NOC-AE-13003060, Cycle 17 Core Operating Limits Report

Cycle 17 Core Operating Limits Report

ML13358A389
Person / Time
Site:	South Texas
Issue date:	12/05/2013
From:	Dunn R South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-13003060
Download: ML13358A389 (18)

Text

Nuclear Operating Company South Texas ProjectElectric GeneratingStation PO Box 289 Wadsworth. Texas 77483 -

December 5, 2013 NOC-AE-13003060 10 CFR 50.36 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 2 Docket No. STN 50-499 Unit 2 Cycle 17 Core Operatinq Limits Report In accordance with Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 2 Cycle 17. The report covers the core design changes made during the 2RE16 refueling outage.

There are no commitments in this letter.

If there are any questions regarding this report, please contact Ken Taplett at (361) 972-8416 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis web

Attachment:

Unit 2 Cycle 17 Core Operating Limits Report, Revision 0 STI: 33793940

NOC-AE-13003060 Page 2 of 2 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A.H. Gutterman, Esquire U.S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal U.S. Nuclear Regulatory Commission Balwant K. Singal Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (MS 8 B1) Jim von Suskil 11555 Rockville Pike NRG South Texas LP Rockville, MD 20852 Kevin PolIo NRC Resident Inspector Richard Peha U.S. Nuclear Regulatory Commission City Public Service P.O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 Peter Nemeth Crain Caton & James, P.C.

Jim Collins City of Austin C. Mele Electric Utility Department City of Austin 721 Barton Springs Road Austin, TX 78704 Richard A. Ratliff Robert Free Texas Department of State Health Services

M M=

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 2 Cycle 17 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

A II3l Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 2 Cycle 17 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.6.

The Technical Specifications affected by this report are:

1) 2.1 SAFETY LIMITS

2) 2.2 LIMITING SAFETY SYSTEM SETTINGS

3) 3/4.1.1.1 SHUTDOWN MARGIN

4) 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT LIMITS

5) 3/4.1.3.5 SHUTDOWN ROD INSERTION LIMITS

6) 3/4.1.3.6 CONTROL ROD INSERTION LIMITS

7) 3/4.2.1 AFD LIMITS

8) 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR

9) 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

10) 3/4.2.5 DNB PARAMETERS 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented below.

2.1 SAFETY LIMITS (Specification 2.1):

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 1.

2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2):

2.2.1 The Loop design flow for Reactor Coolant Flow-Low is 98,000 gpm.

Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 3 of 16 2.2.2 The Over-temperature AT and Over-power AT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values

-1 measured reactor vessel AT lead/lag time constant, T, = 8 sec T2 measured reactor vessel AT lead/lag time constant, r2 = 3 sec 13 measured reactor vessel AT lag time constant, "r3 = 2 sec "r4 measured reactor vessel average temperature lead/lag time constant, -r4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, T5 = 4 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec K1 Overtemperature AT reactor trip setpoint, K1 = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K 2 = 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.00143/psi T' Nominal full power Tavg, T'__ 592.0 'F P' Nominal RCS pressure, P' = 2235 psig fl(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For q,- qb between -70% and +8%, f1(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q,+ qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb exceeds -70%, the AT Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q,- qb exceeds +8%, the AT Trip Setpoint shall be automatically reduced by 2.65% of its value at RATED THERMAL POWER.

Over-power AT Setpoint Parameter Values Tl measured reactor vessel AT lead/lag time constant, ul = 8 sec T2 measured reactor vessel AT lead/lag time constant, "12 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T6 measured reactor vessel average temperature lag time constant, 'r6 = 2 sec 17 Time constant utilized in the rate-lag compensator for Tavg, -;7 = 10 sec K4 Overpower AT reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, K5 = 0.02/'F for increasing average temperature, and K 5 = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K6 = 0.002/ 0 F for T>T",and K6 = 0 forTh_ T" T" Indicated full power Tavg, T":5 592.0 'F f2(AI) = 0 for all (AI)

dLW llUnit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 ON VMr Page4 of 16 2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN MARGIN shall be:

2.3.1 Greater than 1.3% Ap for MODES 1 and 2*

• See Special Test Exception 3.10.1 2.3.2 Greater than the limits in Figure 2 for MODES 3 and 4.

2.3.3 Greater than the limits in Figure 3 for MODE 5.

2.4 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3):

2.4.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure 4.

2.4.2 The EOL, ARO, HFP, MTC shall be less negative than -62.6 pcm/°F.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/IF (300 ppm Surveillance Limit).

Where: BOL stands for Beginning-of-Cycle Life, EOL stands for End-of-Cycle Life, ARO stands for All Rods Out, HFP stands for Hot Full Power (100% RATED THERMAL POWER),

HRP vessel average temperature is 592 'F.

2.4.4 The Revised Predicted near-EOL 300 ppm MTC shall be calculated using the algorithm from Technical Specification 6.9.1.6.b. 10:

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcm/°F If the Revised Predicted MTC is less negative than the COLR Section 2.4.3 limit and all of the benchmark data contained in the surveillance procedure are met, then an MTC measurement in accordance with S.R. 4.1.1.3b is not required.

2.5 ROD INSERTION LIMITS (Specification 3.1.3.5 and 3.1.3.6):

2.5.1 All banks shall have the same Full Out Position (FOP) of either 254 or 259 steps withdrawn.

2.5.2 The Control Banks shall be limited in physical insertion as specified in Figure 5.

2.5.3 Individual Shutdown bank rods are fully withdrawn when the Bank Demand Indication is at the FOP and the Rod Group Height Limiting Condition for Operation is satisfied (T.S. 3.1.3.1).

A lAlFAM Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 5 of 16 2.6 AXIAL FLUX DIFFERENCE (Specification 3.2.1):

2.6.1 AFD limits as required by Technical Specification 3.2.1 are determined by Constant Axial Offset Control (CAOC) Operations with an AFD target band of +5, -10%.

2.6.2 The AFD shall be maintained within the ACCEPTABLE OPERATION portion of Figure 6, as required by Technical Specifications.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):

2.7.1 F' = 2.55.

2.7.2 K(Z) is provided in Figure 7.

(ERTP* ihnseii oepae 2.7.3 The Fxy limits for RATED THERMAL POWER (Fxy within specific core planes shall be:

2.7.3.1 Less than or equalto 2.102 for all cycle burnups for all core planes containing Bank "D" control rods, and 2.7.3.2 Less than or equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.

2.7.3.3 PFxy = 0.2.

These F., limits were used to confirm that the heat flux hot channel factor FQ(Z) will be limited by Technical Specification 3.2.2 assuming the most-limiting axial power distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying variations in the axial xenon and power distributions, as described in WCAP-8385. Therefore, these FXY limits provide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 10 CFR 50.46.

2.7.4 Core Power Distribution Measurement Uncertainty for the Heat Flux Hot Channel Factor 2.7.4.1 If the Power Distribution Monitoring System(PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and Fy(Z) using the PDMS shall be calculated by:

UFQ = (1.0 + (UQ/lOO))*UE Where:

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced by Technical Specification 6.9.1.6.b. 11 UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System (PDMS).

COWA Nuclear Operating Company Unit 2 Cycle 17 Core Operating Limits Report Rev. 0 Page 6 of 16 2.7.4.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and F, (Z) shall be calculated by:

UFQ = UQU*UE Where:

UQU = Base FQ measurement uncertainty of 1.05.

UE = Engineering uncertainty factor of 1.03.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):

2.8.1 F7R = 1.62' 2.8.2 PFAH = 0.3 2.8.3 Core Power Distribution Measurement Uncertainty for the Enthalpy Rise Hot Channel Factor 2.8.3.1 If the Power Distribution Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distribution measurement uncertainty (UFAH) to be applied to the FN using the PDMS shall be the greater of:

UFAH = 1.04 OR UFAH = 1.0 + (U/I100)

Where:

UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced in Technical Specification 6.9.1.6.b.11.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System.

2.8.3.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UFAH) shall be:

UFAH = 1.04 Applies to all fuel in the Unit 2 Cycle 17 Core.

A [in PlD3 Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 7 of 16 2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 The following DNB-related parameters shall be maintained within the following limits:

2.9.1.1 Reactor Coolant System Tavg -s 595 OF2 2.9.1.2 Pressurizer Pressure > 2200 psig 3, 2.9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm4 .

3.0 REFERENCES

3.1 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 2 Cycle 17 Final Reload Evaluation" NF-TG-13-69 (ST-UB-NOC-13003355 dated October 15, 2013.

3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3 STPNOC Calculation ZC-7035, Rev. 2, "Loop Uncertainty Calculation for RCS Tavg Instrumentation," Section 10.1.

3.4 STPNOC Calculation ZC-7032, Rev. 5, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9.

3.5 5Z529ZB0 1025 Rev. 4, Design Basis Document, Technical Specifications /LCO, Tech Spec Section 3.2.5.c.

3.6 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Units 1 and 2 Documentation of the fl(Al) Function in OTAT Setpoint Calculation," NF-TG- 11-93 (ST-UB-NOC- 11003215) dated November 10, 2011.

3.7 Document RSE-U2, Rev. 3, "Unit 2 Cycle 17 Reload Safety Evaluation and Core Operating Limits Report." (CR Action 11-30675-9)

A discussion of the processes to be used to take these readings is provided in the basis for Technical Specification 3.2.5.

2 Includes a 1.9 'F measurement uncertainty per Reference 3.3, Page 37.

3 Limit not applicable during either a Thermal Power ramp in excess of 5% of RTP per minute or a Thermal Power step in excess of 10% RTP. Per Technical Specification 3.2.5 Bases, this includes a 10.7 psi measurement uncertainty as read on the QDPS display, which is bounded by the 9.6 psi averaged measurement calculated in Reference 3.4.

4 Includes the most limiting flow measurement uncertainty of 2.8% from Reference 3.5.

Unit 2 Cycle 17 uclear Operating Company Core Operating Limits Report Rev. 0 Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 0

Uj 600 580 560 540 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

• l llUnit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 A - NoPage 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 Acceptable I I I I I I 1 1- 1 1 1 1 1 1 1 1 1 1 1 1 1(24005 15)1 5

4

.0 - - - - - - -

2 . - -

LO. 1.30)

Unacceptable~

=10'1-1.

"0"ý1) 3 1.0 ____ I - - *..........I..............

- - i..............

__ I__

0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

A 3 ID C~olllny Unit 2 Cycle 17 Nuciear Operating Company Core Operating Limits Report Rev. 0 Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 4.0 3.0 2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

dlIT AM Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 AN Page 11 of 16 Figure 4 MTC versus PowerLevel 7.0 6.0 Unacceptable 5.0 4.0 tz 0u Acceptable 3.0 U

2.0 06 1.0 0.0

-1.0

-2.0

-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

A plli 3a Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 I LLJ I I (79,259): 122 Step Overlap

( 23, 259 ): 122 Step 0 I S Verlinn [ I I I I [ I ] I

( 77 9*AV* vr 1J~1IR~*

I

-- 25 1 2 0 :1St 240 or. 77 24 17 te ()e OFi

• i OF 220 30B1 Ban I

AO 200

-1 .4 ,- -- 100,17 )

180 III 'I."

r 160 I "or

.2 140 z i i6 120 100 I 1 1 1 ~~1I

' 0 1 . ..

80 .00

.00,/

65) - -

60

00 00

i 40 z 1.*1 Control Bank A is already withdrawn to Full Out Position.

20 "1 Fully withdrawn shall be the condition where shutdown and

-/1 J control banks are at the position of either 254 or 259 steps I L I J- -withdrawn.

0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

I l n CompanyUnit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of16 Figure 6 AFD Limits versus Power Level 120 110 100 L (-11,90) 11 '90) 90 80 Unacceptable nacceptable Operation Operation 70 Acceptabl IN I I Oper ation 0

60

[-

50 11 1 1 1 -31 50) 1 1 1 -- 1(31 50)1 40 30 20 10 0

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Flux Difference (% Delta-I)

Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 1.1 I ~

. I I I II. . . . . I. . I . I II I. I. I I. . . .4 11 II1 1 1 111 1 1I 1 1 1 1 1 ; 1 1 1 I1 I-1.0

.1.1 11 11 1 11 1 Iii IIi IIi II I ILIii iii IL! !1-= .L.L1 I 0.9 I I 0.8 I I I I II .  ; I Ii I i I I ! I ! I I I I i I ; iI I I I I I I I I I II I I I II II I I I I I.F I I I iII I I1I I I II I II I III I I I II I I I_ IFI . I I . . . I

' 0.7

.I F .I I  ! ; I II ; I I I I I I I 1 I I I I I II I I I II I! I -I

• 0.6

-- tore Elev. (iff) FQ K(Z) 0.0 2.55 1.0

.t 0.5 7.0 2.55 1.0 14.0 2.359 0.92 z

0.4 1 1 IIJ IT FI I I I 11 11 111 1 11 11 11 -L I I I I i i i i II i I i i i I I ! I I I I II I I !II I i I I II I I I I iI IIi I II I I I I I I 4 II I ii . .

0.3 i  ! ! I I :1  ! i i i i I . .ii. ii I I i i ij !  ! i I ! iA 0.2 I I III i*

• I '~II1 1 1 1 11 1 1 1 1 1lllI11Il 1 1 1 1 1 1 1 1111111 M P

i i i ! i i i i i i i i i i iiii i i i i i i i i i i i i i j i I:i i ;ii i i I II i i :1 iii 11 !i l  ! 1 "! 4 - I1 1 1I 1 -1 I 1I 1 I I !i l I I!1 I I I I I i I I I I I i 0.1 I I i i i ! i i I ii I !I !I  ! I i I I i II I  ! i Ii i I Ii Ii ; ; I I I II I II i i i ii ii iI IiIi i IIi i I Ii i i i i i - -i iI i I I ' Il I ..I .II

. . .I. I. . . . . .II 0.0 I I 7 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

&IF" i*  !*Unit 2 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 W AW ff - Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.0 1 11.314 6.8 37 2.021 13.8 2 8.544 6.6 38 2.076 13.6 3 5.774 6.4 39 2.055 13.4 4 3.313 6.2 40 2.010 13.2 5 2.661 6.0 41 1.989 13.0 6 2.350 5.8 42 1.990 12.8 7 2.203 5.6 43 1.995 12.6 8 2.19 - 5.4 44 1.999 12.4 9 2.088 5.2 45 2.029 12.2 10 2.043 5.0 46 2.073 12.0 11 2.015 4.8 47 2.082 11.8 12 2.011 4.6 48 2.044 11.6 13 2.035 4.4 49 2.002 11.4 14 2.044 4.2 50 2.003 11.2 15 2.019 4.0 51 2.000 11.0 16 1.985 3.8 52 1.990 10.8 17 1.955 3.6 53 1.991 10.6 18 1.934 3.4 54 2.023 10.4 19 1.929 3.2 55 2.055 10.2 20 1.944 3.0 56 1.982 10.0 21 1.978 2.8 57 1.919 9.8 22 1.998 2.6 58 1.926 9.6 23 1.972 2.4 59 1.896 9.4 24 1.957 2.2 60 1.845 9.2 25 1.963 2.0 61 1.822 9.0 26 1.975 1.8 62 1.859 8.8 27 1.985 1.6 63 1.881 8.6 28 2.007 1.4 64 1.785 8.4 29 2.065 1.2 65 1.747 8.2 30 2.121 1.0 66 1.800 8.0 31 2.096 0.8 67 2.048 7.8 32 2.056 0.6 68 2.703 7.6 33 2.043 0.4 69 3.829 7.4 34 2.028 0.2 70 5.444 7.2 35 2.004 0.0 71 7.284 7.0 36 1.985

MPIrAM Nuclear Operating Company Unit 2 Cycle 17 Core Operating Limits Report Rev. 0

- iE- Page 16 of 16 Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 8.689 6.80 37 2.181 13.80 2 6.787 6.60 38 2.216 13.60 3 4.885 6.40 39 2.196 13.40 4 3.195 6.20 40 2.146 13.20 5 2.725 6.00 41 2.113 13.00 6 2.399 5.80 42 2.098 12.80 7 2.185 5.60 43 2.084 12.60 8 2.091 5.40 44 2.072 12.40 9 2.039 5.20 45 2.085 12.20 10 2.013 5.00 46 2.120 12.00 11 2.012 4.80 47 2.127 11.80 12 2.029 4.60 48 2.073 11.60 13 2.067 4.40 49 2.016 11.40 14 2.074 4.20 50 2.017 11.20 15 2.054 4.00 51 2.006 11.00 16 2.038 3.80 52 1.992 10.80 17 2.032 3.60 53 1.991 10.60 18 2.034 3.40 54 2.017 10.40 19 2.033 3.20 55 2.054 10.20 20 2.060 3.00 56 2.005 10.00 21 2.107 2.80 57 1.939 9.80 22 2.129 2.60 58 1.917 9.60 23 2.103 2.40 59 1.914 9.40 24 2.084 2.20 60 1.906 9.20 25 2.086 2.00 61 1.920 9.00 26 2.090 1.80 62 1.965 8.80 27 2.090 1.60 63 1.974 8.60 28 2.101 1.40 64 1.920 8.40 29 2.152 1.20 65 1.924 8.20 30 2.200 1.00 66 1.981 8.00 31 2.166 0.80 67 2.252 7.80 32 2.142 0.60 68 2.597 7.60 33 2.138 0.40 69 2.975 7.40 34 2.142 0.20 70 3.648 7.20 35 2.149 0.00 71 5.207 7.00 36 2.153