ML13324A854
| ML13324A854 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/31/1986 |
| From: | Dzenis E, Gergos B Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13324A853 | List: |
| References | |
| NUDOCS 8605080348 | |
| Download: ML13324A854 (16) | |
Text
RELOAD SAFETY EVALUATION SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1 CYCLE 9 MARCH 1986 Edited by B. W. Gergos Approved:
E. A. Dzenis, Manac Core Operations 8605080348 860506 PDR ADOCK 050020 P L-3 4091 L:6-860303
TABLE OF CONTENTS Title Page
1.0 INTRODUCTION
1 2.0 REACTOR DESIGN 2
2.1 Mechanical Design 2
2.2 Nuclear Design 2
2.3 Thermal and Hydraulic Design 3
3.0 ACCIDENT EVALUATION 4
3.1 Power Capability 4
3.2 Accident Evaluation 4
3.3 Incidents Reanalyzed 5
4.0 TECHNICAL SPECIFICATIONS 6
5.0 REFERENCES
7 4091 L:6-860310
LIST OF TABLES Table Title Page 1
Fuel Assembly Design Parameters 9
2 Core Physics Parameters 10 3
Shutdown Requirements and Margins 11 LIST OF FIGURES Figure Title Page 1
Core Loading Pattern 12 2
F Total Versus Axial Offset 13 0
4091 L:6-860303 11
1.0 INTRODUCTION
AND
SUMMARY
The San Onofre Nuclear Generating Station Unit 1 is in its eighth cycle of operation. The unit will refuel and be ready for Cycle 9 startup in the early part of 1986.
This report presents an evaluation for Cycle 9 operation which demonstrates that the core reload will not adversely affect the safety of the plant. Those incidents analyzed and reported in the FSA(1) and other incidents subsequently analysed(2) which could potentially be affected by fuel reload have been reviewed for Cycle 9 design described herein. The results of new analyses have been included, and the justification for the applicability of previous results from the remaining analyses is presented. These analyses assume that: 1) Cycle 8 operation is terminated between 9100 and 10100 MWD/MTU, 2) Cycle 9 burnup is limited to the end-of-full power capability*,
and 3) there is adherence to plant operating limitations given in the technical specifications.
The San Onofre 1, Cycle 9 core loading pattern is shown in Figure 1. The one Region 7 and 51 Region 8 fuel assemblies from Cycle 8 will be removed and replaced by 52 Region 11 fuel assemblies. A Region 8 fuel assembly will be reused in the central core position.
Nominal design parameters for Cycle 9 are 1347 Mwt core power, 2100 psia system pressure, and 4.64 kw/ft average linear fuel power density. This Reload Safety Evaluation (RSE) covers Cycle 8 RSE( 3) conditions for revision 1 (nominal core inlet temperature 551.5 0F and a 195,000 gpm RCS thermal design flow which accounts for up to an equivalent 20 percent steam generator tube plugging) and for revision 2 (nominal core inlet temperature 528 0 F and a 201,900 gpm RCS thermal design flow which accounts for up to an equivalent 15 percent steam generator tube plugging).
- Definition:
Full rated power and nominal core inlet temperature control rods fully withdrawn, and zero ppm of residual boron.
4091 L:6-860303 1
2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 11 fuel assemblies is the same as the Region 10 assemblies. Table 1 compares pertinent design parameters of the various fuel regions. The Region 11 fuel has been designed according to the fuel performance model in Reference 4. For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in Reference 5, is satisfied.
Clad flattening will not occur during Cycle 9. All fuel regions have a predicted clad flattening time equal to or greater than 50,000 EFPH. No fuel region will receive this exposure.
2.2 NUCLEAR DESIGN Cycle 9 core loading satisfies an ECCS analysis limit of FT x P of on FT of 2.89()
as shown in Figure 2. The limitations on F of 2.89 include the effects of the local power peaking of Figure 3.1 in WCAP 8131(6) to assure that the allowable value for LOCA is satisfied. The points plotted on Figure 2 include maneuvers typically done at San Onofre Unit 1 and variants on these maneuvers done at a number of control rod insertions, times and burnups.
The limiting F has been determined for the combination of the most adverse Q
N FXY and the most adverse Fz that will be experienced during operation in Cycle 9. The most adverse FXY occurs at beginning of life Tand the most adverse FN occurs at end of life. The results shown for FQ in Figure 2 include uncertainty factors of 15% for conservatism and 4% for manufacturing tolerances.
The xenon transient analysis has been evaluated similarly to analyses of N
previous cycles. The most limiting FN, including an uncertainty of 10%
on FN is 1.78. This occurs between 81 and 88% of core height. This is conservative compared to the FN = 1.92 at 85% core elevation required for the Cycle 9 design.
4091 L:6-860303 2
Table 2 provides a comparison of Cycle 9 kinetics characteristics with the current limit based on previously submitted accident analysis. The effect of the Table 2 parameters are evaluated in Section 3. Table 3 provides the end-of-life control rod worths and requirements at the most limiting condition during the cycle. The required shutdown margin is based on a previously submitted accident analysis.(7) The available shutdown margin exceeds the minimum required to meet the accident analysis.
2.3 THERMAL AND HYDRAULIC DESIGN The Cycle 9 safety analysis was based on the use of a design axial power shape of FN = 1.95 at 85 percent core elevation and a FN = 1.57. A total of 4.4 Z
(3) percent DNBR margin is available if a limiting FN = 1.92 at 85 percent core elevation is assumed. For the current cycle even more margin is available since the limiting FN is actually 1.78 (Section 2.2).
The 4.4 percent DNBR margin is z
~(3)N=N comprised of two parts The difference between FN 1.95 and FN 1.92 provides a 1.1 percent DNBR margin.
Pitch reduction gives the other 3.3 percent DNBR margin. This 4.4 percent DNBR margin was used to accomodate the reduction in DNBR margin resulting from the use of a FN = 1.57 rather than a FN = 1.55.
AH With the use of the above peaking factors no significant variations in thermal margins will result from the Cycle 9 reload. The present DNB core limits are conservative for Cycle 9.
4091 L:6-860303 3
3.0 ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability is evaluated considering the consequences of those incidents examined in the FSA(1) and other incidents subsequently analyzed(2,_using the previously accepted design basis. It is concluded that the core reload will not adversely affect the ability to safely operate at 100% of rated power during Cycle 9. For Condition II overpower transients, the fuel centerline temperature limit of 4700*F can be accommodated with margin in the Cycle 9 core. The time dependent densification model (8) was used for fuel temperature evaluations.
The LOCA limit at rated power can be met by maintaining F at or below 2.89.
This limit is satisfied by the power control maneuvers allowed by the technical specifications, which assure that the Interim Acceptance Criteria (IAC) limits are met for a spectrum of small and large breaks.
3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSA(1) and other incidents subsequently analyzed (2) were examined. In most cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis. For those incidents which were reanalyzed, it was determined that the applicable design bases are not exceeded, and, therefore, the conclusions presented in the FSA are still valid.
A core reload can typically affect accident input parameters in the following areas:
core kinetic characteristics, control rod worths, and core peaking factors. Cycle 9 parameters in each of these three areas were examined as discussed below to ascertain whether new accident analyses were required.
A comparison of Cycle 9 core physics parameters with current limits is given in Table 2. The kinetic values fall within the bounds of the current limits.
4091 L:6-860303 4
Changes in control rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Tables 2 and 3 show that the maximum reactivity withdrawal rate, and the shutdown margin with the worst stuck RCCA are within the current limits.
The ejected rod worths and trip reactivity curve are within the bounds of the current limits.
Peaking factor evaluations were performed for the rod out of position, dropped RCCA bank, dropped RCCA, and hypothetical steamline break accident to ensure that the minimum DNB ratio remains above 1.30. These evaluations were performed utilizing Cycle 9 transient statepoint information and peaking factors. In each case, it was found that the peaking factor for Cycle 9 was lower than the value for which DNBR equals 1.30. Consequently, no further investigation or analysis was required. The peaking factors following control rod ejection are within the limits of previous analysis for both BOL and-EOL zero power and full power cases.
3.3 INCIDENTS REANALYZED 3.3.1 LOCKED ROTOR The Locked Rotor event was analyzed for both the 20% and 15% SG tube plugging RCS flow and temperature conditions to verify their acceptability for the cycle 9 reload. For both sets of conditions, the peak RCS pressure and peak fuel clad temperature calculated for the current reload design fall within the bounds of the acceptable safety limit criteria. Calculations indicate that, for the 20% tube plugging conditions, 2% of the fuel rods in the core undergo DNB.
However, because the peak fuel clad temperature criterion has been satisfied, the cycle 9 reload core will remain in place and intact with no loss of core cooling capability. Thus, the applicable safety criteria related to the locked rotor event have been met.
4091 L:6-860303 5
4.0 TECHNICAL SPECIFICATIONS The results of this safety evaluation indicate that no changes to the San Onofre Unit 1 Technical Specifications are required for Cycle 9 operation.
4091 L:6-860303 6
5.0 REFERENCES
- 1. Docket Number 50-206, "San Onofre Nuclear Generating Station, Unit 1, Part 2, Final Safety Analysis."
- 2. "Automatic Initiation of Auxiliary Feedwater System, San Onofre Nuclear Generating Station Unit 1," Attachment to letter, K. P. Baskin to D. M.
Crutchfield, March 6, 1981; "Automatic Initiation of Auxiliary Feedwater System, San Onofre Nuclear Generating Station Unit 1 Amendment 97,"
Attachment to letter, K. P. Baskin to C. M. Crutchfield, November 18, 1981 and "Reactor Coolant Pump Rotor Seizure/Reactor Coolant Pump Shaft Break Analysis San Onofre Unit 1" letter from R. W. Kreiger SCE to D. M.
Crutchfield, July 27, 1983, SEP Topic XV-7.
- 3. Skaritka, J., Editor, "Reload Safety Evaluation - San Onofre Unit 1, Cycle 8," Revision 1, October 1980 and Revision 2, April 1981.
- 4. Miller, J. V. (Ed.), "Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations," WCAP-8785, October 1976.
- 5. Risher, D. H. et. al., "Safety Analysis for The Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.
- 6. "Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station Unit 1, Cycle 4," WCAP-8131, May 1973.
- 7. SCE Report, "Steamline Break Accident Reanalysis, San Onofre Nuclear Generating Station Unit 1, October 1976," Attachment to letter, K. P.
Baskin to K. R. Goller, December 29, 1976, and Trojan (sic) SCE Cycle 8 Steamline Break Reanalysis," Attachment to letter SCE-81-554, R. L. Kelly to H. B. Ray, November 24, 1981.
- 8. Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation," WCAP-8218-P-A, March 1975 (Proprietary) and WCAP-8219-A, March 1975 (Non-Proprietary).
4091 L:6-860303 7
- 9. "Description and Safety Analysis Including Fuel Densification San Onofre Nuclear Generating System, Unit 1, Cycle 5, Attachment to letter from Jack B. Moore to Edson G. Case, March 7, 1975.
4091 L:6-860303 8
0I TABLE 1 SAN ONOFRE UNIT 1 -
CYCLE 9 Fuel Assembly Design Parameters Region 8
9 10 11 Enrichment (w/o U-235), Nominal 4.00 4.00 4.00 4.00 Density (% Theoretical)*
94.59 94.66 94.45 94.68 Number of Assemblies 1
52 52 52 Approximate Burnup at 1020 900 400 0
Beginning of Cycle 9 (EFPD)
- All regions are as-built values.
4091 L:6-860303 9
TABLE 2 San Onofre Unit 1, Cycle 9 CORE PHYSICS PARAMETERS Current Limit Cycle 9 Moderator Temperature
-40.0 to 0(1)
>-40 to 0 Coefficient (pcm/oF)
Doppler Coefficient,
-2.75 to -1.4(3)
>-2.75 to <-1.4 (pcm/oF)
Delayed Neutron Fraction, 0.50 to 0.70(1)
>0.50 to <0.70 eff' Maximum Prompt Neutron 26
<26 Lifetime (P sec)
Maximum Reactivity Withdrawal 40(3)
<40 Rate, (pcm/sec)*
- pcm = 10-5 Ap 4091 L:6-860303 10
TABLE 3 SAN ONOFRE UNIT 1 -
CYCLE 9 SHUTDOWN REQUIREMENTS AND MARGINS Control Rod Worth (PCM)
BOC EOC All Rods Inserted 6837 7458 All Rods Inserted Less Worst Stuck Rod 6052 6571 (1) Less 10%
5445 5914 Control Rod Requirements (PCM)
Reactivity Defects (Doppler, Tavg, 1628 2528 Void, Redistribution)
Rod Insertion Allowance 1488 1255 (2) Total Requirements 3116 3783 Shutdown Margin [(1)-(2)] (PCM) 2330 2131 Required Shutdown Margin (PCM) 1250 1900(7) 4091 L:6-860303 11
Figure 1 CORE LOADING PATTERN SAN ONOFRE UNIT 1 - CYCLE 9 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 9
11 9
A 111 1
i11 10 11 11 11B C
11 11 10 9
9 9
10 11 11 11 11 9
10 9 11 9
10 9
11
- 11.
D 11 11 9
10 10 9
9 9
10 10 9
11 11 E
11 10 10 10 10 10 9
10 10 10 10 10 11 F
9 11 9
9 9
10 10 10 10 10 9
9 9
11 9
G 11 10 9
11 9
9 10 8 10 9
9 11 9
10 11 H
9 11 9
9 10 10 10 10 10 9
9 9
11 9
11 10 10 10 10 10 9
10 10 10 10 10 11 K
11 11 9
10 10 9
9 9
10 10 911 11 L
11 11 10 9 11 9
10 9
11 11 M
11 11 10 9
9 9
10 11 11 N
1 1 11 10 11 11 1
9 11 9
R S
Source Location
- Fuel Region Number 12
NORMAL OPERATION 5.--
4.5
- 4.
0 Axial Offset Limits Z
Cycle 9 Design Limit CE of 2.89 n.I.'
-j.
___)
C__
rM0 4*~~
-h C
40-.
0.(A0.
.C
- 2.
+c+
4&
CO
- 4.
0
-ht (D
-70
-60
-0
-40
-30
-20
-10 0
10 20 50 40 50 PERCENT FLUX IMBALANCE