ML20053D278

Cycle 4 Reload Rept.

ML20053D278
Person / Time
Site:	Crystal River
Issue date:	05/31/1982
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20053D266	List: ML20053D265 ML20053D275 ML20053D278
References
BAW-1684, BAW-1684-R02, BAW-1684-R2, TAC-46962, TAC-48471, NUDOCS 8206040258
Download: ML20053D278 (15)

Text

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BAW-1684, Rev. 2 May 1982 8

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I CRYSTAL RIVER UNIT 3

- Cycle 4 Reload Report -

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BABCOCK & WILCOX Nuclear Power Group I Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 l

8206040259 820528 Babcock & Wilcox I PDR ADOCK 05000302 P pon

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R vipion R (5/25/82)

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7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR 1 accident analysis has been examined with respect to changes in cy-cle 4 parameters to determine the effect of the cycle 4 reload and to ensure that thermal perfomance during hypothetical transients is not degraded.

I The ef fects of fuel densification on the FSAR accident results have been eval-usted and are reported in reference 6. Since batch 6 reload FAs do not contain fuel rods whose theoretical density is lower than those considered in the ref-erence 6 report, the conclusions in that reference are still valid with the exception of the four-pump coastdown and locked-rotor accident which were re-evaluated at 102% of 2568 MWt for the previous cycle of operation and remain valid for cycle 4 operation at 2475 MWt without the pump status trip. The three pump coastdown from three pump operation has been evaluated with the same flux / flow setpoint at a power level of 1399 MWt with acceptable DNB re-I sults.

7.2. Accident Evaluation The key parameters that have the greatest effect on determining the outcome I of a transient can typically be classified in three major areas: core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.

Core thermal properties used in the FSAR accident analysis were design operat-ing values based on calculational values plus uncertainties. The cycle 4 I thermal-hydraulic maximum design conditions are compared to the previous cy-cle 3 values in Table 6-1. These parameters are common to all the accidents considered in this report. A comparison of the key kinetics parameters from the FSAR and cycle 4 is provided in Table 7-1.

Generic LOCA analyses for B&W 177-FA lowered-loop NSSs have been performed using the Final Acceptance Criteria ECCS Evaluation Model. The large-break analysis is presented in a topical report', and is further substantiated in 7-1 Babcock & Wilcox

Revision 2 (5/2'5/82) a letter report 10 The small break analysis is presented in a letter report.ll These analyses used the limiting values of key parameters for all plants in the category. Furthermore, the average fuel temperature as a function of LHR and lifetime pin pressure data used in the LOCA limits analysis' are conser-vative compared to those calculated for this reload. Thus, these analyses and LOCA limits provide conservative results for the operation of Crystal River Unit 3, cycle 4.

Crystal River Unit 3's proposed long-term ECCS modification for small break LOCA is presented in reference 12.

The LOCA analysis used a power level of 2772 MWt, which is conservative rela-tive to the 2475 MWt rating. Table 7-2 shows the bounding values for allow- l2 able LOCA peak LHRs for Crystal River Unit 3, cycle 4 fuel after 50 EFPDs.

The LOCA kW/ft limits have been reduced for the first 50 EFPDs. The reduction g will ensure that conservative limits are maintained while a transition is W being made in the fuel performance codes that provide input to the ECCS analy-sis l" in order to account for mechanistic fuel densification. The limits for the first 50 EFPDs are shown in Table 7-3.

It is concluded from the examination of cycle 4 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core reload will not adversely af fect the Crystal River Unit 3 plant's ability to operate safely during cycle 4. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 4 is g bounded by previously accepted analyses. The initial ' conditions for the tran- W sients in cycle 4 are bounded by the FSAR with the exception of the four-pump coastdown and locked-rotor accidents which were re-evaluated for cycle 3 oper-ation at a core power of' 102% of 2568 MWt. The re-evaluation bounds cycle 4 operation at 2475 MWt without the pump status trip. The three pump coastdown g

has been evaluated without the pump status trip with the same flux / flow set- 2

• point at a power level of 1399 MWt with acceptable DNB results.

7.3. Dose Consequences of Accidents A complete dose evaluation was presented in the cycle 3 reload report2 based on the upgraded power level of 2544 FMt. A complete dose evaluation for cy-cle 4 was not performed because the cycle 4 doses should be essentially the same as the doses presented in the cycle 3 reload report. This conclusion is based on a comparison of the fission product inventory in the cycle 3 and I

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I 4 cores, which were calculated using the appropriate fuel data for the respec-tive cycles. The activities of the xenon, krypton, and iodine nuclides (which

,I are the nuclides that control the accident doses) in both cycles 3 and 4 were

! essentially the same. The ratio of cycle 4 to cycle 3 activities ranged from 0.985 to 1.005 for various nuclides. Thus, the cycle 4 doses are essentially the same as the doses indicated in Tables 7-7 and 7-8 of the cycle 3 relcad report.2 I

Table 7-1. Comparison of Key Parameters for Accident Analysis FSAR 1 , Predicted densif'n cycle 4 5

Cycle 1 15

.I Parameter value value BOL Doppler coeff, 10-5 Ak/k/ F -1.17 -1.47 -1.55 (268 EFPD)

EOL Doppler coeff, 10-5 Ak/k/*F -1.30 -1.66 -1.71 (510 EFPD)

BOL moderator coeff, 10 ak/k/*F 0(a) -0.75 -0.52 (268 EFPD)

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I EOL moderator coeff, 10 Ak/k/*F -4.0 -2.42 (510 EFPD)

-2.89 All-rod bank worth at EOL, HZP, 12.9 9.12 9.583

% ak/k (268 EFPD)

Boron reactivity worth (HFP), 100 101 111 ppm /1% Ak/k Max. ejected rod worth (HFP), % Ak/k 0.65 0.55 0.564 Dropped rod worth (HFP), % ak/k 0.40 0.20 0.20 Initial boron conc'n (HFP), ppm 1150 795 1090

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+0.50 x 10-" ak/k/*F was used for the moderator Jilution accident.

-3.0 x 10 ak/k/*F was used for the steam line failure analysis and dropped rod accident analysis.

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I 7-3 Babccck & Wilcox

Table 7-2. LOCA Limits, CR-3, Cycle 4 After 50 EFPD Elevation, LIIR limits, ft kW/ft 2 15.5 4

4 16.6 6 18.0 8 17.0 10 16.0 I

Table 7-3. LOCA Limits, CR-3, Cycle 4, 0-50 EFPD g

g Elevation, LIIR limits , g ft kW/ft 3 2 14.5 4 16.1 6 17.5 8 17.0 10 16.0 I

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TECHNICAL SPECIFICATIONS PROPOSED MODIFICATIONS TO lI All technical specifications have been reviewed by Florida Power Corporation and B&W and revisions were made to accommodate cycle 4 operation and revised I RPS instrument errors. Table 8-1 lists the Technical Specification changes and cross-references this reload report numbers with the Technical Specifica-tion numbers.

The re-analysis of Technical Specification for cycle 4 operation used the same I analytical techniques as the cycle 3 design.2 The 0-50 EFPD operating limits on rod index, APSR position, and axial power imbalance were established based on the interim LOCA linear heat rate limits which account for mechanistic fuel densif ication. " After 50 EFPD, the FAC LOCA LHR limits were used. The Tech-nical Specifications also provide protection for the over power condition that could occur during an overcooling transient because of nuclear instrumentation errors.

The review of the Technical Specifications based on the analysis presented in this report and the proposed modifications contained in this section, ensure that the Final Acceptance Criteria ECCS linits will not be exceeded nor will the thermal design criteria be violated.

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Revision 2 (5/25/82)

Table 8-1. Technical Specification Changes Tech Spec No. Report page Nos.

(figure, table Nos.) (figure, table Nos.) Reason for change (Figure 2.1-1) 8-4 Revised for cycle 4 opera-l (Figure 8-1) tion. W (Figure 2.1-2) 8-5 Revised for cycle 4 opera-(Figure 8-2) tion withcut pump monitors. l2 (Table 2.2-1) 8-6 Deleted nuclear overpower I

(Table 8-2) based on RCPPM trip.

(Figure 2.2-1) 8-7 Revised for cycle 4 opera-I (Figure 8-3) tion without pump monitors.

{2 2.1.1, 2.1.2 8-8 Revised text to indicate three-pump operation is more restrictive on pressure / temp l limit curve (Figure 2-1) =

2.2.1 6-9 Revised nuclear overpower E 8-9a trip value to include RPS y 5 8-10 instrument error. Deleted nuclear overpower trip based on RCPPMs.

(Figure 2.1) 8-11 Revised press / temp limit (Figure 8-4) curve.

3.1.1.1.2 8-12 Shutdown margin was revised 4.1.1.1.2.1 to reflect the required mode g 4 and 5 margin for cycle 4 3 operation to account for the inadvertant dilution by sodium hydroxide addition.

3.1.2.2 8-13 Revised mode 4 shutdown margin as above.

3.1.2.4.2 8-14 Revised mode 4 shutdown margin above.

3.1.2.7 8-15 Revised mode 4 shutdown margin as above.

3.1.2.9 8-16 Revised borated storage water volume and mode 4 shutdown margin as above 8-2 Babcock & Wilcox g m,

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I Revision 2 (5/25/82)

Figure 8-2. Reactor Core Safety Limits (Tech Spec Figure 2.1-2)

__120 I ( 26.88.112)

ACCEPTABLE 112

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m (20.16.112)

I (-47.15,100.82) 4 PUMP OPERATION __100 e (39.86,99.82) l (-26.88,85.45) ~~

n 85.45 ,

ACCEPTABLE lI ( 47.15,74.27) ( 3 & 4 PUNP OPERATION __

>(39.86,73.27)

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=

-- 60 I  %

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$-- 40 E

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-60 -50 40 30 -20 -10 0 10 20 30 40 50 60 Reactor Power imolance, ",

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Table 8-?. RPS Trip Setpoints Table 2.2-1. Reactor Protection System Instrumentation Trip Setpoints Functional unit Trip setpoint Allowable values

1. Manual reactor trip Not applicable Not applicable

2. Nuclear overpower 5 102.03% of RATED THERMAL POWER 5 102.03% of RATED THERMAL POWER -

with g pumps operating withfgrpumpsoperating 2 5 -M-fr?% of RATED THERMAL POWER s-54n of RATED THERMAL POWER with three pumps operating with three pumps operating

3. RCS outlet temp-high 5 618 F 5 618 F

4. Nuclear overpower Trip setpoint not to exceed the Allowable values not to exceed based on RCS flow and limit line of Figure 2.2-1 the limit line of Figure 2.2-1 AXIAL POWER IMBALANCEa

5. RCS pressure-lowa 1800 psig 2 1800 psig 1 co 1

6. RCS pressure-high 5 2300 prig 5 2300 psig

7. RCS pressure-variable- 2 (11.59 Tout F-5037.8) psig . (11.59 Tout F-5037.8) psig lowd

8. Deleted 1

9. Reactor containment 5 4 psig s 4 psig vessel a 5' Trip may be manually bypassed when RCS pressure s 1720 psig by actuating the shutdown bypass, 1 m provided that (1) the nuclear overpower trip setpoint is s 5% of RATED THERMAL POWER, (2) the

f. shutdown bypass RCS pressure-high trip setpoint of s 1720 psig is imposed, and (3) the shut- h a

g down bypass is removed when RCS pressure > 1800 psig. o n

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Rev,ision.2 (5/25/82)

I Figure 8-3. Reactor Trip Setpoints (Tech Spec Figure 2.2-1)

- 120

- 110

(-17,101.3) 1 01.3 (10,101.3)

Mi = .64375

- 100 M2 = .70825 I ( -33,91 )

ACCEPTABLE 4 PUMP __ 90 (26,90)

OPERATION

(-17,75.67) 75.67 -- 80 (10,75.67)

I ACCEPTABLE -- 70

(-33,65.37) 3 & 4 PUMP (26,64.37)

I OPERATION _- 60

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- 40 m

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=e 35- 10 . .

-60 40 -30 -20 -10 0 10 20 30 40 5U 60 Reactor Power Imuaiance, ",

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8-7 Babcock & Wilcox l

SAFETV LDUTS '

BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core out-let pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two ther-mal limits and account for the effects of potential fuel densification and potential fuel rod bow:

1. The 1.30 DNBR limit produced by a nuclear power peaking factor N

of F = 2.57 or the combination of the radial peak, axial peak d r.d osition of the axial peak that yields no less than a 1.30 DNBR.

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 19.7 kW/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps respective-ly.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1. g The curves of BASES Figure 2.1 represent the conditions at which a minimum g DNBR of 1.30 is predicted at the maximum possible thermal power for the num-ber of reactor coolant pumps in operation.

These curves include the potential effects of fuel rod bow and fuel densification.

l l The DNBR as calculated by the BAW-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher. Extrapo-lation of the correlation beyond its published quality range of 22% is justi- 3 fied on the basis of experimental data. E i

jl For each curve of BASES Figure 2.1, a pressure-temperature point above j' and to the left of the curve would result in a DNBR greater than 1.30 or a I

local quality at the point of minimum DNBR less than 225 for that particular g reactor coolant pump situation. The 1.30 DNBR curve for three pump operation W is more restrictive than any other reactor coolant pump combination because any pressure / temperature point above and to the left of the three pump curve will be above and to the left of the other curves.

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~. Revision 2 (5/25/82)

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l 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0INTS The Reactor Protection System Instrumentation Trip Setpoint specified in i

l I Table 2.2-1 are the values at which the Reactor Trips are set for each param-eter. The Trio Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

lg Operation with a trip setpoint less conservative than its Trip Setpoint but

!g within its specified Allowable Value is acceptable on the basis that the f difference between each Trip Setpoint and the Allowable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures. The purpose I of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower 3 than the normal low pressure trip setpoir. 50 that the reactor must be tripped lg before the bypass is initiated. The Nuc' ear Overpower Trip Setpoint of s 5.0%

prevents any significant reactor power f"om being produced. Sufficient natu-ral circulation would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentation channels and provides manual reactor trip capability.

Nuclear Overpower l A Nuclear Overpower trip at high power level (neutron flux) provides re-actor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip is initiated when the re-actor power level reaches 102.03% of rated power. Due to calibration and in-strument errors, the maximum actual power at which a trip would be actuated 2 I could be 109.15%, which was used in the safety analysis. ,

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Revision 2 (5/25782)

LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temperature - High I

The RCS Cutlet Temperature High trip < 618'F prevents the reactor outlet .

temperature from exceeding the designlirits and acts as a backup trip for all power excursion transients.

Nuclear Overpower Based on RCS Flow and AXIAL p0WER IMBALANCE I The power level trip setpoint produced by the reactor coolant system g flow is based on a flux-to-flow ratio which has been established to ac-commodate flow decreasing transients from high power. '

The power level trip setpoint produced by the power-to-flow ratio pro-vides both high power level and low flow protection in the event the re-actor power level increases or the reactor coolant flow rate decreases, g The power level setpoint produced by the pcwer-to-flow ratio provides 3 overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every pow- a er level there is a minimum pennissible low flow rate. Typical power g level and low flow rate combinations f?r the pump ' situations of Table 2.2-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is > 101.3% and reactor flow rate is 100%, or flow rate I

2 is < 96.03% and power level is 97.28%.

2. Trip would cccur when three reactor coolant pumps are opdrat-ing if power is > 75.67% and reactor ficw rate is 74.7%, or 2 'E flow rate is < ST.29% and power is 55%.

5 For safety calculations the maximum calibration and instrumentation er-rors for the power level were used. E all RC pump coastdown events.

This trip provides protection for i g I

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B-10 Babcock & Wilcox

Revision 1 (4/14/82)

LIMITING SAFETY SYSTEM SETTINGS ig BASES

5 Reactor Containment Vessel Pressure - High The Reactor Containment Vessel Pressure - High Trip Setpoint s 4 psig, I provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant acci Mnt, even in the absence of a RCS Pressure - Lov trip.

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I l ll CRYSTAL RIVER - UNIT 3 Babcock & Wilcox l 8-1%

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