ML20062N432

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Tech Spec Table 3.11-2 Re Max Critical Power Ratio Operating Limits.Update to Core Performance Analysis rept,YAEC-1275, Encl
ML20062N432
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 08/19/1982
From:
VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20062N429 List:
References
NUDOCS 8208230170
Download: ML20062N432 (9)
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Text

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ATTACHMENT 1 VYNPS l Table 3.11-2 MCPR Operating Limits MCPR Operating Limit for Value of "N" in RBM Average Control Rod Cycle Fuel Type (2)

Equation (1) Scram Time Exposure Range 8X8 8X8R P8X8R 42% Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 G1g than L.C.O. E0C-2 GWD/T to EOC-1 GWD/T 1.29 1.29 1.29 Eg 3.3 Col.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 l pm Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 C0 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31

@3 3.3 C.l.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 y 41% . . Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 0 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25 88 3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 OO. Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 TM"- than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 E 3.3 C l.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 3 < 40% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25

'j' -

than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25 3.3 C.l.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25

~..

than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C l.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 75%- Special Testing at Natural Circulation (Note 3, 4) 1.30 1.31 1.31 (1) The Rod Block Monitor (RBM) trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.

(2) The current analyses for MCPR Operating Limits do not include 7X7 fuel. On this basis further evaluation of MCPR operating limits is required before 7X7 fuel can he used in Reactor Power Operation.

(3): For the duration of pump trip and stability testing.

(4) Kg factors are not applied during the pump trip and stability testing.

180-01

ATTACllMENT 2 UPDATE TO Tile VERMONT YANKEE CORE PERFORMANCE ANALYSIS REPORT, YAEC-1275 Note: Replace pages vi, 9, 30, 49, 53, 92, and 94 with the attached.

s i

i-LIST OF FIGURES  ;

i i Number Title Page 3.2.1 Design Reference Loading Pattern, Northeast Quadrant' 7 f

I 4.2.1 Core Averaged Gap Conductance versus Cycle Burnup 13 4.2.2 Core Averaged Volume Average Temperature Versus ,

Cycle Burnup 14 l

5.1.1 VY Cycle 9 Haling Depletion EOC Bundle Average f Relative Powers 20 5.1.2 Cycle 9 Core Average Axial Power Distribution Taken from the Haling Calculation to EOFPL9 21 5.1.3 VY Cycle 9 Rodded Depletion - ARO at EOFPL9, Bundle Average Relative Powers 22' E 5.1.4 Cycle 9 Core Average Axial Power Distribution, 2

Rodded Depletion - ARO at EOFPL9 23 5.2.1 VY Cycle 9 Haling Depletion, EOC Bundle Average Exposures

~

l 24 i.

l 5.2.2 VY Cycle 9 Rodded Depletion, EOC Bundle Average Exposures 25 5.3.1 .VY Shutdown Margin for Cycle 9, Cold Percent Shutdown - l Delta K Versus Cycle- Average Exposure 26 7.1.1 Flow Chart for the Calculation of JkCPR Using RETRAN/TCPYA91 Codes' 53 4

4 7.1.2 Inserted Rod Worth and Rod Position versus Scram Time-

. at EOC, "Mbasured" Scram Time 54 7.1.3 Inserted -Rod Worth and Rod Position versus Scram Time at E0C-1000 MWD /ST, " Measured" Scram Time .55

'7.1.4- Inserted' Rod Worth and Rod Position versus Scram Time at EOC-2000 MWD /ST, "Nhasured" Scram Time 56- '

7.1.5 Inserted Rod Worth and Rod Position versus' Scram' Time:

at EOC , "67B" Scram Time :57-  ;

7.1.6' Inserted-Rod Worth and Rod Position versus Scram Time.

a t 'EOC-1000, MWD /ST , "67B" Scram _ Time 158-7.1.7 Inserted ' Rod ' Worth and Rod Position versus Scram Time at EOC-2000 MWD /ST , "67B"' Scram Time . 59' L

my e k

+

The thermal effects analysis included the calculation of fuel e

temperatures and fuel cladding gap conductance under nominal core steady state and peak linear heat generation rate conditions. Figures 4.2.1 and 4.2.2 provide the core-average response of gap conductance and fuel temperature, respectively. These calculations integrate the responses of individual fuel batch average operating histories over the core average exposure range of Cycle 9. The gap conductance values are weighted axially by power distributions and radially by volume. The core-wide gap conductance values for the RETRAN system simulations described in Section 7.1 are from this data set at the particular exposure point of interest. The fuel temperature values presented in Figure 4.2.2 are weighted axially and radially by volume.

The gap conductance value input to the hot channel (RETRAN/TCPYA91) calculations was evaluated for the P8X8R fresh fuel bundle type for the peak assembly power at the cycle exposure point of peak bundle reactivity. Cap.

conductance calculate i at this point was bounded by a value of 1000 BTU /hr-ft - F. 'With consideration for the hot channel transient response i to bundle power level and gap conductance values calculated for all other fuel I

types in Cycle'9, a gap conductance value of 1000 BTU /hr-ft - F was utilized for all' hot channel calculations at all exposure points and for all fuel bundle types.

-Fuel rod local linear heat generation rates-at fuel centerline incipient' i

j melt were calculated as a function of local axial segment ~ exposure for.the

[ gadolinia concentrations in Vermont Yankee fuel bundles and are displayed'in i

i

-~,

The hot channel calculations are performed with the RETRAN and TCPYA$1 [14] computer codes. The CEXL correlation [11] is used in TCPYA01 to evaluate critical power ratio. The calculational procedure is outlined below.

l The hot channel transient ACPR calculations are performed via a series of

" inner" and " outer" iterations , as illustrated by the flow chart in Figure 7.1.1. The outer loop represents iterations on the hot channel initial power level. These iterations are necessary because the ACPR for a given transient varies with Initial Critical Power Ratio (ICPR), yet only the ACPR corresponding to a transient MCPR equal to the safety limit (i.e. ,1.07 + ACPR

= ICPR) is appropriate. The approximate constancy of the NCPR/ICPR ratio is useful in these iterations. Each outer iteration requires a RETRAN hot channel run to calculate the transient enthalples, flows, pressure and saturation properties at each time-step required for input to the TCPYA91 code. TCPYA61 is then used to calculate a CPR at each time-step during the transient , from which a transient ACPR is derived. The hot channel is modeled ucing a chopped cosine axial power shape with a peak / average ratio of 1.4.

The inner loop represents iterations on'the. hot channel inlet flow.

These iterations are necessary because the RETRAN hot channel model calculates an exit loss coefficient when given the initial power level, flow, and pressure drop as input. The pressure. drop is assumed equal to the core average pressure drop, and the flow is varied. for a given power level until the exit loss coefficient is correct. FIBWR [9] is utilized to estimate the correct inlet flow for a particular power level and pressure drop. -

Y = 'i

TABLE 7.2.1 CORE WIDE TRANSIENT ANALYSIS RESULTS Peak Peak Avg.

Prompt Power Heat Flux (Fraction of (Fraction of ACPR Transient Exposure Initial Value) Initial Value) 8X8 8X8 R/ P8X8R Turbine Trip EOC 3.50 1.23 0.23 0.23 Without Bypass,

" Measured" E0C-1000 2.86 1.18 0.19 0.18 Scram Time EOC-2000 2.20 1.10 0.09 0.09 Turbine Trip EOC 3.97 1.28 0.29 C.28 Without By pa s s ,

"67B" EOC-1000 3.38 1.24 0.24 0.23 Scram Time E0C-2000 2.79 1.15 0.16 0.15 Generator Load EOC 3.35 1.22 0.23 0.22 Rejection Without Bypass, EOC-1000 2.78 1.17 0.19 0.18

" Measured" Scram Time EOC-2000 2.08 1.08 0.08 0.08 Cenerator Load EOC 3.93 1.27 0.29 0.28 Rejection Without Bypass , EOC-1000 3.40 1.23 0.26 0.24 "67B" Scram Time E0C-2000 2.78 1.14 0.17 0.16 Loss of 100 F EOC 1.21 -

0.16 0.16 Feedwater E0C-1000 1.22 -

0.18 0.18 Heating EOC-2000 1.23 1.22 0.18 0.18 BOC 1.21 -

0.16 0.16 Choose ICPR l

Estimate Power T If Estimate Flow with FIBWR 1r RETRAN Flow ,

Initialization Run Is Exit Loss Coefficient Revise Correct? Flow 3r Yes RETRAN/TCPYA01 Hot Channel Run Has ACPR No Converged?

,Yes STOP FIGURE 7.1.I FLOW CilART FOR THE CALCULATION OF ACPR USING RETRAN/TCPYA01 CODES TABLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 9 MCPR OPERATION LIMITS MCPR Operating Limit f or Vclue of "N" uin RBM ' Average Control Rod Cycle Fuel Type (2)

Equation (1) Scram Time Exposure Range 8X8 8X8R P8X8R 42% Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.29 1.29 1.29 3.3 C.l.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C.1.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 41% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. E00-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25

'3.3 C.l.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25

. than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 e- 3.3 C l.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35

" 1.25 1.25 1.25

<40%- Equal or better BOC to EOC-2 GWD/T 1.25 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T- 1.26  ;

-3.3 Col.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 .

than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C l.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35

,75%. Special Testing at Natural Circulation (Note 3, 4) 1.30 1.31 1.31

'(1) The Rod. Block Monitor (RBM); trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.

l(2) .The current analyses for MCPR Operating Limits do not include 7X7 fuel. On this basis further evaluation of MCPR operating limits is ' required before 7X7' fuel can be used in Reactor Power Operation.

(3) -For the duration of. pump trip and stability' testing.

(4) Kf factors are not. applied during the pump trip.and stability testing.

_ - . - - . . . . p

. REFERENCES

1. B. Buteau, " Stability and Recirculation Pump Trip Test , Special Test Procedure No. 81-01", Vermont Yankee Nuclear Power Corporation ,

February 1981.

2. General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P- , GE Company Proprietary , July 1979.
3. D. M. VerPlanck, Methods for the Analysis of Boiling Water Reactors Steady State Core Physics, YAEC-1238, March 1981.
4. E. E. Pilat , Methods for the Analysis of Boiling Water Reactors Lattice Physics: VAEC-1232, December 1980.
5. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code /Model Description Manual, YAEC-1249P, April 1981. ,
6. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code Qualification and Application , YAEC-1265P , June 1981.
7. D. C. Albright, H20DA: An Improved Water Properties Package , YAEC-1237, Nbech 1981.
8. Appendix A to Operating License DPR-28 Technical Specifications and Bases for Vermont Yankee Nuclear Power Station, Docket No. 50-271.
9. A. A. F. Ansari, Methods for the Analysis of Boiling Water Reactors:

Steady-State Core Flow Distribution Code (FIBWR) , YAEC-1234, December 1980.

10. A. A. F. Ansari , R. R. Gay , and B. J. Gitnick , FIBWR: A Steady-State Core Flow Distribution Code for Boiling Water Reactors - Code Verification and Qualification Report , EPRI NP-1923, Project 1754-1 Final Report, July 1981.
11. General Electric Company , GEXL Correlation Application to BWR 2-6 Reactors , NEDE-25422, GE Company Proprietary , June 1981.
12. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling Water Reactors: A Systems Transient Analysis Model (RETRAN) , YAEC-1233, April 1981.
13. EPRI, RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems , CCM-5, December 1978.
14. A. A. F. Ansari, K. J. Burns and D. K. Beller, Methods for the Analysis of Boiling Water Reactors: Transient Critical Power Ratio Analysis

( RETRAN-TCPYA01 ) , YAEC-1299P , Ma rch 1982.

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