factor l

does not exceed the limits of Specification 2.10.4(3).

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

I of rated thermal power.

_(

• • This number is an actual limit and corresponds to an indicated flow rate of 202,500 gpm. All other values in this listing are indicated values p

and include an allowance for measurement uncertainty (e.g., 540 F, indicated, allows for an actual T of 542 F).

• • • The AXIAL SHAPE INDEX.

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

FORT CALHOUN 2-57c Amendment No. 32, 43, E7, 70, 77

3.0 SURVEILLANCE REQUIREMENTS 3.10 Reactor Core Parameters (Continued)

(6) Azimuthal Power Tilt (Tq)

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

The excore detectors with at least four safety channels a.

operable, or b.

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

(7) DNB Parameters The cold leg temperature, pressurizer pressure, and axial a.

shape index shall be verified to be within the limits of Section 2.10.4(5) at least once per shift.

b.

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

I Amendment flo. 32, 76 3-63b

JUSTIFICATION, DISCUSSION, AND SIGNIFICANT HAZARDS CONSIDERATIONS FOR CYCLE 10 RELOAD The Fort Calhoun Technical Specifications are being amended to reflect enanges which are a result of the Cycle 10 core reload. Table B-1 presents a summary of the Technical Specification changes and the explanation for the changes.

Justification for the changes is contained in the attached Fort Calhoun Cycle 10 Core Reload Evaluation.

Significant Hazards Considerations:

It has been detemined, based on the analytical infomation supplied in the Cycle 10 Core Reload Evaluation, that this amendment request does not involve a significant hazards consideration. This conclusion was derived by applying the Commission's guidance for implementation of 10 CFR 50.92. The Commission provided this guidance concerning the application of these standards through certain examples in the Federal Register, Volume 48, Number.67, Wednesday, April 6,1983, Rules and Regulations.

Example 111 of actions involving no significant hazards considerations, 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 different from those found previously acceptable to the NRC for a previous core at the facil-ity in question are involved. This 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 10 Core Reload Evaluation, no fuel assemblies to be loaded into the Cycle 10 core will be of new or different design than those used previously and found to be acceptable to the NRC. No proposed change to the Technical Specifications for Cycle 10 involve acceptance criteria which are significantly different fran those previously found acceptable to the NRC.

The minimum acceptable DNBR limit has been decreased to 1.18 from 1.22 (using the CE-1 correlation) to be consistent with both the NRC Staff's Safety Evalu-ation Report approving a 1.15 CE-1 correlation limit (rather than the 1.19 interim value) and also in accordance with the employment of the Statistical Combination of Uncertainties methodology used by the District since Cycle 9.

l The analytical methods used to demonstrate confomance with the Technical Specifications and regulations are consistent with previous NRC approvals (or documented in report OPPD-NA-8301-P, Rev. 01, submitted to the Commission in June,1985) and involve no significant changes.

It is concluded that the proposed license amendment does not include signifi-cant hazards considerations in that:

1.

The probability or consequences of accidents previously evaluated are not increased. All events / accidents not enveloped by Cycle 9 parameters were evaluated and shown to have acceptable consequences, with violation of no safety limits.

ATTACHMENT B

2.

The Cycle 10 core reload does not create the possibility of a new or dif-ferent kind of accident from any previously evaluated. The core loading utilizies fuel management techniques which have previously been proven acceptable.

3.

The Cycle 10 core reload does not result in a significant reduction in the margin of safety because the Cycle 10 reload evaluation, which uses NRC approved methodologies, demonstrates that the margin of safety is maintained in the revised Technical Specifications limits.

l 1

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TABLE B-1 Explanation for Cycle 10 Technical Specification Changes Tech. Spec. Number Changes Reasons 1.

1.1,1.3(2),1.3(4),

Change the minimum The CE-1 limit for 14 x 14 fuel 1.3(8),1.3(9),2.1.1 DNBR value from 1.22 was approved by the NRC at 1.15.

Pg. 1-2, 1-7, 1-8, to 1.18.

The SCU analysis was revised to 1-9, 2-2b reflect the approved limit.

2.

1.1 Change the total un-The unrodded planar radial peak Pg. 1-2 rodded planar radial is being raised for Cycle 10.

peak from 1.78 to 1.85.

3.

1.1 Change the unrodded The unrodded integrated radial Pg. 1-2 integrated radial peaking factor is being raised peaking factor fran for Cycle 10.

1.73 to 1.80.

4.

Figure 1-1 Replace Figure 1-1 The TM/LP safety limits have with enclosed Figure been changed to reflect changes 1-1.

in peaking factors and inclusion of ASI input into the TM/LP cal-cula to rs.

5.

Figure 1-3 Replace Figure 1-3 The TM/LP trip LSSS equation has with enclosed Figure been adjusted to reflect ASI in-1-3.

put into the TM/LP calculators.

(Item 11, Reload Evaluation).

6.

1.3.(2) - Pg. 1-7 Delete references to The Fort Calhoun License is lim-less than 4-Pump oper-ited to 4-Pump operation.

ation.

7.

1.3(4)

Include axial shape The modified TM/LP calculators Pg. 1-8 index as a thermal-monitor ASI.

hydraulic parameter.

8.

Table 1-1 Delete references to The Fort Calhoun License is lim-No. I and 2 3-and 2-Pump Oper-ited to 4-Pump operation.

Pg. 1-10 tion.

9.

Figure 2-6 Replace Figure 2-6 The LHR excore LCO has been with Enclosed Figure changed to reflect higher radial 2-6.

peaking factors.

10. Figure 2-9 Replace Figure 2-9 The FxyT and F T limits have been R

with enclosed Figure changed to reflect higher peaking 2-9.

fa' to rs. The asymetric loading pattern impacted the shape of the FxyT limit line.

i a

11. 2.10.4(2)

Change limited to The F T changes have been made to R

Pg. 2-57a

< 1.73 to limited reflect proposed changes in Tech.

to < 1.80 and wi th Spec. 1.1.

F T > 1.73 to with R

F T }[ 1.80.

R

12. 2.10.4(3)

Change limited to The FxyT changes have been made to Pg. 2-57a

< 1.78 to limi ted reflect proposed changes in Tech.

to < 1.85 and wi th Spe c. 1.1.

F T > 1.78 to wi th R

F T }[ 1.85.

R

13. 2.10.4(4)

Mi ni-CECOR/BASSS Adds a change in tilt when Mini-Pg. 2-57b Tilt Limits CECOR/BASSS is being used to mon-itor Technical Speci fications.

14. 2.10.4(5)(a)(i)

Change < 545*F to The Cold Leg Tenperature limits Pg. 2-57c

< 540*FT have been changed from 545'F to 540*F.-

15. 2.10.4(5)(a)(iv)

Add *** footnote to The Fort Calhoun Station has add-Pg. 2-57c identify incore mon-ed incore DNB LCO monitoring sys-itoring with Better tem in addition to the excore LCO.

Shape Selection Sys-ten.

16. 3.10(6)a DELETE two symmetric Asymmetric fuel loading pattern Pg. 3-63b safety channels and for Cycle 10 prevents this com-two symmetric control bination from properly detecting channels.

azimuthal tilts.

FORT CALHOUN UNIT 1 CYCLE 10 1

RELOAD EVALUATION l

1

Fort Calhoun -

Cycle 10 License App 1tcation CONTENTS 1.

INTRODUCTION AND

SUMMARY

2.

OPERATING HISTORY OF THE REFERENCE CYCLE

-3.

GENERAL DESCRIPTION 4.

FUEL SYSTEM DESIGN 5.

NUCLEAR DESIGN 6.

THERMAL-HYDRAULIC DESIGN a

7.

TRANSIENT ANALYSIS 8.

ECCS PERFORMANCE ANALYSIS 9.

STARTUP TESTING 10.

Mini-CECOR/BASSS LC0 Monitoring System i

11.

TM/LP MODIFICATION ~

12.

REFERENCES.

l

1.0 INTRODUCTION

AND

SUMMARY

i This report provides an evaluation of the design and performance for the operation of Fort Calhoun Station Unit 1 during its tenth fuel cycle at full rated power of 1500 MWt. All planned operating condi-

~ tions remain the same as those :for Cycle 9 with the exception of a i

reduction in core inlet temperature from 545'F to 540*F. This reduc-tion is being made to minimize the susceptability to steam generator "U" tube stress corrosion cracking..The reduction in inlet tenpera-ture was implemented by-Operations at the startup of Cycle 9, so this i

Technical Specification change for Cycle 10 reflects controls already in effect.

The core will consist of 68 presently operating J and K assemblies, j

44 fresh Batch L assemblies and 12 G and 9 H assemblies discharged from previous cycles.

The Cycle 10 analysis is based on a Cycle 9 tennination point between 12,500 MWD /T to 13,500 MWD /T.

In performing analyses of design basis events, detennining limiting safety settings and establishing limit-ing conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 10 conditions would be envel-oped, provided the Cycle 9 termination point falls within the above i-burnup range. The analysis presented herein will accommodate a Cycle 10 length of up to 13,000 MWD /T.

The evaluation of the reload core characteristics have been conducted 7

i with respect to the Fort Calhoun Unit No.1 Cycle 9. safety analysis i

described in the 1985 update of the USAR, hereaf ter referred to as

]

the " reference cycle" in this report unless otherwise noted.

Specific core di fferences have been accounted for in the present anal-i ysis.

In all cases, it has been concluded that either the reference l

cycle analyses envelope the new conditions or the revised analyses i

presented herein continue to show-acceptable results. Where dictated by variations from the previous cycle, proposed modifications to the t

j:

plant Technical Specifications have been provided.

l The Cycle 10 core has been designed to further reduce fluence to crit-i{

ical reactor pressure vessel welds through the use of part-length poi-son rods inserted in the CEA guide tubes of selected peripheral assem-

[

blies and thus minimize the RTPTS shift of these welds. This will de-1ay the time when the reactor vessel welds reach the Pressurized Ther-mal Shock RTpis screening criteria contained in 10 CFR 50.61. The Cycle.10 use of a-low radial leakage core design has resulted in in-1 creased radial peaking factors. The increased peaking factors have been accommodated in the~ safety analysis through the continued use of i

the Statistical Combination of Uncertainties methodology (Reference i

1), incorporation'of axial shape index as an input to the Thermal Margin / Low Pressure Trip Function, and the use of a Mini-CECOR/Better

~

Axial Shape Selection' System (BASSS)'for incore monitoring of the linear heat rate and DNB LCO's. The Mini-CECOR/BASSS methodology change is an improvement over the previous excore monitoring system i

and will be installed during the upcoming outage. A full description is supplied in Section 10.

i

The analysis presented in this report was perfonned utilizing the methodology documented in the District's reload analysis methodology reports (References 2, 3 and 4). - These methodologies were previously transmitted in References 5 and 6.

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2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE Fort Calhoun Station is presently operating in its ninth fuel cycle util-izing Batch G, H, I, J and K fuel assemblies. Fort Calhoun Cycle 9 oper-ation began on July 8,1984, and reached full poer on August 3,1984.

The reactor has operated up to the present time with the core reactiv-ity, power distributions and peaking factors having closely followed the calculated predictions.

It is estimated that Cycle 9 will be tenninated on or about Septenber 27, 1985. The Cycle 9 termination point can vary between 12,500 MWD /T 4

and 13,500 MWD /T and still be within the assumptions of the Cycle 10 analyses. As of August 1,1985, the Cycle 9 burnup had reached 11,311 MWD /T.

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3.0' GENERAL DESCRIPTION l'

The Cycle 10 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 10 is the addition of part-length poison rods in selected peri-1 pheral assemblies for power reduction to minimize the reactor vessel weld RTNDT shifts. The part-length poison rods are essentially full 4

strength CEA fingers active only in the middle 50% of the core height.

2 Four G assemblies, 21 N assemblies, and 40 I assemblies will be dis-charged this outage. They will be replaced by 24 fresh unshimmed Batch i

L assemblies (3.80 w/o enrichment),12 fresh shimmed Batch L assenblies l

(3.80 w/o enrichment, 0.01904 gm B o/ inch), eight fresh shimmed Batch L i

assemblies (3.80 w/o enrichment, 0.01190 gm B10/ inch), twelve Batch G assemblies (3.03 w/o initial enrichment) discharged from Cycle 8, and nine Batch H assemblies (3.50 w/o initial enrichment) discharged from i.

Cycle 8.

l Figure 3-1 shows the fuel management pattern to be employed in Cycle 10 l

including the location of the part-length poison rods. The location j

within an assembly of the part-length poison rods is shown in Figure j

3-2.

Figure 3-3 shows the locations of the poison pins within the lat-tice of shimmed assemblies and the fuel rod locations in unshimmed assemblies.

Figure 3-4 shows the beginning of Cycle 10 assembly burnup distribution i

for a Cycle 9 termination burnup of 13,000 MWD /T. The average discharge i

exposure at the End of Cycle 9 fuel is projected to be 35,239 MWD /T.

The initial enrichment of the fuel assemblies is also shown in Figure i,

3-4.

Figure 3-5 shows the end of Cycle 10 assembly burnup distribution.

l The end of Cycle 10 core average exposure is approximately 27,500 MWD /T.

i 4

4 1-i l

1-r

,7-c v

,e,,-%w,,

~y

.e,-.,

p.._,-,y-<

.9.-.m----,

,,,..m,,,,,.,

r..g

--e

,,r,~

.,e,-

,y

.,7---ai-e -w

l Table 3-1 Fort Calhoun Cycle 10 Core Loading

. Initial BOC Batch EOC Batch Poison Poison

. Assembly Number of Average Burnup (MWD /T)

Average Burnup (MWD /T)

Rods per Loading l

Designation Assembli es

[E0C 9 = 13,000 MWD /T]

[EOC 10 = 12,500 MW/T]

Assenbly gm B o/ inch l

i l

G(1) 12 35,038 38,952 0

0 HIII 9

29,421 33,838 0

0 J

28 24,491 37,242 0

0 J(2) 8 14,407 29,093 0

0 K

-12 12,429 27,468 0

0 K/

20 17,643 30,796 8

.0238 L

24 0

13,247 0

0 L/

12 0

14,675 8

.01904 L*

8 0

17,010 8

.01190 TOTAL 133 (1) Assemblies Discharged From Cycle 8 (2) Assemblies Delivered for Cycle 8, But First Loaded Into Cycle 9 l

Figure 3-1 FORT CALHOUN CYCLE 10 CORE LOADING PATTERN AA

-ASSEMBLY LOCATION 01 02 BB

= FUEL TYPE H

H 6

hPART-LENGTH XX POISON RODS XX XX OS 04 05 06 07 G

L L

L J

XX 08 09 10 11 12 IS G

L L/

J K/

L/

XX 14 l' 5 16 l 17 18 19 G

L+

K/

J+

K/

J l

20 21 22 23 24 25 L

J*

K K

J L/

26 K/

27 28 29 30 31 32 L+

J K/

J K

J 33 L

34 35 36 ST 38 39 J

L/

J L/

J H

I

Figure 3-2 FORT Call 10Uti CYCLE 10 PART LErlGTH POIS0TI R00 LOCATI0115 4 PART LENGTH POISON ROD ASSEMBLY ASSEMBLY LOCATIONS i,2,3 S 8 IN OUARTER CORE X

X I

l X

X

]

FUEL A00 LOCATION POISON R0D LOCATION

Figure 3 FORT CALHOUN CYCLE 10 ASSEMBLY FUEL AND POISON R00 LOCATIONS

[

UNSHIMMED ASSEMBLY i

P 4

.e l

l

\\

K/,L/ D - 8 POISON ROD ASSEMBLY i

X x

X X

r I

X X

i r

X X

f i

FUEL ROD LOCATION K

POISON ROD LOCATION

l Figure 3-4 FORT CALHOUN CYCLE 10 i

80C10 ASSEMBLY AVERAGE EXPOSURE AND INITIAL ENRICHMENT AA

-ASSEMBLY LOCATION 01 02 88

- FUEL TYPE H

H C.CC

- ENRICHt1ENT,

W/O U-235 3.50 3.50 DDDDD -ASSY AVG EXPOSURE, MMD/T 29984 29929 OS 04 05 06 07 G

L L

L J

3.03 3.80 3.80 3.80 3.50 88082 0

l 0

0 24893 l08 09 10 11 12 13 G

L L/

J K/

L/

S.08 S.80 S.80 9.50 9.50 S.80 38953 0

0

26785, 15430, 0

i 14 15 16 17 18 19 G

L+

K/

J*

K/

J 8.08 S.80 3.50 3.50 3.50 3.50 S3077 0

18412 14985{

17962 26253 20 21

'22 i 23 24 25 L

J+

K K

J L/

3.80 3.50 3.50 3.50 S.50 3.80 26 I

O 14429l 11767l

13779, 2095Sf 0

K/

3.50 27 28 29 30 31 32 i

18498l L+

J K/

J K

i J

S.80 3.50 9.50 9.50 9.50 9.50 SS 0

26833 17915 20927 11742 24854 L

l i

3.80 94 SS SG B7 38 99 0

J L/

J L/

J H

S.50 9.80 S.50 S.80 3.50 S.50 j

l248SSi 0

28258 0

24854 25140 NOTE:

EOC 9 CORE AVERAGE BURNUP 13000 NUD/T

=

.-..--.m.

+ -,.

---

y.

Figure S-5 FORT CALHOUN CYCLE 10 ASSEMBLY AVERAGE BURNUP AT EOC10 (MWD /T)

AA

- ASSEMBLY LOCATION 01 02 BBBBB =ASSY AVG EXPOSURE, MWD /T SSS8G SSGGG OS 04

'05 08 07 j

86643 11713 18957 14432 35098 l

I 08 09 10 11 12 IS l

41593 19242 17427 39304

30192, 17884 l

l 1

14 15 18 17 18 19 38G20 16832 83626 29110 32270 39070 t

20 21 l22 23 24 25 19998 29075' 27387 28827 34618 1 17712 26 i

25662 l

l 27 28 29 30 al 32 17187 39658 32231 34572 26209 37155 SS 12141 34 35 36

'97 38 39 37793 17797 38990 17583 97054 SGSSG I

)

4.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch L reload fuel is identical to that of the Batch K fuel described in the 1985 update of the USAR and the j

Cycle 9 reload submittal. The fuel system design and analysis for ENC fuel in the Fort Calhoun reactor is described in Reference 7.

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a W

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5.0 NUCLEAR DESIGN 5.1 PHYSICAL CHARACTERISTICS 5.1.1 Fuel Management The Cycle 10 fuel management uses a low-radial leakage design, with twice, thrice and fourth burned assemblies predoninately loaded on the periphery of the core.

This low-radial leakage fuel pattern is utilized to minimize the flux to the pressure vessel welds.

In addition, selected assemblies located adjacent to crit-ical welds will contain part-length poison rods to fur-ther shield the welds.

While this type of fuel manage-ment results in reduced pressure vessel flux over a standard out-in-in pattern, the radial peaking factors are increased.

As described in Section 3.0, the Cycle 10 loading pat-tern incorporates 44 fresh Batch L assemblies (12 shimmed L/, 8 shimmed L*, and 24 unshimmed, L) with an enrichment of 3.80 w/o.

Twelve 4-cycle burned Batch G assenblies and 9 thrice burned Batch H assenblies, all i

of which were removed at E0C8, are combined with 32 once burned Batch K assenblies, 8 once burned Batch J assemblies, and 28 twice burned Batch J assemblies to produce a Cycle 10 pattern with a cycle energy of 12,500 t 500 MWD /T. The Cycle 10 core characteristics have been examined for a Cycle 9 tennination between 12,500 MWD /T and 13,500 MWD /T and limiting values estab-11shed for the safety analysis.

The loading pattern is valid for any Cycle 9 endpoint between these values.

Physics characteristics including reactivity coefft-cients for Cycle 10 are listed in Table 5-1 along with the corresponding values fran the reference cycle (Cycle 9).

It should be noted that the values of para-meters actually employed in safety analyses are differ-ent from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted values with accommodation for appropriate uncertainties and allow-ances.

Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the end of Cycle 10 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 10 values calculated for minimum scram worth exceed the required Technical Specification limit and thus provide an adequate shutdown margin.

i 5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (AR0) planar radial power distributions at 80C10, M0C10 i

_________________._______________j

5.0 NUCLEAR DESIGN (Continued) 5.1

~ PHYSICAL CHARACTERISTICS (Continued) 5.1.2 Power Distribution (Continued) and E0C10, respectively, that are characteristic of the high burnup end of the Cycle 9 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 25 and 75 percent of the fuel height. The high burnup end of the Cycle 9 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 10. The planar radial power distribu-tions for the above region with Bank 4 fully inserted at beginning and end of Cycle 10 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 9 shutdown window.

The radial power distributions described in this sec-tion are calculated data without uncertainties or other l

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 10. These conservative values, which are used in Section 7 of this document, establish the allowable lim-1ts for power peaking to be observed during operation.

Figures 3-3 and 3-4 show the integrated assembly burnup values at 0 and 12,500 MWD /T, respectively, based on an E0C9 burnup of 13,000 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 4

i 10 during nonnal base load, all rods out operation at full power is 1.99, not including uncertainty allow-ances.

5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data a

The maximum reactivity worth and planar pow-er peaking factors associated wiu an eject-2 ed CEA event are shown in Table 5-3 for both beginning and end of Cycle 10. These values encompass the worst conditions anti-cipated during Cycle 10 for any expected

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5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.3 Safety Related Data (Continued) 5.1.3.1 Ejected CEA Data (Continued) i Cycle 9 tennination point. The values shown i

for Cycle 10 are calculated in accordance with Reference 4.

In addi tion, Table 5-4 lists those values used in the Reference Analysis (Cycle 6) for canparison.

5.1.3.2 Dropped CEA Data i

The Cycle 10 safety related data' for the dropped CEA analysis were calculated iden-tically to that used in Cycle 9.

The data is reported in the dropped CEA analysis.

5.2 ANALYTICAL INPUT TO IN-CORE MEASUREMENTS i

In-core detector measurement constants to be used in evaluting the reload cycle power distributions will be calculated in the same manner as for Cycle 9.

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

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS I

The power distribution measurement uncertainties which are applied to Cycle 10 are the same as those presented in Reference i

2.

. m.

TABLE 5-1 FORT CALHOUN CYCLE 10 NOMINAL PHYSICS CHARACTERISTICS Reference Units Cycle

• Cycle 10 Critical Baron Concentration Hot Full Power, AR0, Equilibrium Xenon, BOC PPM 1108 1066 Inverse Boron Worth Hot Full Power, BOC PPM /%Ap 110 110 Hot Full Power, E0C PPM /%Ap 88 88 Reactivity Coefficients (CEAs Withdrawn)

Moderator Temperature Coefficients Beginning of Cycle, HZP 10-4ap/* F

+0.36

+0.25 End of Cycle, HFP 10-4Ap/*F

-2.5

-2.41 Doppler Coefficient Hot Zero Power, BOC 10-53p/oF

-1.78

-1.79 Hot Full Power, B0C 10-53p/or

_1,40

-1.42 Hot Full Power, E0C 10-53pfoF

-1.58

-1.59 Total Delayed Neutron Fraction, geff BOC 0.00640 0.00631 E0C 0.00545 0.00541 Neutron Generation Time, t*

B0C 10-6 sec 23.1 23.8 E0C 10-6 sec 30.7 30.7 Cycle 9

TABLE 5-2 FORT CALHOUN UNIT 1 CYCLE 10 LIMITING VALUES OF PEACTIVITY WORTHS AND ALLOWANCES FOR HOT ZERO POWER STEAM LINE BREAK, %Ap END-0F-CYCLE Reference Cycle l

(Cycle 9)

Cycle 10 1.

Worth of all CEA's Inserted 9.95 9.08 2.

Stuck CEA Allowance 1.88 3.06 l

t 3.

Worth of all CEA's Less Worth of Most Reactive CEA Stuck Out 8.07 6.02 4.

Power Dependent Insertion Limit CEA Worth 1.11 1.25 5.

Calculated Scram Worth 6.96 4.77

^

6.

Physics Uncertainty plus Bias 0.69 0.48 7.

Net Available Scram Worth 6.27 4.29 8.

Technical Specification Shutdown Margin 4.00 4.00 9.

Margin in Excess of Technical Specification Shutdown Margin 2.27 0.29 l

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9 TABLE 5-3 1

FORT CALHOUN UNIT 1 CYCLE 10 l

CEA EJECTION DATA Cycle 6 Value BOC10 Value EOC10 Value i

Maximum Radial Power Peaking Factor Full Power with Bank 4 inserted; worst CEA ejected 6.00 3.30 3.96 Zero power with Banks 4+3 inserted; worst CEA ejected 13.00*

4.03 4.89

(

Maximum Ejected i

CEA Worth (%Ap)

Full power with Bank 4 inserted; worst CEA ejected 0.30 0.21 0.29 Zero Power with Banks 4+3 inserted; worst CEA ejected 0.90*

0.32 0.44 r

• Banks 4+3+2 inserted i

Figure 5-1

{

FORT CALHOUN CYCLE 10 ASSEMBLY RELATIVE POWER DENSITY O MWD /T,

HFP, EG. XENON AA

= ASSEMBLY LOCATION 01 02 B.BBBB -RELATIVE POWER DENSITY 0.2496 0.2683 f

I 03 04 05 06 07 i

0.2606 0.9804 1.1514 1.1897 0.8052 i

1 08 09 10 11 12 IS l

0.1881 1.0420 1.3822 1.0170 1.2031 1.4010 1

14 15 16 17 l 18 19 0.4084 1.3179 1.2539 1.2907 1.1749 1.0298 i

X i;

20 21 22 23 24 25 l

1.0952 1.1857 1.3065 1.2730 1.127G 1.4166 26 0.5113 27 28 29 30 31 32 1.3020 1.0004 1.1608 1.1203 1.1887 0.9917 99 0.9171 34 35 3G 37 38 39 O.9827 1.3534 0.9965 1.9935 0.9791 0.8877 i

i X=l16XI!1Ul1 1-PIN PE6K=1.7380 i

i i

l 1

Figure 5-2 FORT CALHOUN CYCLE 10 ASSEMBLY RELATIVE POWER DENSITY l

6000 MWD /T.

HFP.

EG. XENON AA

= ASSEl18LY LOCATION 01 02 i

B.BBBB = RELATIVE POWER DENSITY 0.2714 0.2969 OS 04 05 06 07 0.2922 0.9369 1.1086 1.1418 0.8155 j

j 08 09 10 11 12 IS 0.2171 1.0601 1.3935 0.9967 1.1738 1.4050 14 15 16 17 18 19 0.4571 1.3555 1.2146 1.1673 1.1938 1.0167 X

20 21 22 23 24 25 1.1293 1.1761 1.2394 1.1881 1.0792 1.4005 1

26 i

0.5846 27 28 29 30 31 82

,1.3981 1.0334 1.1916 1.0806 1.1911 0.9714 SS 0.9863 34 35 S8 97 38 39 1.0538 1.4352 1.0184 1.3944 0.9644 0.8821 X =ll AX Il1Ull 1-PIN PEAK =1.7569 4

I

. I

/

i Figure 5-3 FORT CALHOUN CYCLE 10 ASSEMBLY RELATIVE POWER DENSITY l

13000 MWD /T,

HFP, EG. XENON AA

.- ASSEl1BLY LOCATION 01 02 B.BBBB -RELATIVE POWER DENSITY 0.9081 0.9450 OS 04 05 06 07 O.3903 0.9610 1.1055 1.1485 0.8582 08 09 10 11 12 13 0.2508 1.0755 1.4010 0.9968 1.1598 1.4251

(

14 15 16 17 18 19 i

0.4965 1.3383 1.1694 1.1279 1.1189 1.0261 6

1.1385l22 20 1 23 24 25 21 1

1.1180 1.1779 1.1375 1.0666 1.4144 I 26 0.6296 27 28 29 20 31 32 1.3825 1.0274 1.1185 1.0656 1.1329 0.9894

'SS 0.9981, S4 35 l SE 37 38 39 1.0565 1.4422 1.0198 1.4078, 0.9837 0.9142 l

X 0

l

)

f X =l1 AXIL 1Ull 1-PIN PEAK =1.6851 1

i J

l Figure 5-4 FORT CALHOUN CYCLE 10 RPD WITH BANK 4 INSERTED 0 MWD /T, HFP.

EQ. XENON l

AA hASSEMBLYLOCATION 01 02 B.BBBB [ RELATIVE POWER DENSITY

0. 2G 17 ! 0.2895 OS 04 05 06 07 0.1674 0.7928 1.1378 1.2566 0.8727 1

08 09 ///, 10 11 12 13 0.1117 0.5224 1.1345 1.0016 1.2774 1.5204

/

14

1. 5 1G 17 18 19 0.8640 1.1109 1.1589 1.2484 1.2460 1.10G1 20 21

.22 28 24 25 1.1689 1.2180 1.3454 1.3809 1.1793 1.4769 26 I

0.5864 l1.1081l29 27 28 30 Si l32 1.4811 l1.2569 l1.1811 1.1787[0.9176[

l aa 1.0757 34 35 SG 87 38

[39//,

1.1981

1. SAGO 1.09G7 1.4G84 0.9079 0.5G25 X = 11 AX I MUN 1-PIN PEAK =1.8050

'/,

CEA BANK 4 LOCATION w

Figure 5-5 FORT CALHOUN CYCLE 10 RPD WITH BANK 4 INSERTED 13000 MWD /T,

HFP, EG. XENON AA

- ASSET 1BLY LOCATION 01 02 B.BBBB [ RELATIVE POLLER DENSITY 0.3244 0.3753 9

0.7928l05 l06

$07 03 04 O.2035 1.0800l1.2222 0.9335 08

'O'9 / / /

10 11 12 13 0.1482 0.5105 1.1145 0.9694 1.2329 1.5518 14

'15 16 17 f18 19 O.4382 1.1082 1.0603 1.1359 1.1829 1.1055 I

20 21 22 l

I 24 l25 23 1.2112 1.1754 1.2153 I I.1996 1.1155 1.4720 l

! 26 I

0.73G8 l

27 28 29 30 31 32 1

1.6150 '1.1642 l1.2244 1.1301 1.1170 0.9028 i

33 l

I 1.2072,

'34 I 35

'36 0 37 38 59'///,'

l1.2589 1.8772 1.1930 1.4932 0.9024 0.5503 i

i i

i i

IWS:t X = I1 AX II1Ull 1-PIN PEAK =1.9530 CEA BANK 4 LOCATION

6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 10 at the rated power of 1500 MWt have been perfonned using the TORC computer code described in Reference 1, the CE-1 critical heat flux correlation described in l

Reference 2, and the CETOP-D computer code described in Reference 3.

This conbination was used in the Cycle 8 and 9 Fort Calhoun reload analyses (References 4 and 5) and the reload methodology can be found in Reference 6.

Table 6-1 contains a list of pertinent thermal-hydraulic parame-ters used in both safety analyses and for generating reactor pro-tective system setpoint information. The calculational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor)-listed in Table 6-1 have been conbined statistically with other uncertainty fac-tors at the 95/95 confidence / probability level (Reference 7) to define the revised design limit on CE-1 minimum DNBR. The MDNBR limit was revised from 1.22 to 1.18 to reflect the NRC approval of a 1.15 limit for 14 x 14 CE type fuel (Reference 8). The interim limit for the CE-1 correlation had been 1.19.

6.2 FUEL R0D B0 WING 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 Ifmit in the statistical combination of uncer-tainties (Reference 7).

TABLE 6-1 Fort Calhoun Unit 1 Thennal-Hydraulic Parameters at Full Power Unit Cycle 10*

Total Heat Output (Core Only)

MWt 1500 106 BTU /hr 5119 Fraction of Heat Generated in Fuel Rod

.975 Primary Systen Pressure Nominal psia 2100 Minimum In Steady State psia 2075 Maximum In Steady State psia 2150 Inlet Temperature

• F 540 1

Total Reactor Coolant Flow gpm 202,500 I

(Steady State) 106 lbm/hr 76.98 (Through the Core) 106 lbm/hr 73.55 Hydraulic Diameter ft

.044 (Nominal Channel)

Average Mass Velocity 106 lbm/hr-ft2 2.26 Core Average Heat Flux BTU /hr-ft2 179,722

( Accounts for Heat Generated in Fuel Rod)

Total Heat Transfer Surface Area ft2 28,485**

Average Core Enthalpy Rise BTU /lbm 70.1 Average Linear Heat Rate kw/ft 6.1* * ~

Engineering Heat Flux Factor 1.03***

Engineering Factor on Hot Char.nel Heat Input

. 1.0 3* *

• Rod Pitch and Bow-1.065***

Fuel Densification Factor ( Axial) 1.01***

• Design inlet temperature and nominal primary system pressure ~

twere used to calculate these parameters.

• • Based on Cycle 10 specific value of 320 shims.

• • • These factors were combined statistically (Reference 7) with other uncertainty factors at 95/95 confidence / probability

>1evel 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 10 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 groups:

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 themal 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 10 safety anal-yses are listed in Table 7-1.

Core parameters input to the safety analyses for evaluating approaches to DNB and centerline tenperature to melt fuel design limits are presented in Table 7-2.

As indicated in Table 7-1, no reanalysis was perfomed 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 9, Reference 1). For these DBEs the results and conclusions quoted in the reference cycle analysis are valid for Cycle 10.

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 summary 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 10 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 Drop 5 Not Reanalyzed 7.2.5 Transients Resulting from the Malfunction of One Steam Generator 4 Not Reanalyzed 7.3 Postulated Accidents 7.3.1 CEA Ejection Not Reanalyzed 7.3.2 Steam Line Break Reviewedb 7.3.3 Steam Generator Tube Rupture Not Reanalyzed 7.3.4 Seized Rotor 3 Reviewed 0 1 echnical Specifications preclude this event during operation.

T 2 equires High Power and Variable High Power Trip R

3 equires low Flow Trip R

4 equires trip on high differential steam generator pressure R

5 ounded by Full' Length CEA Drop B

6 vent bounded by reference cycle analysis. A negative 10 CFR 50.59 deter-E mination was made for this event.

TABLE 7-2 FORT CALHOUN UNIT 1, CYCLE 10 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle (Cycle 9)

Physics Parameters Units Values Cycle 10 Values Radial Peaking Factors For DNB Margin Analyses (F T)

R

- Unrodded Region 1.75*

1.80*

Bank 4 Inserted 1.79*

1.83*

For Planar Radial Component (F vT) of 3-D Peak x

(CTM Limit Analyses)

Unrodded Region 1.78*

1.85*

Bank 4 Inserted 1.93*

2.15*

Maximum Augmentation Factor 1.057 1.057 Moderator Temperature Coefficient 10-4ap/* F

-2.7 to +0.5

-2.7 to +0.5 Shutdown Margin (Value l

Asstrued in Limiting l

E0C Zero Power SLB)

%ap

-4.0

-4.0

• For the Loss of Coolant Flow and CEA Drop Events, the effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calcu-lations. The DNBR analysis utilized the methods discussed in Section 6.1 of this report. The procedures used in the Statistical Combination of Uncertain-ties (SCU) as they_ pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c, 2d..

(

TABLE 7-2 (Continued)

Safety Parameters Units Cycle 9 Values Cycle 10 Values Power Level MWt 1530*

1530*

Maximum Steady State Temperature

• F 547*

542*

Minimum Steady State Pressurizer Pressure psia 2053*

2053*

Reactor Coolant Flow gpm 202,500*

202,500*

Negative Axial Shape LCO Extreme Assumed at Full Power (Ex-Cores)

I

-0.18

-0.18 p

Maximum CEA Insertion

% Insertion at Full Power of Bank 4 25 25 Maximum Initial Linear Heat Rate for Transient Other than LOCA KW/ft 15.22 15.22 Steady ' State Linear Heat Rate for Fuel CTM Assumed in the Safety Analysis KW/ft 21.0 21.0 CEA Drop Time to 100%

Including Holding Coil Delay sec 3.1 3.1 Minimum DNBR (CE-1) 1.22*

1.18*

• For the Loss of Coolant Flow and CEA Drop Events, the effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calcu-lations.. The DNBR analysis utilized the methods discussed in Section 6.1 of this report. The procedures used in the Statistical Canbination of Uncertain-ties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c, 2d.

.--r

TABLE 7-3

- DESIGN BASIS EVENT REANALYZED FOR FORT CALHOUN CYCLE 10 Reason for Acceptance Summa ry Event Reanalysis Criterion of Results (changes relative to reference cycle)

Boron Dilution Increased critical boron Dilution to critical Acceptance criteria concentrations from Cycle time limits of 30 minutes met. See Section 9.

for refueling and 15 7.1.1 for details.

minutes for all other subcritical modes must be met.

Excess Load Change in TM/LP trip func-Pbias = 35.5 psia.

tion (Pvar) trip equation.

which is more limit-Reevaluate Pbias term.

ing _(as in Cycle 9) than the RCS Depres-suriza tion.

RCS Depressurization Reevaluate Pbias term.

P as = 20.7 psia b

wh' ch is less limiting than that of Excess Load event.

Sequential CEA Group Withdrawal Increased Tech. Spec.

Minimum DNBR greater MDNBR = 1.39 limits on radial peak-than 1.18 using CE-1 PLHGR < 21 kw/ft.

ing factors.

correlation. Transient PLHGR < 21 kw/ft.

TABLE 7-3 (Continued)

DESIGN BASIS EVENT REANALYZED FOR FORT CN.HOUN CYCLE 10 Reason for Acceptance Summa ry Event Reanalysis.

Criterion of Results (changes relative to reference cycle)

Loss of Coolant Flow Increased Tech. Spec.

Minimum DNBR greater Minimum DNBR = 1.50 limits on radial peak-than 1.18 using CE-1 ing factors.

correlation.

Full Length CEA Drop Increased Tech. Spec.

Minimum DNBR greater-Minimum DNBR = 1.47 limits on radial peaking.

than 1.18 using CE-1 factors.

correlation.

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-(Conti nued)-

7.1.1 Boron Dilution Event 4

The Boron Dilution event was reanalyzed for Cycle 10 to 4

]

detennine if sufficient time is available for an opera-tor to identify the cause and to terminate an approach to criticality -for all sd) critical 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 compares.the values of the key transient parameters assumed in each mode of operation for Cycle i

10 and the reference cycle (Cycle 9).

Table 7.1.1-2 compares the results of the analysis for j

Cycle 10 with those for Cycle 9._ The key results are the minimum times requiredito lose prescribed negative reactivity.in each operational mode. As.seen from Table 7.1.1-2, sufficient time exists for the operator j

to initiate appropriate action' to mitigate the conse-quences of this event.

e i

0 t

.c,++y-,-,,,,r

'v-4-<t

• ,4-----T y

w

+e 9-y v+

i TABLE 7.1.1-1 FORT CALHOUN CYCLE 10 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Parameter Cycle 9 Cycle 10 Critical Boron Concentration, PPM (All Rods Out, Zero Xenon)

Mode Hot Standby 1560 1540 Hot Shutdown 1560 1540 Cold Shutdown - Normal RCS Volume 1360 1400 Cold Shutdown - Minimum RCS Volume

• 1190 1070 Refueling 1290 1330 Inverse Boron Worth, PPM /%Ap Mode Hot Standby

-90

-90 Hot Shutdown

-55

-55 Cold Shutdown - Normal RCS Volume

-55

-55 Cold Shutdown - Minimum RCS Volume

-55

-55 Refueling

-55

-55 Minimum Shutdown Margin Assumed, %Ap Mode Hot Standby

-4.0

-4.0 Hot Shutdown

-4.0

-4.0 Cold Shutdown - Normal RCS Volume

-3.0

-3.0 Cold Shutdown - Minimum RCS Volume

-3.0

-3.0 Refueli ng

• 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 10 RESULTS OF THE BORON DILUTION EVENT Criterion For Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown Mode Margin (Min)

Margin (Min)

Cycle 9 Cycle 10 Hot Standby 92.7 93.8 15 Hot Shutdown 45.2 45.8 15 Cold Shutdown - Normal RCS Volume 39.3 38.2 15 4

Cold Shutdown - Minimum RCS Volume 16.4 18.2 15 Refueling 35.0 31.2 30 1

T

-'~r

7.0 TRANSIENT ANALYSIS (Continued) 7.1 (Continued) 7.1.4 Excess load Event The Excess Load event was reanalyzed for Cycle 10 to detennine the pressure bias tenn for the TM/LP trip set point.

The Excess Load event is one of the DBEs analyzed to determine the maximum pressure bias tenn input to the TM/LP trip. The methodology used for Cycle 10 is des-cribed in References 3 and 4.

The pressure bias tenn accounts for margin degradation attributable to mea-surenent and trip system processing delay times.

Changes in core power, inlet tenperature and RCS pres-sure during the transient are monitored by the TM/LP trip airectly.

Consequently, with TM/LP trip setpoints and the bias tena detennined in this analysis, adequate protection will be provided for the Excess Load event to prevent the acceptable DNBR design limit from being exceeded.

The assunptions used in the analysis to maximize the pressure bias tenn 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 stean 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 assumption of the most negative moderator and fuel temperature coefficients dur-ing the cycle.

j l

(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-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 35.5 psia is required. This is greater than that input from the RCS Depressuriza-tion event, the other event for which a pressure bias term is calculated. Hence, the use of the pressure bias factor detennined by this event in conjunction with the TM/LP trip,- will ~ prevent ex-ceeding the DNBR design limit for A00's which re-quire TM/LP trip protection.

7.0 TRANSIENT ANALYSIS (Continued) 7.1 (Continued) 7.1.7 RCS Depressurization Event The RCS Depressurization event was reanalyzed for Cycle 10 to detennine the pressure bias tenn for the TM/LP setpoint.

The RCS Depressurization event is one of the DBEs anal-yzed to determine the maximum pressure bias term input to the TM/LP trip.

The methodology used for Cycle 10 is the same as that used for Cycle 9 and is described in References 3 and 4.

The pressure bias tena accounts for margin degradation attributable to measurement and trip systen processing delay 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 detennined in this analysis, adequate protection will be provided for the RCS Depressurization event to prevent the acceptable DNBR design limit fran being ex-ceeded.

l The assunptions 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 4

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 assunec' 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 scram worth versus insertion used in the anal-ysis is a bottom peaked shape. This power distri-bution maximizes the time required to tenninate the decrease in DNBR following a trip.

The analysis of this ' event shows that a pressure bias tenn of 20.7 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 tenn 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.

7.0 TRANSIENT ANALYSIS (Continued) 7.2 (Conti nued) 7.2.1 CEA' Withdrawal Event The CEA Withdrawal event was reanalyzed for Cycle 10 to detennine 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, High Pres-surizer Pressure, or Axial Power Distribution Trips).

The methodology contained in Reference 3 was employed in analyzing the CEA Withdrawal event. This event is classified as one for which the acceptable DNBR and centerline to melt limits are not violated by virtue of maintenance of sufficient initial steady state thermal 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 Withdrawal, the Variable High Power Trip and High Pressurizer Pressure Trip in conjunction with the initial steady state LCOs, prevents DNBR limits fran 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 fran the excore detectors) and High Pressurizer Pressure Trip in both the detennination of the required initial overpower margin for DNBR using CETOP/CE-1 and the peak linear heat generation rate for the CTM SAFDL.

For the HFP CEAW DNBR analysis, an MTC identical to that utilized in Reference 5 and the gap thennal con-ductivity consistent with the assumption of Reference 3 were used in cordunction with a variable reactivity in-sertion rate. The range of. reactivity insertion rates examined is given in Table 7.2.1-1.

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 dcmonstrate 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 thermal power, was assuned in the analysis.

The zero power case initiated at the limiting condi-tions of operation results in a minimum CE-1 DNBR of 7.48.

Also, the analysis shows that the fuel-center-line tenperatures 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 (Conti nued) 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 case.

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-trenes of the LCOs, is 1.39.

The HFP maximum reactivity insertion rate analysis shows that the fuel centerline tenperatures are well below those corresponding to the acceptable CTM limit.

The sequence of events for the full power case with the 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, t

RCS coolant tenperatures, 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 required) will not lead to a DNBR or fuel temperature which exceed the DNBR and centerline to melt design limits.

I i

TABLE 7.2.1-1 FORT CALHOUN CYCLE 10 KEY PARAMETERS ASSUMED IN THE CEA WITHDRAWAL ANALYSIS Parameter Units HZP HFP Initial Core Power Level MWt i

102% of 1500*

Core Inlet Coolant Temperature

• F 532*

542*

Pressurizer Pressure psia 2053*

2053*

Moderator Temperature Coefficient x10-4ap/* F

+0.5

+0.5**

Doppler Coefficient Multiplier 0.85 0.85 CEA Worth at Trip 10-2ap

-4.65

-5.70 Reactivity Insertion Rate Range x10-4ap/sec 0 to 1.0 0 to 1.0 CEA Group Withdrawal Rate in/ min 46 46 Holding Coil Delay Time sec 0.5 0.5

• The effects of uncertainties on these parameters were accounted for deter-ministically and the DNBR calculations used the methods discussed in Sec-tion 6.1 of this document and detailed in References 2a, 2b, 2c and 2d.

• • DNBR analysis assunes MTC consistent with Reference 5.

TABLE 7.2.1-2 FORT CALHOUN CYCLE 10 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM ZERO POWER Time (sec)_

Event Setpoint or Value 0.0

.CEA Withdrawal Causes Uncontrolled Reactivity Insertion 34.4-Variable High Power Trip Signal 29.1% of 1500 MWt Generated 34.8 Reactor Trip Breakers Open 35.3 CEAs Begin to Drop Into Core t

35.8 Maximum Core Power 39.3% of 1500 MWt 36.7 Maximum Heat Flux 26.7% of 1500 MWt 36.7 Minimum CE-1 DNBR 7.48 2

39.9 Maximum RCS Pressurt, psia 2224 a--,

e~

TABLE 7.2.1-3 4

FORT CALHOUN CYCLE 10 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM FULL POWER (MAXIMUM REACTIVITY INSERTION RATE)

Time (sec)

Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 4.5 High Power Trip Signal Generated 112% of 1500 MWt 4.9 Reactor Trip Breakers Open 5.4 CEAs Begin to Drop Into Core 5.5 Maximum Core Power 113.92% of 1500 MWt 5.8 Maximum Heat Flux 108.51% of 1500 MWt 6.7 Maximum RCS Pressure, psia 2098 1

f.

+-,

n.

+

I l

5 100 i

i i

i i

i l

90 u

j

-80 l

~o o

V 70 u.

o 60 x2 t.

_j j

u-50 1

s<w I

40 l.

LLJ 4

e 4

xy 30 i

w.

x o

20 o

4 10 0

0 10 20 30

-40 50 60

~ TIME, SECONDS 4

CEA Withdrawal (Zero Power)

OmahaPublicPowerDistrict Figure CoreAverageHeatFluxvs. Time FortCalhounStation-ljnitNo.i 7.2.1-2

100 i

i i

90 80 70 z

8.

S 60.

j ci l

w 50 b

E 40 W

8 30 20 l

10 i

I I

l t

0 i

i 0

10 20 30 40 50 60 TIME, SECONDS-1 3

CEA Withdrawal (Zero Power)

OmahaPublicPowerDistrict Figure-CorePowervs. Time FortCalhounStation-UnitNo.i 7.2.1-1

1 570 i

i i

I I

560 u.

i lEa m'

550 T

~

H s

22 5

I e

540 AVG W

m 530 -

T 520 I

I I

I I

0 10 20 30 40 50 60 TIME, SECONDS 1

CEA Withdrawal-(Zero Power)

OmahaPublicPowerDistrict Figure RCSTemperaturesvs. Time FortCalhounStation-UnitNo.i 7.2.1-3

2350 i

i i

i 2300 2250 2200 --

N E 2150 i

Y a 2100 E

2050 m

EE 2000 1950 1900

=

-1850 0

10 20 30 40 50 60 TIME, SECONDS CEA Withdrawal (Zero Power)

OmahaPublicPowerDistrict Figure RCSPressurevs. Time FortCalhounStation-ljnitNo.i 7.2.1-4

120 i

i 110 100 90 E

80 a

8 70 Ei a

60 h

50 E

40 w

Es 30 20 10 0

0 10 20 30 40 TIME, SECONDS CEA Withdrawal (Full Power)

OmahaPublicPowerDistrict Figure CorePowervs. Time FortCalhounStation-UnitNo.i 7.2.1-5

120 i

i i

110^

/

a 100 g

8 90 E

85 80 k

a 70 4

5d 60 tiy 50 40 Fu E

30 E

8 20 10 0

0 10 20 30 40 TIME, SECONDS CEA Withdrawal (Full Power)

OmahaPublicPowerDistrict Figure CoreAverageHeatfluxvs. Time FortCalhounStation-UnitNo.i 7.2.1-6

r

'620 610

~

T H

u.

590

.a i

0 580

.g.

Q.

[T 3yg s'

570 i

E-W 560 m

550 540 530 I

0 10 20 30 40 TIME, SECONDS i

l t

a CEA Withdrawal (full Power)

OmahaPublicPowerDistrict Figure-RCSTemperaturesvs. Time FortCalhoun~ Station-UnitNo.i 7.2.1-7 1

-2

~ - -

2250 iz 2200 2150 2100 N

$ 2050 l

@ 2000 E

E 1950 g

oC w

1900 1850 l

L 1800 l

1750 I

I I

0 10 20 30 40 TIME, SECONDS CEA Withdrawal (Full Power)

OmahaPublicPowerDistrict Figure RCSPressurevs.-Time FortCalhounStation-UnitNo.i 7.2.1-8

7.0 TRANSIENT ANALYSIS (Continued) 7.2 (Conti nued) 7.2.2 Loss of Coolant Flow Event The Loss of Coolant flow event was reanalyzed for Cycle 10 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 paranetrically in initial axial shape and rod configuration using the methods described in' Reference 3 (which utilizes the statistical conbina-tion of uncertainties in the DNBR analysis as described in Appendix C of Reference 2c and 2d).

I-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 initial steady state thennal 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 6 and 7) which utilizes implicit modeling of the l

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 10 analysis and compares them to the reference cycle (Cycle 9) values.

Table 7.2.2-2 presents the NSSS and RPS responses dur-ing a four pump loss of flow initiated at an axial i

shape index of -0.182 which bounds the DNBR related ax-1 ial shape index LCO.

The low flow trip setpoint is reached at 2.00 seconds and the scram rods start drop-ping into the core 1.15 seconds later. A minimum CE-1 DNBR of 1.50 is reached at 3.88 seconds. Figures 7.2.2-2 to 7.2.2-S 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.

TABLE 7.2.2-1 FORT CALHOUN CYCLE 10 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Cycle 9 Cycle 10 l

Initial Core Power Level MWt 1530*

1530*

l Initial Core Inlet Coolant Temperature

• F 547*

542*

l Initial RCS Flow Rate gpm 202,500*

202,500*

Pressurizer Pressure psia 2053*

2053*

Moderator Temperature Coefficient 10-4ap/ F

+0.5

+0.5 Doppler Coefficient Multiplier 0.85 0.85 LFT Analysis Setpoint

% of initial flow 93 93 LFT Response Time sec 0.65 0.65 4-Pump RCS Flow Coastdown Fig. 7.2.2-1 Fig. 7.2.2.

CEA Holding Coil Delay sec 0.5 0.5 CEA Time to 100% Insertion sec 3.1 3.1 (Including Holding Coil Delay)

CEA Worth at Trip (all rods cut)

%p

-6.87

-5.58 A

Total Unrodded Radial Peaking 1.75 1.80 Factor (F T)

R

• The uncertainties on these parameters were cabined statistically rather than deter-ministically. The values listed represent the bounds included in the statistical combi nation.

I 1

i TABLE 7.2.2-2 FORT Call 10VN CYCLE 10 SEQUENCE OF EVENTS FOR LOSS CF FLOW Time (Sec)

Event Setpoint or Value

~0.0 Loss of Power to all Four Reactor Coolant Pumps 2.0 Low Flow Trip Signal Generated 93% of 4-Pump Flow 2.7 Trip Breakers Open 3.2 Shutdown, CEAs Being to Drop into Core 3.9 Minimum CE-1 DNBR 1.50 5.6 Maximum RCS Pressure, psia 2116

1 I

I I

t

.9

.8

.7 Ep

.8 W

E

.5 2

S u.

s

.4 8

.3

.2 -

.1 0

I I

i 0

4 8

12 16 20 TIME, SECONDS

.LossofCoolantFlow OmahaPublicPowerDistrict figure CoreFlowFractionvs. Time FortCalhounStation-UnitNo.i 7.2.2-1

110 i

100 90 80 1

70 0

8 g

60 w

50 E

40 w

E a

30 20 10 I

0 0

4 8

12 16 20 TIME, SECONDS.

LossofCoolantFlow OmahaPublicPowerDistrict Figure CorePowervs. Time FortCalhounStation-UnitNo.i 7.2.2-2 4

4 l

110 i

i i

i 100 go y-x g

80 i

E o

70 a

N 60 if D

50 w

l 52 40 1

?

30

_g Eo

)

20 i

10 0

0

'4 8

12 16 20 TIME, SECONDS LossofCoolantFlow OmahaPublicPowerDistrict Figure CoreAverageHeatFluxvs. Time FortCalhounStation-UnitNo.i 7.2.2-3 A

610 i

i i

I 600 590 TH

~

u_

g 580 O

O TAVG 560 5

g 550 cr Tc 540 530 520 0

4

.8 12 16 20 q

TIME, SECONDS LossofCoolantFlow OmahaPublicPowerDistrict Figure

.RCSTemperaturevs. Time FortCalhounStation-UnitNo.i 7.2.2-4

2200 i

i i

i 2150 2100 y2050 i

a 2000

!3 E

g 1950 cc 1900 1850 1800 0

4 8

12 16 20 TIME, SECONDS LossofCoolantFlow OmahaPublicPowerDistrict Figure ACSPressurevs. Time FortCalhounStation-UnitNo.i 7.2.2-5

7.0 TRANSIEifT ANALYSIS (Continued) 7.2 (Continued) 7.2.3 Full Length CEA Drop Event The Full Length CEA Drop event was reanalyzed for Cycle 10 to detennine the initial thennal margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNBR and fuel centerline to melt de-sign 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 paraneters used for Cycle 10 and compares them to the reference cycle (Cycle

9) values.

Conservative assunptions used in the analy-sis are consistent with those discussed in Reference 3 and include:

1)

The most negative moderator and fuel tenperature 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.

2)

Charging pumps and proportional heater systems are assumed to be inoperable during the transient.

This maximizes the pressure drop during the event.

3)

All other systens are assuned to be in the manual mode of operation and have no impact on this eve nt.

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 conditions described in Table 7.2.3-1.

The transient behavior of key NSSS para-meters are presented in Figures 7.2.3-1 to 7.2.3-4.

The transient was conservatively analyzed at ful1 power with an ASI of -0.182, which is outside of the LC0 limit of -0.06.

This results in a minimum CE-1 DIBR of 1.47.

A maximum allowable initial linear heat generation rate of 16.8 KW/ft could exist as an initial condition with-out exceeding the acceptable fuel centerline to melt limit of 21 KW/ft during this transient. This amount of largin is assured by setting the Linear Heat Rate re-

'ated LCOs based on the more limiting allowable linear i eat rate for LOCA.

tIT can be concluded that the CEA Drop event when initi-a$d from the Tech. Spec. LCOs will not exceed the DNBR ar] centerline to melt design limits.

)

}

s TABLE 7.2.3-1 FORT CALHOUN CYCLE 10 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Parameter Units Cycle 9 Cycle 10 Initial Core Power Level MWt 102% of 1500*

102% of 1500*

Core Inlet Temperature

'F 547*

542*

Pressurizer Pressure psia 2053*

2053*

Core Mass Flow Rate gpm 202,500*

202,500*

Moderator Temperature Coefficient x10-4ap/*F

-2.7

-2.7 Doppler Coefficient Multiplier 1.15 1.15 i

CEA Insertion at Maximum Allowed

% Insertion of 25 25 i

Power Bank 4 Dropped CEA Worth

%Ap unrodded

-0.2261

-0.2308 PDIL

-0.2238

-0.2238 Maximum Allowed Power Axial Shape Index at Negative Extreme of LC0 Band

-0.18

-0.18 Radial Peaking Distortion Factor Integrated Radial Peaking Unrodded Region 1.1585 1.1741 Bank 4 1.1557 1.1727 Inserted Region Planar Radial Peaking Unrodded Region 1.213 1.251 Bank 4 1.205 1.221 Inserted Region i

• The uncertainties on these paraneters were conbined statistically rather than deterministically. The values listed represent the bounds included in the statistical conbination.

m

TABLE 7.2.3-2 FORT CALHOUN ~ CYCLE 10 SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP Time (Sec)

Event Setpoint or Value 0.0 CEA Begins to Drop into Core 1.0-CEA Reaches Full Inserted Position 100% Inserted 1.1 Core Power Level Reaches Minimum and 70.7% of 1500 Begins to Return to Power due to Reactivity Feedbacks 96.5 Core Inlet Temperature Reaches a 534.9 F Minimum Value 19 8.9 Reactor Coolant System Pressure Reaches 2015 a Minimum Value 200.0 Core Power Returns to its Maximum Value 96.2% of 1500 MWt 200.0 Minimum DNBR is Reached 1.47

[

0 4

0 l

P 110 i

100 -

90 80 70 u

~

a 60 -

ti Y

a 50 E

8 40 30 -

20 i

10 0

0 20 40 60 80 100 120 140 160 180 200 TIME, SECONDS FullLengthCEADrop OmahaPublicPowerDistrict Figure CorePowervs. Time FortCalhounStation-UnitNo.i 7.2.3-1

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80 E

70 n

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40 5

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0 20 40 60 80 100 120 140 160 180 200 TIME, SECONDS 1

FullLengthCEADrop OmahaPublicPowerDistrict Figure CoreAverageHeatFluxvs. Time FortCalhounStation-UnitNo.!

7.2.3-2

600 i

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TH 590 580 S

Q

.570 I

gyg g

560 i

W 550 y

T g 540 530 520 0

20 40 60 80 100 120 140.160 180 200 TIME, SECONDS

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Figure RCSTemperaturesvs. Time FortCalhounStation-UnitNo.i 7.2.3-3

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FulllengthCEADrop OmahaPublicPowerDistrict Figure RCSPressurevs. Time FortCalhounStation-UnitNo.i 7.2.3-4

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d T

FIGURE 11 1 S

THERM AL MARGINi8.OW PRESSURE TRIP FUNCTIONAL DIAGRAM FOR FORT CAlll0Ui!T Uf!IT 1 TC Y

5

^

BM f

=T+KB C

C

\\ Axial a

KC Al a

M Axial

\\FMET o

Offset c:

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F STCAL PVAR PF 1

y (Continued pp E

PF,,

Below)

APF

\\

p I

., APF

-. /

/

X BW h

~

X 0

Bpp 100%

PF x 100 m

' 100-Opp - 2 gpp l0000 l

+ TRIP PVAR MAX a (APF)

ALARM:

TRIP P

CALCULATION:

PRETRIP UNIT P<PPRETRIP n(PF)(AI)B + 6TCAL + Y p

P

=

TRIP:

VAR C + K B, Q a MAX (4,B)

ME TRIP MERE TCAL=T C

PR SURE PTRIP = MAX (PVAR, PMINI

7.0~

TRANSIENT ANALYSIS (Continued) 7.3 (Conti nued) 7.3.1 CEA Ejection Input paraneters.to the CEA Ejection accident were examined and found to be bounded by the previous analysis of Cycle 6.

Therefore, under the guidelines of 10CFR 50.59 no reanalysis for Cycle 10 was performed.

4 h

7.3.2 Steam Line Break Accident This accident was evaluated for Cycle 10 using the methodol-L ogy discussed in References 8 and 3.

The Stean Line Break accident was previously analyzed in the Fort Calhoun FSAR and satisfactory results were reported therein. The Stean Line i

Break accidents at both HZP and HFP were examined in the i

reference cycle (Cycle 8) safety evaluation with acceptable i

i results obtained. Both the FSAR and reference cycle evalu-ations' are reported in the 1985 update of the Fort Calhoun USAR.

1 l

The Cycle 10 Full Power Steam Line Break accident way evalu-tive effective MTC of -2.7 x 10-%p/*F '

j ated for a more neggAp/*F value that was used in the Cycle 8 than the -2.5 x 10-i a nalysis. The Cycle 10 neg'ative MTC limit of -2.7 x 10-4 ap/*F, however, remains unchanged. from Cycle 9, and the cooldown curve for Cycle 10 is bounded by that of Cycle 9-(as 4

shown in' Figure 7.2.3-1).

The cooldown curves for Cycles 1, 8 and 10 are shown in Figure 7.2.3-2.

This figure shows that i

the-reactivity insertion for the Cycle 10 core with an MTC of

-2.7 x 10-4ap/*F due to a 'Stean Line Break accident at full i

power is substantially less than the value used in the Cycle 8 analysis.

(This smaller reactivity insertion is-due to the use of the DIT cross-sections which are' valid for a range of moderator tenperatures fran roon tenperature to 800 K while the' analyses prior to Cycle 9 were perfonned with cooldown curves derived by conservatively extrapolating CEPAK cross-3 section values to low temperatures.) The fuel temperature 4

coefficient used in the Cycle 8 analysis is conservative with

. respect to the fuel temperature coefficient calculated for the Cycle 10 core including uncertainties. The Cycle 10 min-

, imum available shutdown worth is 6.45%ap compared to a Cycle 8'value of 6.68%ap.

The reduction of 0.23%Ap in scran worth from Cycle 8 to Cycle 10 is offset by the 3.83%Ap gain in j

moderator cooldown reactivity.

The-net gain assures that the overall reactivity insertion for a Cycle 10 Steam Line Break is less than that of the reference cycle analysis.

There-fore, the return to power is less than that of the reference cycle and Cycle 1 FSAR analyses.

A similar evaluation was performed for the Zero Power Steam i

Line Break accident. - Again the Cycle 10 cooldown for an MTC of -2.7 x 10-4ap/*F shows a substantially smaller reactivity 6

insertion than was used in the Cycle 8 analysis'(as seen in

7.0 TRANSIENT ANALYSIS (Continued) 7.3 (Conti nued) 7.3.2 Steam Line Break Accident (Continued)

Figure 3.2.2-2.

Since the minimum available shutdown margin for Cycle 10 remains unchanged from the reference cycle value (4%Ap), the overall reactivity insertion for the Cycle 10 Steau Line Break accident will be substantially less than that of the reference cycle. Therefore, the consequences of a zero power Stean Line Break accident for Cycle 10 will be less severe than that reported for the reference cycle and

+

the FSAR (Cycle 1) cases.

Based on the evaluation presented above, it is concluded 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.

Since a negative 10 CFR 50.59 determination was made for the Cycle 10 Stean Line Break Accident, no reanalysis was per-

' fo rmed.

C

--~,

,e-n-

5.0 b

i 4.0 CYCLE 9 i

I

.0 CYCLE 10

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SteamLineBreakIncident OmahaPublicPowerDistrict Figure j

~ ReactivityvsModeratorTemperature FortCalhounStation-UnitNo.-i 7.3.2-1 i

2 2-

l 8.0 i

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6.0 4

CYCLEi x

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CYCLE 8 FULLPOWER

{

CYCLE i 2'0 C

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\\ CYCLE 10

-2.0 200 300 400 500 600 700 CORE AVERAGE MODERATOR TEMPERATURE,'F i

l SteamLineBreakIncident OmahaPublicPowerDistrict figure ReactivityvsModeratorTemperature FortCalhounStation-UnitNo.i 7.3.2-2 s._

,.,,e m.

,,.,.-y,..

7.0 TRANSIENTANALYSIS(Continued) 7.3 (Conti nued) 7.3.4 Seized Rotor Event The Seized Rotor event was evaluated for Cycle 10 to demon-strate that only a small fraction of fuel pins are predicted to fail during this event.

Cycle 10 is bounded by the Cycle 9 analysis because an F T of 1.85 was assuned in the Cycle 9 R

analysis and the Cycle 10 Technical Specification of 1.80 renains conservative with respect to the F T value used in R

the Cycle 9 analysis.

Therefore, the total number of pins predicted to fail will continue to be 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 limits of 10CFR100.

Since a negative 10 CFR 50.59 detennination was made for the Cycle 10 Seized Rotor Event, no reanalysis was performed.

i

)

8.0 ECCS PERFORMANCE ANALYSIS t

The Loss of Coolant accident evaluation was perfonned using the methodology discussed in Reference 1.

The District has verified that the physics input assunptions and the maximum rod burnup are within the bounds asstaned in the large break analysis for the reference cycl'e (Cycle 8) as reported in the l

1985 update of the USAR. Therefore, under the guidelines of 10CFR50.59 no f

reanalysis for Cycle 10 was performed.

I L

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I

9.0 STARTUP TESTING The startup testirg program proposed for Cycle 10 is identical to the pro-gram outlined in the Cycle 6 Reload Application, with two exceptions.

First, a CEA exchange technique for zero power rod worth measurements will be performed, in addition to the normal boration/ dilution technique. Also, low power CECOR flux naps and psuedo-ejection rod measurements will be sub-stituted for the full core symmetry checks.

The CEA exchange technique is a method for measuring rod worths which is both faster and produces less waste than the typical boration-dilution me thod. The sd)stitution of a zero power pseudo-ejection rod worth mea-surement and low power CECOR maps for full core symmetry checks will provide a better assessnent of any azimuthal power tilts because of the asymmetric design of the Cycle 10 core. The combination of the pseudo-ejection tech-nique at zero power and low power CECOR maps provides a less time consuming but equally valid technique for detecting azimuthal power tilts during re-load core physics testing. The psuedo-ejection rod measurement involves the dilution of a bank into the core, borating a CEA out, and then exchanging (rod swap) the CEA against other symmetric CEA's within the bank to measure rod worths. The acceptance and review criteria for these tests are:

Test Acceptance Criteria Review Criteria Pseudo-ejection None The greater of: 2.5c devi-rod worth mea-ation from group average surement or 15% deviation from group average.

Low Power CECOR Technical Specif f-Azimuthal tilt less than maps cation limits on 2 01.

F T, Fxy, and Tq R

T OPPD has reviewed these tests and has concluded that no unreviewed safety question exists for implementation of these procedures.

l

10.0 MINI-CECOR/BASSS LCO MONITORING SYSTEM The Better Axial Shape Selection System (BASSS) monitors the Limiting Conditions for Operation on peak linear heat rate and departure from nucleate boiling using as input the data available from the MINI-CECOR code and the ERF plant computer. This arrangement is similar to the one used by Baltimore Gas and Electric at their Calvert Cliffs Units, and described in the Combustion Engineering Setpoint Methodology Top-ical, CENPD-199-P, Revision 1-P, dated April 1982.

MINI-CECOR is a mini-computer version of Combustion Engineering's CECOR code.

It uses the same algorithms as the mainframe version but has a reduced level of editing options to enable it to fit on a mini-compu-l ter. Combustion Engineering developed MINI-CECOR and will install and benchmark the code on the District's ERF computer system prior to the beginning of Cycle 10 operation.

It is used in this application to syn-l thesize the following parameters from readings of the fixed incore de-tectors:

1.

The three-dimensional power peaking factor (F )

q 2.

The core average axial shape index (T) 3.

The total planar radial peaking factor (Fxy )

T l

4.

The total integrated radial peaking factor (F T)

R l

These inputs to BASSS are descriptive of the existing core power distri-bution.

The inputs to BASSS obtained from the plant computer are the following:

l 1.

Measured core power level 2.

Percent insertion of the lead CEA regulating group.

BASSS consists of two algorithms: one for peak linear heat rate moni-toring and another for DNB monitoring. The peak linear heat rate al-gorithn uses the 3-D power peaking factor and the measured core power l

level to calculate the core peak linear heat rate. The algorithn ap-plies appropriate uncertainties and allowances (per the Technical Spec-ifications) to the 3-D peaking factor. The measured peak linear heat l

rate is compared to the monitoring limit, which is based on both LOCA and A00 transient analysis considerations, and an alann is activated when the monitoring limit is exceeded. The power operating limit on linear heat rate is also calculated and displayed as an indication of the available operating margin. The DHB algorithm is an improvement over the excore ASI monitoring systen in that it uses the incore axial shape index, CEA group position and the radial peaking factors to es-I l

tablish the plant's power operating limit. An alann is activated when the power operating limit is exceeded. A gain in operating margin re-suits from the following:

1.

A reduction in ASI uncertainty due to the use of incore ASI ver-j sus excore ASI.

l 2.

Knowledge of the actual CEA group position versus the excore sys-tem's assumption that the CEAs are inserted to the PDIL's trans-l tent insertion limit.

l 3.

Knowledge of the actual radial peaking factors versus the excore system's assumptions that radial peaks are at the Technical Spec-ification limits.

BASSS is also provided with the capability to monitor the Limiting Con-ditions for Operation on FxyT and FnT.

If the Technical Specification T or F T are exceeded during nomal plant operation, BASSS limits of Fxy R

will activate an alarm and calculate the proper tradeoff with maximum allowed power that ensures that the Axial Power Distribution and Ther-l mal Margin / Low Pressure Trips remain conservative. An alarm is acti-vated if the measured power is higher than the allowed power level.

The uncertainties for the use of Mini-CECOR/BASSS to monitor DNB A LHR has been developed by Combustion Engineering and documented in Refer-l ence 1.

l l

m l

l l

l

11.0 TM/LP MODIFICATION The TM/LP calculators at the Fort Calhoun Station originally monitored core power, reactor coolant inlet temperature and core coolant pressure. The axial shape index was not a monitored parameter, but was assumed (in the setpoint analysis) to always be at the Axial Power Distribtulon LSSS value.

For Cycle 10, the TM/LP calculators were modified to include modules for the ASI monitoring. This change makes the TM/LP calculators functionally like the CE " standard systen" (Ref.1).

Figure 11-1 is a simplified functional diagram of this system. The signal l

representing the core power (B) is the auctioneered highest of the neutron I

l heat flux and the delta-T power. This signal is used to calculate the PF fun ction. The PF tenn multiplied by the core power is the equivalent to the QR1 tenn in the " standard" TM/LP equation. The measured axial shape index signal (Y), which includes the adjustment for shape annealing and represents the peripheral axial shape index, is used to calculate A1. The A1, PF, B, and the constant a are multiplied together to generate a signal representing the first term in the PVAR equation.

The second and third tenns of the Pvar equation renain the same.

The second tenn is the product of the constant 6 and Teal. The third tenn is the con-stant Y. These terms are added to the first tenn to calculate the variable low pressure trip limit.

The calculated limit is then capared to a fixed low pressure trip limit (Pmin). The auctioneered highest of these signals P rio is capared to the measured reactor becmes the trip limit (P rip).

t t

coolant pressure (P) and a trip signal is generated when P is less than or equal to Ptrip.

The change to the TM/LP calculators was the addition of the Al function.

The balance of the system is unchanged.

The uncertainty associated with ASI monitoring has been cabined statistic-ally into the overall uncertainty of the TM/LP trip system. This work was perfonned by Combustion Engineering and documented in Reference 2.

l l

t i

i

U I

12.0 REFERENCES

References (Chapters 1-5) 1.

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

2.

OPPD-NA-8301-P, Revision 01, " Reload Core Analysis Overview", June 1985.

3.

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

September 1983.

4.

OPPD-NA-8303-P, " Transient and Accident Methods and Verification",

September 1983.

5.

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

l 6.

Letter from R. L. Andrews to E. J. Butcher, LIC-85-237, dated June 13, 1985.

7.

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

I l

d J

I r

m.

m

- mm

12.0 REFERENCES

(Continued)

References (Chapter 6) 1.

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

2.

CENPD-152-PA (Proprietary) " Critical Heat Flux Correlation For CE-Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," April 1975.

l l

3.

CEN-191(B)-P "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 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 Pubite Power. District, Fort Calhoun Station, Unit No.1 Docket No. 50-285, March 15,1983.

5.

Safety Evaluation by, the Office of Nuclear Reactor Regulation Supporting Amendment No. 77 to Facility Operating License No. DPR-40 for the Omaha Public Power District, Fort Calhoun Station, Unit No.1 Docket No. 50-285, April 26,1984.

6.

OPPD-NA-8301-P, Revision' 01, " Omaha Pub 1tc Power District Nuclear Analysis Reload Core, Analysis Methodology Overvieu," June 1985.

7.

CEN-257(0)-P, " Statistical Conbination of Uncertainties,, Part 2,"

Supplement 1-P, August 1985 8.

Safety Evaluation Report on CENPD-207-P-A, "CE Critical Heat Flux:

Part 2 Hon-Uniform Axial Power Distribution," letter, Cecil Thomas (NRC) to Mr. A. E. Scherer (Combustion Engineering), November 2,1984.

t I

k

\\

s F f 5

c 1.

a i ll t

12.0 REFERENCES

(Continued)

References (Chapter 7)

-1.

" Amendment to Operating License DPR-40, Cycle 9 License Applica-tion", Docket No. 50-285 February 8.1984.

2a.

" Statistical Combination of Uncertainties Methodology, 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 Systen Paraneter~ Uncertainties in Thennal Margin Analysis for Fort Calhoun Unit 1", CEN-257(0)-P, November 1983.

l 2c.

" Statistical Combination of Uncertainties Methodology, Part 3:

l Departure from Nucleate Boiling and Linear Heat Rate Limiting l

Conditions for Operation for Fort Calhoun", CEN-257(0)-P, 1

November,1983.

2d.

" Statistical Combination of Uncertainties Methodology for Fort Calhoun, Supplement-1-P," CEN-257(0)-P, August 1985.

3.

" Omaha Public Power District Reload Core Analysis Methodology -

Transient and Accident Methods and Verification", OPPD-NA-8303-P, Septenber 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.

j i

7.

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

8.

" Omaha Public Power District Reload Core Analysis Methodology.

Neutronics Design Methods and Verification", OPPD-NA-8302-P, Septenber 1983.

12.0 REFERENCES

(Continued)

References (Chapter 8) 1.

"0maha Public Power District Reload Core Analysis Methodology -

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

References (Chapter 10) 1.

CEN-257(0)-P, " Statistical Combination of Uncertainties, Part 3,"

Supplement 1-P, August 1985.

Renerences (Chapter 11) 1.

CEl4PD-199-P, Revision 1-P, " Combustion Engineering Setpoint Methodology," March,1982.

2.

CEN-257(0)-P, " Statistical Combination of Uncertainties, Part 1,"

Supplement 1-P, August 1985.

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