ML17261A870
| ML17261A870 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/28/1989 |
| From: | Skaritka J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17261A867 | List: |
| References | |
| NUDOCS 8903080049 | |
| Download: ML17261A870 (20) | |
Text
RELOAD SAFETY EVALUATION R. E. GINNANUCLEAR PLANT CYCLE 19 February 1989 Edited by: J. Skaritka Approved:
E. H. Novendstern, Manager T/H Design and Fuel Licensing Commercial Nuclear Fuel Division 8903080049 890224 PDR Al3OCK 05000244 P
PDC 3702F:6/890202
R; E. Glnna, Cycfe 19 e
TABLEOF CONTENTS Section Title Page
1.0 INTRODUCTION
AND
SUMMARY
1.1 Introduction 1.2 General Description 1.3 Conclusion 2.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Mydraulic Design 3.0 POWER CAPABILILTYAND ACCIDENT EVALUATION 3.1 3.2 Power Capability Accident Evaluation 3.2.1 Kinetic Parameters 3.2.2 Control Rod Worths 3.2.3 Core Peaking Factors
4.0 REFERENCES
R. E. Glnna, Cycle 19 January 1989 LIST OF TABLES Table Title Page R. E. Ginna Cycle 19 Fuel Parameters 10 Kinetic Characteristics R. E. Ginna Cycle 19 12 Shutdown Requirements and Margins 13 Thermal-Hydraulic Design Bases for R. E. Ginna Cycle 19 LIST OF FIGURES Figure Title Page R. E. Ginna Cycle 19 Core Loading Pattern 15 Control Rod Insertion Limits 16
R. E. Glnna, Cycle 19
1.0 INTRODUCTION
AND
SUMMARY
1.1 INTRODUCTION
This report presents an evaluation for R. E. Ginna Cycle 19, which demonstrates that the core reload will not adversely affect the safety of the plant. The Cycle 19 evaluation was accomplished utilizing the methodology described in WCAP-9273-A, "Westinghouse Reload Safety Evaluation Methodology."(")
R. E. Ginna is operating in Cycle 18 with a mixed fuel core comprised of Westinghouse Optimized Fuel Assemblies (OFAs) and Exxon Nuclear Corporation (ENC) fuel assemblies.
It is planned, for Cycle 19, to refuel the R. E. Ginna core with fresh Westinghouse 14x14 OFAs, and Westinghouse and ENC irradiated assemblies.
In a licensing submittal'(
) to the NRC, approval was requested for the transition from ENC fuel to Westinghouse OFA and associated proposed changes to the R.
E. Ginna Technical Specifications.
This submittal justified the compatibility of Westinghouse OFAs with ENC fuel assemblies in a mixed-fuel core, and contained bounding reference analyses which are applicable to the Cycle 19 safety evaluation.
NRC approval was received via Reference 3.
'ln a licensing submittal to the NRC(4), approval was requested for an increase in the maximum allowable steam generator tube plugging level.
The changes to the R. E.
Ginna Technical Specifications required as a result of the increased tube plugging were also included with this submittal'( ). The evaluation presented in this report is based on the analysis and results described in the steam generator 15% tube plugging license submittal'(
) as well as the OFA licensing submittal(
).
All of the accidents comprising the licensing bases(
) which could potentially be affected by the fuel reload have been reviewed for the Cycle 19 design described herein.
Justifications for the applicability of previous safety analyses are provided.
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R. E. Glnna, Cycle 19 1.2 GENERAL DESCRIPTION The R. E. Ginna Cycle 19 reactor core is comprised of 121 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1.
During the Cycle 18/19 refueling, the feed regions (21A, and 21 B) will consist of 28 OFAs, Of the twelve Region 21A assemblies, four contain 32 part-length IFBAs per assembly and eight contain 40 part-length IFBAs per assembly.
The Region 21B assemblies contain 24 fuel rods per assembly with part-length IFBAs. Eight ENC Region 10.and four ENC Region 11 fuel assemblies from the spent fuel pit will be inserted into the Cycle 19 reactor core.
A summary of the Cycle 19 fuel and IFBAinventory is given in Table 1.
Consistent with the use of the Westinghouse Improved Thermal Design Procedure (ITDP)( ) for the analyses(
- 4) of both Westinghouse and ENC fuel, the core design parameters utilized for Cycle 19 are as follows:
Core Power (MWt
~
System Pressure psia)
Vessel Average Temperature ('F)
Minimum Measured Flow (gpm).
Average Linear Power Density (kw/ft)
(based on average active fuel stack length of 141.4 inches)
1.3 CONCLUSION
1520 2250 573.5 89,600/Loop*
5.80 From the evaluation presented in this report, it is concluded that the Cycle 19 design does not result in the safety limits for any incident being exceeded.
This conclusion is based on the following:
1.
Cycle 18 bumup of 12,000+300/-500 MWD/MTU.
An evaluation(
) has been performed which shows that all safety limits are satisfied for 15% steam geneator tube plugging with a corresponding 86,900 gpm/loop minimum measured flow and a 85,100 gpm/loop thermal design flow.
1 3702F:6/890202
R. E. Glnna, Cycle 19 January 4989 2.
Cycle 19 burnup will not exceed 10800 MWD/MTUwhich includes an allowance for a 500 MWD/MTU EOC power coastdown.
During a power coastdown the following applies:
(a)
Coastdown by the normal programed power reduction, (b)
The low-lowTavg setpoint is not changed, and (c)
The Tavg program for rod control is not changed.
3.
There is adherence to the following:
(a)
(b)
(c) plant operating limitations as given in the Technical Specifications, the proposed changes in the 15% steam generator tube plugging licensing amendment(
), and the Bank D RIL conditions at HFP given in Section 2.2 of this report.
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R. E. Glnna, Cycle 19 January 1989 2.0 REACTOR DESIGN 2.1 MECHANICALDESIGN The Cycle 19 feed fuel assemblies consist of Region 21 W OFAs. Thirteen ENC fuel assemblies are inserted into the Cycle 19 core, one demonstration assembly from the Cycle 18 core and 12 assemblies from the spent fuel pit. The mechanical description and justification of the compatibility of the W OFAs with ENC fuel assemblies in a transition core were presented in the licensing submittal( ).
The mechanical design of the Region 21 fuel assemblies is the same as the Region 20 fuel assemblies.
Ten Enhanced Performance Rod Cluster Control Assemblies (EP-RCCAs) will be inserted into the Cycle 19 core.
They are of the same design as the ten EP-RCCAs inserted into the Cycle 18 core.
These RCCAs have a thin chrome electroplate applied to a specified length of absorber rodiet cladding in contact with the reactor internal guides to provide increased resistance to cladding wear.
In addition, the absorber diameter is reduced slightly at the lower extremity of the rodlets in order to accommodate absorber swelling and minimize cladding interaction with the absorber.
Table 1
presents a comparison of pertinent design parameters of the various fuel regions.
The Westinghouse fuel has been designed utilizing the Westinghouse improved fuel performance models(
). The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad. flattening model
( ).
For all Westinghouse fuel regions, the fuel rod internal pressure design
- basis, which is discussed and shown acceptable in Reference 10, is satisfied.
Westinghouse has had considerable experience with Zircaloy clad fuel, as described in Reference 11.
This report also describes the ope'rating experience that has been obtained from OFAs and IFBAfuel rods.
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R. E. Glnna, Cycle 19 2.2 NUCLEAR DESIGN The nuclear design of the Cycle 19 transition core used the standard calculational methods described in the Westinghouse Reload Safety Evaluation Methodology.(")
Although the physics characteristics are slightly different for the OFA when compared to the ENC fuel assembly, evaluations show that the differences are within the normal changes seen from cycle to cycle.
Table 2 provides a comparison of the Cycle 19 kinetics characteristics with the evaluation limits based on the accident analyses submitted to the NRC.(
)
It can be seen from the Table 2 parameters that all of the Cycle 19 values fall within the evaluation limits. These parameters are evaluated in Section 3.0.
Table 3 provides the end-of-life control rod worth and requirements at the most limiting condition during the cycle.
The available shutdown margin exceeds the minimum required.
If Cycle 18 operates to the high end of its burnup window (12,300 MWD/MTU), the maximum calculated HFP F~H, with Bank D at the HFP insertion limit (20% inserted), is 0.33%
(including 8% design allowance, Reference 1 2) above the 1.66 Technical Specification limit. The violation is expected to be in the Cycle 19 burnup range from 5250 to 9500 MWD/MTU. There is no violation if the Bank D insertion is not greater than 17.75% or the Cycle 18 burnup is 12,150 MWD/MTUor less.
To preclude a potential violation of the F~H limit, Bank D insertion should be limited to 17.75% (5 steps above the current RIL) at full power as shown in Figure 2. This limit needs only to be applied for the Cycle 19 burnup range from 5250 to 9500 MWD/MTUif Cycle 18 operates to a burnup greater than 12,150 MWD/MTU. This temporary change in the insertion limit poses no adverse impact on other core safety parameters or technical specifications.
2.3 THERMALAND HYDRAULICDESIGN No significant variation in the thermal margins will result from the Cycle 19 reload.
The
. DNB core limits and safety analyses used for Cycle 19 are based on the conditions in Section 1.2.
Sufficient DNB margin exists to satisfy the design criteria for the Cycle 19 reload.
3702F:6/890202
R. F Glnna, Cycle19 January 1989 3.0 POWER CAPABILITYANDACCIDENTEVALUATION 3.1 POWER CAPABILITY The plant power capability for Cycle 19 is evaluated considering the consequences of those UFSAR incidents on the licensing basis accident analysis'(
). It is concluded that the core reload will not adversely affect the ability to safely operate at the current 1520 MWt rated power during Cycle 19.
For overpower transients, the fuel centerline temperature limit of 4700'F can be accommodated with margin in the Cycle 19 core. The revised Fuel Thermal Safety
- Model, incorporated into the improved fuel models
'( ), was used for fuel temperature evaluations.
The LOCA limit at 1520 MWt for Westinghouse fuel is met by maintaining [FQ(z)xp] at or below [2.32* xK(Z)t. This limit is satisfied by the power control maneuvers allowed by the Technical Specifications, which assure that the Final Acceptance Criteria (FAC) limits are met for a spectrum of small and large break LOCAs.
3.2 ACCIDENTEVALUATION The effects of the reload, including the mechanical design features described in Section 2.1, on the design basis and postulated incidents analyzed in the UFSAR were examined.
In all cases it was found that the effects can be accommodated within the conservatism af the initial assumptions used in the applicable safety analysis.
A core reload can affect accident analysis input parameters in the following areas:
core kinetic characteristics, control rod worths, and core peaking factors.
Cycle 19 parameters in each of these areas were examined as discussed below to ascertain whether new accident analyses were required.
3.2.1 Kinetics Parameters A corn'parison of Cycle 19 core physics parameters with current evaluation limits is given in Table 2. Allthe kinetics values remain within the bounds of the analysis limits.
The maximum FQ value allows a maximum of 15 percent steam generator tube plugging.
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R. E. Glnna, Cycle 19 January 1989 3.2.2 Control Rod Worths Changes in control rod worths may affect differential rod worths, shutdown
- margin, ejected worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 19 is less than or equal to the analysis limit. Table 3 shows that the Cycle 19 shutdown margin requirements are satisfied.
Cycle 19 ejected rod worths are within the bounds of the analysis limits.
Cycle 19 has a normalized trip reactivity insertion rate which is slightly different from the current limit. The effects of this reduced normalized trip reactivity rate have been evaluated for those accidents affected and compared to previous analyses.
The only significant non-conservative deviations, with respect to the current limit between the two trip insertion curves, occur for the last 25 percent of rod insertion.
The remaining portion of the trip insertion curve is conservative with respect to the current limit.
Slow transients are relatively insensitive to the trip reactivity insertion rate.
Fast transients are evaluated to confirm that the limiting transient conditions are unchanged.
Results show that the previous analyses remain applicable.
3.2.3 Core Peaking Factors Evaluation of peaking factors for the rod out of position, dropped RCCA, and dropped bank incidents show. that the DNBR is maintained above the appropriate safety analysis limit DNBR value listed in Table 4 for each fuel type.
The hypothetical steamline break transients were evaluated for Cycle 19.
This evaluation showed that the predicted results are within the bounds of the submitted analysis.@
3702F:6/890202
R. E. Glnna, Cycle 19
4.0 REFERENCES
1.
- Bordelon, F. M., et al., "Westinghouse Reload Safety Evaluation Methodology,"
WCAP-9273-A, July 1985.
2.
Letter from J. E. Maier (RGE) to H. R. Denton (NRC), December 20, 1983,
Subject:
"R. E. Ginna Application for Reload License Amendment Using Westinghouse Optimized Fuel Assemblies."
3.
NRC Safety Evaluation Report," Use of Westinghouse Optimized Fuel Assembly (OFA) as reload fuel for R. E. Ginna", Docket Number 50-244, May 1, 1984.
4.
R. W. Kober (RGE) to C. Stahl (NRC), October 27, 1987,
Subject:
R.
E.
Ginna Application for Licensing Amendment for 15%
Steam Generator Tube Plugging Level.
5.
Updated Final Safety Analysis Report - R. E. Ginna, Docket Number 50-244.
6.
- Chelemer, H., et.
al., "Improved Thermal Design Procedure,"
WCAP-8567, July 1975.
7.
Miller, J. V. (Ed), "Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations," WCAP-8785, October 1976.
8.
- Weiner, R. A., et. al., "Improved Fuel Performance Models for Rod design and Safety Evaluations," WCAP-1 1873-A, August 1988.
9.
George, R.A., et al., "Revised Clad Flattening Model," WCAP 8381, July 1974.
10.
Risher, D.H., et al., "Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP 8964-A, August 1978.
8702F:6/890202
R. E. Glnna, Cycle 19 January 1989 11.
- Foley, J.
and
- Skaritka, J.,
"Operational Experience with Westinghouse Cores (through December 31, 1987)," WCAP-8183, Revision 16, August 1988.
- 12. Spier, E. M., et. al., "Evaluation of Nuclear Hot Channel Factor Uncertainties,"
Section 7.0, WCAP-7308-L-P-A, June 1988.
13.
Motley, F. E., et al., "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976.
3702F:6/890202
R. E. Glnna, Cycle 19 J
ry 1989 TABLE 1 R. E. GINNACYCLE 19 FUEL PARAMETERS
~Re drool Enrichment (w/o)
Density (% theoretical)
No. of Assemblies Region Loading (MTU)
LTA(1)
QD 3.70 94.0 0.325 8
4 8
16 16 2.970 1.485 2.843 5.640 5.620 1.400
~0
~M, QI~U
~9PgVV
~19B V 19C(VV 3 pg 3 21 3 3gg(2) 3.4p1(2) 3.611(2) 3 4p1(2) 94.0 94.0 95.36 95.09 94.94 95.09 Approximate Bumup at Beginning of Cycle 19 (GWD/MTU)
Length of Natural U Axial Blankets (3)
Top (inches)
Bottom (inches)
Number of IFBA (4)
Fuel Rods 43 29 31 32 6.2 6.2 27 6.2 6.2 640(5) 6.2 6.2 24 6.2 6.2 (1
Exxon Annular demo 2
Enrichment in the enriched region 3
Only W fuel contains axial blankets Exxon fuel has no axial blanket (4)
IFBA-Integral Fuel Burnable Absorber Thin boride coating on surface of fuel pellets 5) 92 inches IFBA axial length - centered 6) 104 inches IFBA axial length - centered 3702F:0/890202 10
R. E. Glnna, Cycle 19 Ja 1989 TABLE1 (Cont'd)
R. E. GINNACYCLE l9 FUEL PARAMETERS
~eceion Enrichment (w/o)
Density (% theoretical)
No. of Assemblies Region Loading (MTU)
Approximate Bumup at Beginning of Cycle 19 (GWD/MTU)
Length of Natural U Axial'Blankets (3) 95.37 95.37 95.37 95.00 95.00 95.00 12 8
16 4
8 16 4.230 2.820 5.620 1.410 2.820 5.640 16 RQ~~
2OOl2~W
,'>~BLWW.
~~V 2~UJ
~21 Y 3 4Q4(2) 3 4Q4(2) 4 QQ8(2) 3 6Q(2) 3 6Q(2) 4 0(2)
Top (inches)
Bottom (inches)
Number of IFBA '(4)
Fuel Rods 6.2 6.2 6.2 6.2 192(
)
192(5) 6.2 6.2 6.2 6.2 6.2 6.2 6.2 6.2 128(
)
320(
)
384'(
)
1)
Exxon Annular demo 2)
Enrichment in the enriched region 3
Only W fuel contains axial blankets Exxon fuel has no axial blanket (4)
IFBA-Integral Fuel Burnable Absorber Thin boride coating on surface of fuel pellets
~
~
~
5) 92 inches IFBA axial length - centered 6) 104 inches IFBA axial length - centered 3702F:6/890202
R. E. Glnna, Cycle 19 TABLE2 KINETICS CHARACTERISTICS R. E. GINNA-CYCLE 19 Reference Analysis V ues 2
~CgLe~9""
Moderator Temperature Coefficient, (PCM/'F)"
+5 to -42.9
+5 to -42.9 Doppler Coefficient (PCM/'F)*
-2.9 to -0.91
-2.9 to -0.91 Delayed Neutron Fraction jef
.0043 to.0073
.0043 to.0073 Maximum Prompt Neutron Lifetime (N, sec) 26
<26 Maximum Differential Rod Worth of Two Banks Moving Together at HZP (PCM/sec)*
97.5
<97.5 1 PCM = 1.0 x 10 5 hp Actual values fall within the bounds indicated 3702F:6/890202 12
~
R. E. Glnna, Cycle 19 January 1989 I
~
TABLE3 SHUTDOWN REQUIREMENTS AND MARGINS R. E. GINNA<<CYCLE 'I9 ogle 19 EOC o
r Rod Wo
'e ce
)
All Rods Inserted Less Worst Stuck Rod 2 loops in operation 5.96 (A) Less 10%
5.36 n
RdR i
Reactivity Defects (Doppler, Tavg Void, Redistribution) 2,52 Rod Insertion Allowance 0.64*
(B) Total Requirements 3.16 wn Mar ercen h
-B 2.20 ge
'edShu o
e cen 1.80
- Based on current Bank D RIL of 20% at full power.
3702F:6/890202
R. E. Glnna, Cycle 19 I,I~
~i TABLE4 THERMAL-HYDRAULICDESIGN BASES'OR THE RGE CYCLE 19 SAFETY EVALUATION a.
LANTP
~parame r
ITDP Value Non-ITDP Value Reactor Power, MWt 1520 1520 2 2%
Primary Flow, gpm 179,200 (173,800)*
174,000 (170,200)*
- Pressure, psia 2280 2250 g 30 psia Tin 'F 543;7 543.7 2 4'F Radial Peaking Factor (F~H')
1.60[1+0.3(1-P)]
1.66[1+0.3(1-P)]
- b. FUEL RELATED P Paarame er DNB Correlation 14 14 Exx n Fue W-3 14 14 WOF WRB-1 (13)
Safety Analysis Design Limit 1.62 Typical Cell 1.52 Typical Cell 1.54 Thimble Cell 1.51 Thimble Cell Thermal Design Procedure ITDP ITDP An evaluation(
) has been performed for 15% steam generator tube plugging: This flow reduction was offset by using available plant DNBR margin.
All DNB safety limits were met.
3702F:6/890202 14
. R. E. Glnna, Cycle 19 r
January1989 F'
FIGURE 'I R. E. GINNACYCLE 19 CORE LOAOING PATTERN 180 M
L K
EI I
H G
F E
D C
8 A
198 V30 19C V20 218 Y27 198 V27 218 Y20 198 V36 19A V06 208 W35 208 W32 10 L16 218 Y15 20A W19 19A V05 11 M23 21A Y01 198 V33 21A Y06 10 L29 218 Y18 20A W13 19A V10 19A V02 208 W22 208 W31 198 V25 218 Y26 198 V26 19C V18 218 Y14 198 V29 19A vO4 208 W28 208 W33 20A WOB 20A W09
'1 8 U07 20A 20A W04 W07 208 W21 208 W34 19A V12 10 L30 218 Y28 20A W16 19A V 13 20A W11 18 U29 21A YOS 18 U25 20A W01 19A V07 20A
'W14 218 Y16 10 L15 90 7
11 M24 21A Y02 198 V24 21A 18
- Y09, U28 21A Y10 178 XT03 21A Y11 18 U03 21A Y05 198 V31 21A Y03 11 M22 270 10 12 13 10 L13 218 Y23 19A V16 198 V22 20A W15 208 W25 218 Y21 19C V19 19A VOB 208 W26 198 V21 218 Y17 198 V34 20A W12 20A W02 208 W24 208 W29 19A VO9 18 U13 20A W06 19A V 15 20A W17 218 Y13 10 L31 21A Y12 18 U01 21A Y07 198 V23 21A Y04 11 M21 18 U12 20A
'W05 19A V14 20A W18 218 Y25 10 L14 20A W03 20A W10 208 W36 208 W30 19A V01 19A V03 208 W23 198 V32 218 Y19 198 V28 20A W20 208 W27 218 Y22 19C V17 218 Y24 19A V 11 198 V35 10 L32 0 DEG.
KEY:
R R
~
REGION NUMBER ID IC
~
FUEL ESSEMBLY IDENTIFICE IDN 3702F:6/890202 15
FIGURE 2
R.
E.
GINNA CYCLE l9 CONTROL RO ION LIM O
INSERT 19,10 69,10 ANK d,d I
(100,S~
80 oo,so)
BA K
C I 60 BAN D
)I Li BANK It D
Te pora Ch (o,z
)
20 30,0 30, 0
10 20 X
40 50 60 70 80 PERCEHT OF FUIL POWER 90 100 TS AS A
FUNC,TION OF POWER