ML17249A374

Cycle 10 SAR W/Mixed Oxide Assemblies.

ML17249A374
Person / Time
Site:	Ginna
Issue date:	12/14/1979
From:	Busselman G, Johnson B, Sofer G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17249A368	List: ML17249A367 ML17249A369 ML17249A370 ML17249A372 ML17249A373 ML17249A374
References
XN-NF-79-103, NUDOCS 7912280240
Download: ML17249A374 (42)

Text

0 KN-N 103 P E (lllklk PlljlllLEkk PILklll'NllllLEJIB MFET7 kkkL7SIIS PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA 99352

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0: XN-NF-79-103 IR/R14 79 R. E. GINNA NUCLEAR PLANT CYCLE 10.

SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared:

G. J. Buss man, Manager Neutronics and Fuel Management Approved:

G. A . Sofe nager Nuclear Fue s Engineering Concurred:

J. N. Morgan, Manage Licensing and Safety Engineering Concurred: l'~/ 7 L. J. Federico, Manager Nuclear Fuels Project E)j(ON NUCLEAR COMPANY, Inc.

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being Sub.

mitted by Exxon Nuclear to the USNRC as part of a technical contri.

bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear. fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors anH, with the measurement uncertainties backed out, are 2.15 and 1.60. Additionally the predicted axial F distri-butions are well below the axially dependent Technical Specification limits N

on F~. The BOC F value of 1.745 compares with the measured Cycle 9 value in Table 5.1 of 1.758.

12 XN-NF 79-,103 The control of the core power distribution is accomplished by following the procedures as discussed in the report, XN-76-40, "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors", September 1976 and its addendum. The results reported in these documents demonstrate that the Power Distribution Control (PDC) procedures defined in the report will protect an axially dependent F limit with a peak value of 2.30. The Technical Specification limit for R. E. Ginna has a peak of 2.32 and an axial dependence identical to that supported by the procedures. The physics characteristics of the Ginna Cycle 10 core are similar to those utilized in the PDC supporting analysis. The Ginna Technical Specification limits on F can therefore be protected by operation under the PDC procedures as stated in XN-76-40.

5.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 10 are compared with Cycle 9 data in Table 5.2. The ENC Plant Transient Simulation (PTS) Analysis indicates that the minimum required shutdown margin is 1,800 pcm based upon the steamline break accident analyzed for ENC fuel at the EOC conditions. A value of 1,900 pcm is used at EOC in the evaluation of the shutdown margin to be consistent with'the Technical Specifications. The Cycle 10 analysis indicates excess shutdown margins of 1,414 pcm at the BOC and 344 pcm at the EOC. The Cycle 9 analysis indicates excess shut-down margins for that cycle of 1,795 pcm at the BOC and 393 pcm at the EOC.

The slightly lower Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to slightly lower calculated rod worths.

13 XN-NF-79 103 The control- rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 9. With these limits the'nominal worth of the control bank, D-bank, inserted to the insertion limits'at HFP is 122 pcm at, BOC and'70 pcm at EOC. The control rod shutdown requirements in Table 1

5.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both BOC and EOC.

5.1.3 Moderator Tem erature Coefficient Considerations The reference Cycle 10 design calculations indicate that the moderator temperature coefficient is negative at all times during the cycle as shown in Table 5.1. This meets the Technical Specification requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be nega-tive. The least negative moderator temperature coefficient occurs at BOC HZP and is -2.0+2pcm/ F. This compares with the BOC9 HZP value of -2.0 pcm/ F.

5.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 10 core analyses are described in References 6, 7, and 8. These methods have been verified for both U02 and Pu02-U02 lattices. In summary, the reference neutronic design analysis of the reload core was performed using the XTG (Reference 9) reactor simulator system-. The input exposure data were based on quarter core depletion calcu-lations performed from Cycle 5 to Cycle 9 using the XTG code. The BOC5 exposure distribution was obtained from plant data. The fuel shuffling between c'ycles was accounted for in the calculations.

14 XN-NF-79-103 Predicted values of F~, Fx , and F were studied, with the XTG reactor model. The calculational thermal-hydraulic feedback and axial exposure distribution effects on power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.

15 XN-NF-79-103 Table 5.1 R. E. Ginna Neutronics Characteristics of Cycle 10 Compared with Cycle 9 Data C cle 9 C cle 10 BOC EOC BOC EOC

(

Critical Boron

')

HFP, ARO, Equilibrium Xenon (ppm) 961 12( 921 HZP, ARO No Xenon (ppm) 1,41 0(2) 1,414 Moderator Temperature Coefficient HFP, (pcm/oF) -7.6(2) -30.4 -8.1 -30.4 HZP, (pcm/oF) -2.0 -21.5 -2.0 -21.6 Doppler Coefficient, (pcm/ F) -1.25 to -2.0 -,1.35 -1.84 Boron Worth, (pcm/ppm)

HFP -8.12 -8.72 -7.95 ,-8.62 HZP -8.58 Total Nuclear Peaking Factor Fq, HFP 1.758( 1.745 1.517 Delayed Neutron Fraction .0061 .0051 .0058 .0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm) 5,751 ) 5,821. 5';341 5,696 Excess Shutdown Margin (pcm) 1,795( ) 393( 1,414 344 Moderator Pressure Coefficient (pcm/psi) 0.35 0.35 (1) Extrapolated from measured data (2) Measured Data (3) 70/ Power Map (4) Reference 5

16 XN-NF-79-103 Table 5.2, R. E. Ginna Control Rod Shutdown Margins and Requirements for Cycle 10 C cle 9 ** C cle 10 BOC EOC BOC EOC Control Rod Worth HZP , cm All Rods Inserted (ARI) 6,407 6,634 5,949 6,420 ARI less most reactive (N-1) 5,751 5,821 5,341 5,696 N-1 less lOX allowance L(N-1)* 9l 5,176 5,239 4,807 5,125 Reactivit Insertion cm Moderator plus Doppler 1,431 1,996 1,443 1,932 Flux Redistribution 600 600 600 600 Void 50 50 50 50 Sum of the above three 2,081 2,'646 2,093 2,582 Rod Insertion Allowance 300 300 300 300 Total Requirements 2,381 2,946 2,393 2,882 Shutdown. Margin (N-l)*.9 - Total Requirements 2,795 2,293 2,414 2,244 Required Shutdown Margin* 1,000 1,900 10000 1,900 Excess Shutdown Margin 1,795 393 1,414 344

• Technical Specification 3.10

"* Calculated values from Reference 5

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20 XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.

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21 XN-NF-79-103 7.0 ACCIDENT AND TRANSIENT ANALYSIS 7.1 PLANT TRANSIENT AND ECCS ANALYSES fOR R. E. GINNA The ECCS analysis provided in Reference 3 is applicable to all ENC fuel residing in the core during Cycle 10 operation.

(2) for the The Plant Transient Analysis reported in XN-NF-77-40 R.

E. Ginna plant was intended to cover all anticipated ranges of values for all significant fuel dependent plant parameters for Cycle 8 and for all future 7 reloads. Table 7.1 presents a comparison of the kinetic parameters used in the Plant Transient Analysis and the parameters calculated specifically for Cycle 10. Due to the introduction of the 4 MOX assemblies the reactivity worth of the -boric acid used by the HPSIS (High Pressure Safety Injection System) and the BOC delayed neutron fraction have been calculated to be outside the range reported in the XN-NF-77-40 analysis. The analysis was reviewed and it was found that the change in boric acid worth affects the smal,l and large steamline break transients and that the delayed neutron fraction most affects the fast uncontrolled rod withdrawal transient.

The enveloping data for both steamline breaks are the EOC data and for the fast uncontrolled rod withdrawal are BOC data. The impact of the Cycle 10 parameters (see Table 7.1) have been evaluated for each of the transients. The results of the evaluation for the transients were found to be nearly equivalent to the previous results and that the figure of merit for the transients were not violated, i.e. for the small steamline break the system does not go critical, for the large steamline break the MDNBR is greater than the 1.30 limit and for the uncontrolled rod withdrawal the MDNBR margin is not altered.

22 XN-NF-79-103 7.2 ROD EJECTION ANALYSIS FOR R. E. GINNA CYCLE 10 A Control Rod Ejection Accident is defined as the mechanical fail-ure of a control rod mechanism pressure housing, resulting'in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaft. The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel damage.

~The rod ejection accident analysis presented in the document XN-NF-78-53 is still applicable to Cycle 10 operation. The location of the 4 MOX assemblies introduces minimal effects on ejected rod worths and hot pellet peaking factors. The e'jected rod worths and hot pellet peaking factors are cal'culated using the XTG code. No credit was taken for the powei flattening effects, of Doppler or moderator feedback in the calculation of ejected rod worths'r peaking 'factors. The calculations made for Cycle 10 using XTG were two-dimensional (x-y) with appropriate axial buckling correc-tion term's. The total'eaking factors (F~) were determined as the product of the radial peaking facto'r (as calculated using XTG) and a conservative axial peaking factor; The pellet energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.

The rod ejection accident was found to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77 'and provides a greater energy deposition marg'in than that determined by Reference 4. The results of the control rod ejection transient for this case are presented in Table 7.2 along with results 'from References 4 and 5.

23 XN-NF-79-103 7.1 R. E. Ginna Kinetic Parameters Reference Cycle (1) Cycle 10 Parameters BOC E C B C E C

, Moderator Temperature Coefficient (pcm/oF) 0.0 -35.0 -8.1 -30.4 Moderator Pressure Coefficient (pcm/psia) +.25 +.35 +.09 +.35 Moderator Density Coefficient (pcm/gm/cm3) 0.0 +29635.0 +6858.0 +25740.0 Doppler Coeffi:cient (pcm/ F) -1.25 -2.00 -1.35 -1.84 Boron Worth Coefficient (pcm/ppm) -8.75 -8.72 -7.95 -8.62 Delayed Neutron Fraction .0061 .0051 .0058 .0052 Reference 2

XN-NF-7.9.-103 Table 7.2 Ejected Rod Worth and Peaking Factors'~

~C1 8( )

~C1 9( )

~C1 10(

HFP HZP HFP HZP HFP HZP Before Ejection 2,25 2.82 2.24 2.62 2.15 2.59 N

F~ After Ejection 4.36 '.30 2.96 5.59 g g4( ) 6 01(

Maximum Rod Worth from a Full Inserted Bank (X hp) 0.470 0.640 0.362 0.553 0.280 0.435 Energy Deposition (cal/gm) 171 37 (1) Includes a conservative estimate of F at HFP of 1.4 and at HZP of 1.8.

(2) Reference 4, calculated with XTRAN.

(3) Reference 5, calculated with XTGPWR.

(4) Calculated with XTGPWR.

25 XN-NF-79-103

8.0 REFERENCES

1. XN-NF-77-52, "R. E. Ginna Reload Fuel Design", November, 1977.

2. 'XN-NF-77-40, "Plant Transient Analysis for R. E. Ginna, Unit 1 Nuclear Power Plant", Revision 1, July, 1979.

3. XN-NF-77-58, "ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-II PWR Evaluation Model", December, 1977.

4. XN-NF-77-53, "R. E. Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.

5. XN-NF-78-50, ="R. E. Ginna Cycle 9 Safety Analysis Report,"

December, 1978.

6. F. B. Skogen, "Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.

7. XN-75-27(A), Supplement 1 to Reference 6, April, 1977.

8. XN-75-27, Supplement 2 to Reference 6, December, 1977.

9. XN-CC-28, Rev. 3, "XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.

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26 XN-NF-79-103 R. E. GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K. H. Blank L. A. Nielsen G. J. Busselman G. F. Owsley L. J. Federico J. F. Patterson R. L. Feuerbacher A. W. Prichard R. G. Grummer F. B. Skogen B. L. Johnson (2) G. A. Sofer M. R. Killgore A. V. Wojchouski T. L. Krysinski C. H. Wu C. E. Leach RG&E/L. J. Federico (80)

J. N. Morgan Document Control (10)

W. S. Nechodom

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