ML20076J916: Difference between revisions
StriderTol (talk | contribs) (StriderTol Bot change) |
StriderTol (talk | contribs) (StriderTol Bot change) |
||
| Line 213: | Line 213: | ||
T Safety Analysis A per=anent hydrogen igniter system has been shown by experiment to be capable of mitigating the effects associated with hydrogen production ! | T Safety Analysis A per=anent hydrogen igniter system has been shown by experiment to be capable of mitigating the effects associated with hydrogen production ! | ||
during degraded core accidents. | during degraded core accidents. | ||
5.0 DROPPED ROD ACCIDENT ANALYS'IS - REMOVAL OF OPERATING RESTRICTIONS Our July 22, 1982 letter to Ms. E. Adensam (see appendix D) formally requested your staff review the material submitted to you as NS-EPR-3545, January 20, 1982, and subsequently remove the interim operational restrictions before startup of Sequoyah unit 2, cycle 2. | 5.0 DROPPED ROD ACCIDENT ANALYS'IS - REMOVAL OF OPERATING RESTRICTIONS Our {{letter dated|date=July 22, 1982|text=July 22, 1982 letter}} to Ms. E. Adensam (see appendix D) formally requested your staff review the material submitted to you as NS-EPR-3545, January 20, 1982, and subsequently remove the interim operational restrictions before startup of Sequoyah unit 2, cycle 2. | ||
We believe the removal of these operating restrictions is justified and request your concurrence. | We believe the removal of these operating restrictions is justified and request your concurrence. | ||
Latest revision as of 11:41, 27 September 2022
| ML20076J916 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/01/1983 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20076J913 | List: |
| References | |
| NUDOCS 8307070005 | |
| Download: ML20076J916 (25) | |
Text
{{#Wiki_filter:- .
'}
ENCLOSURE 1 RELOAD SAFETY EVALUATION SEQUOYAH UNIT 2. CYCLE 2 8307070005 830701 PDR ADOCK 05000328 P PDR
TABLE OF CONTENTS Page
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 REACTOR DESIGN 2 2.1 Mechanical Design 2 2.2 Nuclear Design 2 2.3 Thermal and Hydraulic Design 4 3.0 ACCIDENT EVALUATION 6 3.1 Power Capability 6 3.2 Accident Evaluation 6 3.3 Incidents Reanalyzed 8 4.0 TECHNICAL SPECIFICATION CHANGES 10 5.0 DROPPED ROD ACCIDENT ANALYSIS - REMOVAL OF 15 OPERATING RESTRICTIONS
6.0 REFERENCES
15 Appendix A Technical Specification Changes Appendix B Peaking Factor Limit Report l Appendix C Reload Test Program l Appendix D Removal of Rod Control Operating Restrictions i l
I y
/
,-+
/ r g
LIST OF TABLES Title Pace Table 1 Fuel Assembly Design Parameters 16 2 Kinetic Characteristics 17 3 Shutdown Requirements and Margins l8 4 Rod Ejection Parameters l9 LIST OF FIGURES Figure Title Page 1 Core Loading Pattern 2C) 2 Maximum Calculated Values of Fa "H with Respect to Technical Specification Limits 21
'3 Revised Technical Specification Figure 2.1-1. 22 l
ii
n / .
1.0 INTRODUCTION
AND
SUMMARY
Sequoyah Unit 2 is in its first cycle of operation. The unit is expected to refuel and be ready for Cycle 2 startup in November 1983. This report presents an evaluation for Cycle 2 operation which demonstrates that the core reload will not adversely affect the safety of the plant. Those incidents analyzed and reported in tne r5AR(1) which could potentially be affected by fuel reload have been reviewed for the Cycle 2 design described herein. The applicability of the current nuclear design limits was verified for Cycle 2 using the methods described in Reference 2. The results of new analyses have been included, and the justification for the applicability of previous results from the remaining analyses is presented. It has been concluded that the Cycle 2 design does not cause the previously acceptable safety limits for any incident to be exceeded. The above operational conclusions are based on the assumption that: (1) Cycle 1 nominal burnup is 15,100 + 500 MWD /MTV, (2) Cycle 2 burnup is limited to 11,500 MWD /MTV, and (3) there is adherence to plant operating limitations given in the technical specifications and their proposed modifications presented herein. The impact of this reload has been evaluated and found to be within the safety limit as described in this RSE for Cycle 2. During the Cycle 1/2 refueling, sixty-eight Region 1 fuel assemblies will be replaced by sixty-eight Region 4 assemblies. See Table 1 for tne most signficant fuel assembly design parameters for each region and Figure 1 for the Cycle 2 core loading pattern. Nominal design parameters for Cycle 2 are 3411 MWt core power, 2250 psia core pressure, nominal core inlet temperature of 546.7 F, and core average linear power of 5.43 kw/ft. 1
N
, j ..
7 . . 2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 4 fuel assemblies is the same as the Region 3 assemblies.* The Region 4 fuel has been designed according to the fuel performance model in Reference 3. The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse model.b) For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in Reference 5, is satisfied. It is also planned to use Wet Annular Burnable Absorber (WABA) rods, which are described and evaluated in the WABA Evaluation Report.(12) Westinghouse has had considerable experience with Zircaloy clad fuel. This experience is extensively described in WCAp-8183, " Operational Experience with Westinghouse Cores."I ) This report is updated an-nually. 2.2 NUCLEAR DESIGN Cycle 2 core loading is designed to meet an Fg (z) x P ECCS analysis limit of 52.237 x K(z). Table 2 provides a comparison of the Cycle 2 kinetics characteristics with the current limit based on previously submitted accident analysis. With the exception of the most negative Doppler Temperature Coefficient, all of the Cycle 2 values fall within the current limits. These parameters are evaluated in Section 3. Table 3 prvvides the end of life control rod worths and requirements at the most limiting condition during the cycle. The required shutdown margin is based on previously submitted accident analysis. The avail-able shutdown margin exceeds the minimum required. The control rod insertion limits remain unchanged from Cycle 1, as given in the tech-nical specifications.
- With the exception of minor grid modifications to minimize potential grid to grid interaction during fuel handling and a reconstitutable bottom nozzle design.
2
)
j - _ y.. The PALADON CodeN) was used in the nuclear analyses. The NRC has found this code acceptable for use on reload designs. Twenty-eight Region 4 fuel assemblies will contain 288 new WABA rods arranged as shown in Figure 1. Two symmetrically located Region 3 fuel assemblies will contain secondary source rods that were irradiated in Cycle 1. N limit si pe was changed from 0.2 In the Cycle 2 analysis, the F AH to 0.3. The change in F N with power is described by the fol-AH lowing relationship: N F AH f 1.55 D + 0.3' (1-M This allows an increase in allowable F"AH at reduced power in com-parison to the previous Technical Specification limit while maintaining the same F AH limit at full power. The increase in allowable F AH N at reduced power allows for optimization of the core loading pattern for N full power operation by minimizing the restriction on F AH at low power. This eliminates the need to change the rod insertion limits to satisfy peaking f actor criteria at low power with the control rod banks at the N insertion limit. The variatian in the maximum calculated F AH with power with the control rods at the insertion limit for Cycle 2 is shown in Figure 2. Relaxed Axial Offset Control (RAOC) will be employed in Cycle 2 to en-hance operational flexibility. RAOC makes use of available margin by expanding the allowable Al band, particularly at reduced power. The RAOC methodology and application is fully descriFed in Reference 11. The analysis for Cycle 2 indicates that no change to the safety para-meters is required for RAOC operation. 3
f I_ /
,/,
[/ Adherence to the gF limit is obtained by using the F Surveillance g Technical Specification, also described in Reference 11. Fg surveil-lance replaces the previous F,y surveillance by comparing a measured Fg , increased to account for expected plant maneuvers, to the F g limit. This provides a more convenient form of assurin,' plant operation below the Fg limit while retaining the intent of using a measured parameter to verify operation below Technical Specification limits. Fg surveillance is only a change to the plant's surveillance require-ments and as such has no impact on the results of the Cycle 2 analysis or safety parameters. 2.3 THERMAL AND HYDRAULIC DESIGN No significant variations in thermal margins result from the Cycle 2 reload. However, the reactor core safety limits, Figure 2.1.1 in the i Technical Specifications, and the axial offset limits have been revised to reflect the increase in K from 0.2 to 0.3 in the following rela-tionship. FtH < 1.55 [1 + K (1-P)] Where P = fraction of rated power for power levels less than 100's. i. The core limits at 1775 and 2000 psia remain unchanged from the current limits. At 2250 and 2400 psia the proposed core limits are slightly more limiting below 100% power. The core limits have these minimal changes because at most conditions below full power, the restriction f that the average enthalphy at the vessel exit is less than the enthalpy of saturated liquid is more limiting than DNB considerations. This vessel exit enthalpy limit is not core peaking factor dependent. The change in axial offset limits are discussed in Section 3.3. 1 l j 4
f ,<
/e se The thermal-hydraulic methods used to analyze axial power distributions generated by the RAOC methodology are similar to those used in the Constant Axial Offset Control (CAOC) methodology. Normal operation power distributions are evaluated relative to the assumed limiting normal operation power distribution, which for Sequoyah Unit 2, Cycle 2, is the 1.55 cosine, used in the accident analysis. Limits on allowable operating axial flux imbalance as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA gF considerations.
The Condition 11 analyses were evaluated relative to the axial power distribution assumptions used to gen + ate DNB care limits and resultant Overtemperature Dc!ta-T setpoints (including the f( AI) function). No changes in these limits are required for RAOC operation. 5
/ . /.
A .. / , . 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability is evaluated considering the consequences of those incidents examined in the FSAR,0 ) using the previously accepted design basis. It is concluded that the core reload will not adversely affect the ability to safely operate at 100 percent of rated power dur-ing Cycle 2. For the evaluation performed to address overpower con-cerns, the fuel centerline temperature limit of 4700*F can be accom-modated with margin in the Cycle 2 core using the methodology described in Reference 2. The time dependent densification model( } was used for these fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining Fg at or below 2.237. 3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSAR for four loop operation have been examined. In most cases, it was found that the effects can be 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 basis limits are not exceeded, and, therefore, the conclusions presented in the FSAR are still valid. A core reload can typically affect accident analysis input parameters in three major areas: kinetic characteristics, control rod worths, and core peaking factors. Cycle 2 parameters in each of these three areas were examined as discussed in this section to ascertain whether new accident analyses were required. 6
.__j
Kinetics Parameters A comoarison of Cycle 2 kinetics parameters with the current limits is presented in Table 2. All parameters in Table 2 were found to be within the limiting range of values used in previou's safety analysis, except for the Doppler Temperature Coefficient (DTC). However, this change is small and since the DTC respresents only a small portion of the total negative reactivity feedback the effect is negligible and no accidents were reanalyzed as a result. Control Rod Worths Changes in control rod worths may affect shutdown margin, differential rod worths, ejected rod worths, and trip reactivity. Table 3 shows that the Cycle 2 shutdown margin requirements are satisfied. As shown in Table 2, the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 2 is less than the current limit. Cycle 2 ejected rod worths were less than those used for the Cycle 1 analyses, however, the hot-zero power beginning-of-life rod ejection case required reanalysis due to the peaking factors. Cycle 2 has a trip reactivity insertion rate which is different from that used for Cycle 1. Non-conservative deviations occur in various portions of the insertion. Fast transients are affected since the limiting point in the transient is reached during the insertion. Slow transients are less sensitive to trip reactivity. An investigtion' of the transients has shown chat only the locked rotor and loss of flow analysis may be af fected by these deviations. These transients have been reanalyzed and do not change the safety conclusions of the FSAR. Core Peaking Factors Peaking f actor evaluations were performed for the rod out of position and hypothetical steamline break accidents to ensure that the minimum 7 i
+,. ,.
DNB ratio remains above the DNBR design limits. These evaluations were performed utilizing the existing transient statepoint information from the reference Cycle 1 and peaking factors determined for the re. load core design. In each case, it was found that the peaking factor for Cycle 2 resulted in a minimum DNBR whicn was greater than the design limit DNBR. Consequently, for these accidents no further investigation or analysis was required. The Cycle 2 control rod ejection peaking factors were within the bounds of the Cycle 1 values, except for the beginning-of-life hot-zero power cases which were reanalyzed (Section 3.3). Cycle 2 peaking factor and power distribution evaluations have been performed for the dropped RCCA accident according to the new dropped rod methodology described in Reference 9. i- 3.3 INCIDENTS REANALYZED The hot-zero power beginning-of-life rod ejection accident case was I reanalyzed due to the Cycle 2 maximum gF exceeding the Cycle 1 values. Table 4 shows the pertinent rod ejection parameters used in the reanalysis. } The analyses were performed using the same methods as described in i References 1 and 10. The results for rod ejection show that the fuel i i rod conditions at the hot spot satisfies all the acceptance criteria i specified in Reference 10. Therefore, the safety conclusions given in i Reference 1 remain valid. The change in the allowable F 3g as a function of power resulted in a change to the K constants in the Overtemperature Delta-T and Overpower : Delta-T setpoint equations and a change to the Overtemperature Delta-T f(aI) function. l l l l 8
Since the Overtemperature Delta-T trip is used in tne bank withdrawal at power accident, this accident was reanalyzed with the new Overtem-perature Delta-T setpoints. The results show that the minimum DNBR remains above the limit value. This verifies that the conclusions in Reference 1 remain valid. In the LOCA analysis 2% uniform steam generator tube plugging was assumed. A total of six purge lines, four 24 in. diameter and two 12 in. diameter, were assumed to be open at the time of the accident. Initial temperatures used in the analysis were 105 F in the upper compartment and 125 F in the lower compartment. The loss of flow and locked rotor analyses were reanalyzed for the change in shape of the normalized trip reactivity versus position curve. The loss of flow reanalysis shows that the minimum DNBR remains above the limit value. The locked rotor analysis verifies that the conclusions of Reference 1 remain the same. 1 d 9
4.0 Technical Specification Changes To ensure plant operation consistent with design and safety evaluation conclusion statements made in this report and to ensure that these conclusions remain valid, several technical specification changes will be needed for cycle 2. These changes are summarized below. The changed technical specifications accompany this document (see appendix A). Description Incorporate the increase in K from 0.2 to 0 3 in the following relationship: Ph# $ [,1.55 1.0 + K (1.0-P{}. The technical specification changes are as follows: a) Replace figure 2.2-1, b) Change Table 2.2-1 as indicated on: page 2-7, Kj from SE 1.14 to $1.15 page 2-7, K from 0.009 to 0.011 - page 2-8, K from 0.00043 to 0.00055 page 2-10, 6 from 0.0012 to 0.0011 page 2-8, item (i) change 30% to 29% page 2-8, item (1) change 4% to 5% page 2-9, item (ii) change 30% to 29% l page 2-9, AT trip set-point change from 0.89 to 1.5 page 2-9, item (iii) change 0.8% to 0.86% page 2-9, item (iii) change 4% to 5% c) Change page 3/4 2-8,. Equation a R3 relationship change 0.2 to 0.3 Change page 3/4 2-11, Equation a R3 relationship change 0.2 to 0 3 Values in figure 3.2 3 remain unchanged, d) Revise page B 2-1 equation for Fjk to 1.55[1+0.3(1.0-P , e) Replace pages B 3/4 2-1 through B 3/4 2-6 Justification ThechangesprovideanincreaseinallowableF[hatreducedpowerin comparison with the cycle 1 technical specifications. The increase in allowable Fjk at reduced power allows optimization of the core loading pattern for full power operation. Safety Analysis Increasing the slope of the allowable F$h as a function of the power design limit from O'.2 to 0 3 requires reevaluation of the DNB protection setpoints. The setpoints for Sequoyah unit 2 cycle 2 have been updated to 4 10
r- \
=
+, ..
account for this increase in slope. ThemaximumcalculatedF$gthroughthe power range of Sequoyah unit 2 cycle 2 has been verified to be less than i tne value allowed with the 0.3 gF gslope multiplier. The effect on specific parameters is discussed in this report. Description Incorporate RAOC methodology for power distribution control into the Sequoyah technical specifications. The attached changes are as follows: a) Replace sections 3.2.1, 4.2.1.1, 4.2.1.2, B 3/4.2, and B 3/4.2.1, b) Replace figure 3 2-1, c) Delete sections 4.2.1.3 and 4.2.1.4, d) Delete figure B 3/4 2-1. Reference
- 1. R. W. Miller, et.al.; Relaxation of Constant Axial Offset Control For Sequoyah Unit 2, Cycle 2
- 2. Millstone Nuclear Plant, unit 2, cycle 4 SER, Amendment 61, October 6, 1980.
Justification In a plant incorporating RAOC operation, the technical specifications are modified to-remove all references to CAOC in section 3/4.2.1 and the corresponding bases. HAOC application has the following advantages: a) Maneuvering capability is enhanced and boron system duty can be minimized or smoothed, b) Operator action required to conform to power distribution technical specifications is reduced because rod motion corrections are reduced, c) Return to power capability after a trip is greatly increased. Safety Analysis The RAOC methodolgy utilizes the plant-specific LOCA and DNB margin to set l the allowable AI band. Limits on allowable operating axial flux imbalances l as a function of power level considering limiting condition I power distributions were found to be less restrictive than those resulting from LOCA FO considerations. Condition II analyses were evaluated relative to the axial power distribution assumptions used to generate DNB core limits. No changes in these limits are required for RAOC operation. 11
r The RAOC methodology is similar to CAOC methodology with the following exception. The method used for generating the xenon shape library is different. Previously, a library based upon xenon oscillation studies was used. For cycle 2, Westinghouse generates a xenon parameter range library and systematically reconstructs the xenon distribution when needed. In both methods, the entire range of xenon and rod insertion limits are covered. A detailed description of RAOC is included in reference 1. Description Delete the last sentence of action A of Limiting Condition for Operation 3.2.2. This deletes the requirement for going to hot standby to reduce the overpower Delta-T trip setpoint with Fg exceeding its limit. Justification The overpower Delta-T trip setpoint can be reduced one channel at a time while at power. It is not necessary to go to hot standby to make these setpoint changes. Safety Analysis The purpose of this action statement is to compensate for a measured P0 (z) exceeding its limit by reducing the overpower Delta-T setpoint. Reducing this setpoint provides a more conservative reactor trip. This action coupled with required power reduction and reduction of power range neutron flux high trip setpoint ensures FSAR assumptions remain valid should an accident occur under these conditions. Description Replace Fxy(z) surveillance currently in Sequoyah technical specifications with F n(z) surveillance. The attached changes are as follows: a) Sections 3.2.2, 4.2.2.2, 4.2.2 3, and 6.9.1.14 are replaced, b) The appropriate bases are changed (B 3/4.2). c) F xy requirements removed from section 3.3.3.2. i l l Reference R. W. Miller, et.al.; Relaxation of Constant Axial Offset Control For l Sequoyah Unit 2, Cycle 2 12
O e Justification Fxy(z) is implicitly included in the F (z) g measurement, and the intent of the technical specification is to monitor Fg(z) using a measured parameter. Therefore, the Fxy(z) surveillance requirements in the technical specifications are replaced with 0F (z) surveillance. Fg (z) surveillance provides the following advantages: a) Credit can be taken for the actual power distribution (and resulting F (z) values) measured in the plant. n b) Monitoring F 0(z) and increasing the value for expected plant maneuvers provides for a more convenient form of ensuring plant operation below the Fg(z) limit. c) The cycle dependent factors will be reported in a peaking factor report which will reduce technical specification changes. A description of the peaking factor report is included in section III.B.2 of the re ference . Safety Analysis F (z) surveillance implicitly includes F x z and retains the use of a 0 measured parameter to verify operation befo(w)the technical specification limits. FO surveillance is only a change to the plants' surveillance requirements and as such has no impact on the results of the cycle 2 analyses or safety parameters. A detailed description of the FO IZ) surveillance is included in section III.B.1 of the reference. Description For limiting condition for operation 3.6.1.5, change the upper limits for the upper and lower containment air temperatures to 105 F and 1250F, respectivcli. Justificaticn The Sequoyah LOCA analysis has been repeated with the upper limits of the containment upper and lower compartment air temperatures at 1050F and 125 F, respectively. Safety Analysis The upper limit on containment air temperature ensures that the containment air mass is limited to an initial air mass sufficiently high so that blowdown of the reactor coolant system (RCS) subsequent to a LOCA is consistent with analytical assumptions. The new LOCA analysis shows that l I the conclusions presented in the FSAR are still valid and the peak clad I temperature remains below 22000F. 13
. ". Description For limiting condition for operation 3.6.1.9, change the number of purge supply and exhaust lines allowed open to three pairs.
Justifications The Sequoyah LOCA analysis has been repeated with the supply and exhaust lines to the upper and lower containment and the instrument room all open at the initiation of the LOCA. The Tennessee Valley Authority (TVA) has also assessed the site boundary dose subsequent to a I )CA with seven (7) 24-inch purge lines opened. Safety Analysis The new LOCA analysis shows that the conclusions presented in the FSAR with respect to reactor coolant system (RCS) blowdown are still valid. Peak clad temperature remains below 22000F. Further, the TVA assessment of the site boundary dose subsequent to a LOCA shows the limits of 10 CFR 100 are met. Description Remove unnecessary statement in the bases describing quadrant power tilt ratio (section B 3/4.2.4). Justification The paragraph above this statement defines the purpose for the limit, therefore, this additional statement is unnecessary. Safety Analysis There are no safety implications. Description The number of igniters has been increased to 68 of which 66 will be required operable for Loo 3.6.4.3 Justification These changes reflect the installation of the permanent system. The permanent system hydrogen mitigation system is a two-train system with 34 igniters in each train. l The permanent hydrogen mitigation system employs controlled ignition to mitigate the effects of hydrogen during potential degraded core accidents , or class 9 accidents. The containment structures and key equipment have i been shown by analysis or testing to survive the pressure and temperature I loads from selected degraded core accidents and to continue to functon. The extensive research program has confirmed our analytical assu=ptions, demonstrated equipment survivability, and shown that controlled ignition can indeed mitigate the effects of hydrogen releases in closed vessels. The permanent hydrogen mitigation system is an adequate hydrogen control system that would perform its intended function in a manner-that provides adequate safety margins. 14 l
T Safety Analysis A per=anent hydrogen igniter system has been shown by experiment to be capable of mitigating the effects associated with hydrogen production ! during degraded core accidents. 5.0 DROPPED ROD ACCIDENT ANALYS'IS - REMOVAL OF OPERATING RESTRICTIONS Our July 22, 1982 letter to Ms. E. Adensam (see appendix D) formally requested your staff review the material submitted to you as NS-EPR-3545, January 20, 1982, and subsequently remove the interim operational restrictions before startup of Sequoyah unit 2, cycle 2. We believe the removal of these operating restrictions is justified and request your concurrence.
6.0 REFERENCES
- 1. Sequoyah Unit 2 Final Safety Analysis Report, USNRC Docket No. 50-328
- 2. Bordelon, F. M., et. al. , " Westinghouse Reload Safety Evaluation Methodology," WCAP-9273, March 1978.
3 Miller, J. V. (Ed.), " Improved Analytical Model used in Westinghouse Fuel Rod Design Computations," WCAP-8785, October 1976.
- 4. George, R. A., et. al., " Revised Clad Flattening Model," WCAP-8331, July 1974.
- 5. Bisher, D. H., et. al., " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.
~
- 6. Jones, R. G. and Iorii, J. A., " Operational Experience with Westinghouse-Cores," WCAP-8183 Revision 11, May 1982.
- 7. Camden, T. M., et, al., "PALADON - Westinghouse Nodal Computer Code,"
WCAP-9486, December 1978.
- 8. Hellman, J. M. (Ed.), " Fuel Densification Experimental Results and Model for Reactor Operation," WCAP-8219-A, March 1975.
- 9. Letter; Rahe (Westinghouse) to Berlinger (NRC), " Dropped Rod Methodology for Negative Flux Rate Trip Plants" NS-EPR-2545, January P0, 1982.
- 10. Risher, D. H., "An Evaluation of the Rod Ejection Accident in i
Westinghouse PWR's Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.
- 11. Letter; Hahe (Westinghousa) to Berlinger (NRC), " Relaxation of Constant l
Axial Offset Control (RAOC)," NS-EPR-2649, August 31, 1982. 15
TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS SEQUOYAH UNIT 2 - CYCLE 2 Region 1 2 3 4 Enrichment (w/o 2.124 2.617 3.101 3.50 U235)* Geometric Density 94.6 94.7 94.6 94.5 (percent Theoretical)* Number of Assemblies 5 72 48 68 Approximate 14400 16400 10500 0 Burnup at Beginning of Cycle 2 (MWD /MTU) I All fuel regions except region four are as-built values: Region four values are nominal. An average density of 94.5% theoretical was used for Region 4 evaluations. 0531L:6 16
1 TABLE 2 KINETICS CHARACTERISTICS SEQUOYAH UNIT 2 - CYC1.E 2 Previous Analysis Cycle 2 Value (1) (7) Value Moderator Density Coefficient 0 to 0.43 0 to <0.43 (ap/gm/cc) Least Negative Doppler - Only Power -10.2 to -6.7 -10.2 to -6.7 Coefficient, Zero to Full Power (pcm/% power)* Most Negative Doppler - Only Power -19.4 to -12.6 -19.4 to -12.6 Coefficient Zero to Full Power (pcm/% power)* Delayed Neutron Fraction .0044 to .0075 .0044 to .0075 Maximum Prompt Neutron Lifetime 5 26 5 26 (p sec) Maximum Reactivity Withdrawal 5 100 5 100 Rate from Subcritical (pcm/sec)* Doppler Temperature Coefficient -1.0 to -2.2 -1.0 to -2.9 (pcm/ F)*
-5
- pcm = 10 3, 1
l 17 0531L:6 l l
TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS SEQUDYAH UNIT 2 - CYCLE I AND 2 Four Loop Operation Cycle 1 Cycle 2 BOC EOC BOC EOC Control Rod Worth (% ap) All Rods Inserted Less Worst 6.61 6.18 5.32 6.22 Stuck Rod Less 10% II) 5.95 5.56 4.79 5.60 Control Rod Requirements (% ap) Reactivity Defects (Doppler, T,yg, 2.16 2.94 1.85 3.09 Void, Redistribution) Rod Insertion Allowance (RIA) 0.50 0.50 0.50 0.50 Total Requirements (2) 2.66 3.44 2.35 3.59 Shutdown Margin (1)-(2) (% ap) 3.29 2.12 2.44 2.01-Required Shutdown Margin (% ap) 1.60 1.60 1.60 1.60 18 0531L:6
. _a
TABLE 4 ROD EJECTION PARAMETERS FOUR LOOP OPERATION SEQUOYAH UNIT 2 Previous Analysis Cycle 2 Used in Values Analysis Values (1) HZP-BOL 0.83 Max Ejected Rod Worth, %Ap 0.88 0.624 N 14.05 14.90 14.90 Max F g
.0075 <.0075 .0075 Max B,ff HZP - Hot Zero Power BOL - Beginning of Life 19 0531L:6
Figure 1 CORE LOADING PATTERN Sequoyah Unit 2, Cycle 2 R P N M L K J H G F E D C B A 4 4 4 1 4 4 4
.)
4 4 4 2 4 1 4 2 4 4 4 12 8 8 12 2 4 4 2 2 4 2 3 Z 4 2 2 4 4 8 12 12 8 3 SS 4 2 3 3 2 2 2 2 2 3 3 2 4 4 4 4 2 3 2 3 3 3 3 3 2 3 2 4 4 5 12 12 4 2 4 2 3 2 3 2 3 2 3 2 4 2 4 12 12 6 4 4 2 2 3 3 2 3 2 3 3 2 2 4 4 - 8 8 7 o 2 1 3 2 3 2 3 1 3 2 3 2 3 1 2 90 8 4 4 2 2 3 3 2 3 2 3 3 2 2 4 4 8 8 9 4 2 4 2 3 2 3 2 3 2 3 2 4 2 4 12 12 10 4 4 2 3 2 3 3 3 3 3 2 3 2 4 4 12 12 11 4 2 3 3 2 2 2 2 2 3 3 2 4 4 4 2 2 4 2 3 2 4 2 2 4 4 8 12 12 13 sq 8 4 4 4 2 4 1 4 2 4 4 4 12 8 8 12 14 4 4 4 2 4 4 4 15 0 X - Region Number Y - Number of Burnable Poison Rods 2 - Secondary Source Rods 20
=
FIGURE 2 N MAXIMUM CALCULATED VALUES OF FAH WITH RESPECT TO TECHNICAL SPECIFICATION LIMITS 2.1 i :jr :i= =+
.:u:.purI.;=:=2x:== !r=~i = ui=r =: :.7c l ":
1 n:u==== x.
- 2. : = = t a.: =:n= = . - -
q==1n;= =:= = n = u r n :j::
-h . . .-
- . _ ;-:: .. 1=::: =:= ,
- n .. . r = == = : = : ; l
-I-
.:t = nlu- :- . ut
.._=_
= rut =: cirt: -
. . = . =..:.=. . .:u L::. .-......lr._.._....... - =: =
. . l:..
- 4.t..- .
._ :x. rs.=. - . .-. l =. _.:.x_.-
2.0 [ . .. .. ; _. l _l.=. x. -.,t.. =j =_.
. .: =. .. .22-.. _ - + . _ _ _ ._-l__ _- { xn2 =l:.
=: ====:=4 ==r = = .
c !:r x-=q.n : =t= : ci =t==-= : = 1- - - =x:n=[1 u- . .. t r . ; . = =ir G{ .. . . - _ _ . . . g 3
- n. .: .. n 1.o: _:. I c . . . g _. . ._,t.=. . ._ .:.=__._ . . . . ._ .:== _. _.__._
._ . _ q
- - _ . - - ._=.=_ ..r -: n. g n. n._ . . . .-
. . un. . _.
-4 .- _ ( +
; +
h.x=N-: n-.. t t=:=l=:r= ;=l =; ::== 4 =:r = l =. 1=_^. 1.9 . e4 . . - _ = , . . _ ;. c . , . . .r
. :- i. 7 .g,
- uir . m a . ;l=:a.: . :=
1.t:.n _ 2n
* : / r ..nir = :- nt r i = 1:n ...
- t: gi
=r= = :- u t .=~:= \ =T .
-v j $g .: f:== =. a.=;
..u u = i =.:= .:
;= =
a
- x. .
- = =:=t=:=
#+ t u_n. . j . _. // x -
= . _ _ = . _ =_=__~__. ,
l=__. :=_ .._ _.. 1.8
~
T_ . . . ( e+ + T.__.,1 ' . .. . - . . .._ _ ___. E_ .T_-'{7---'_Z 3.d. _ ...= =:= :: =l1
- nel .
/ n O.J
# =Yu . / ; -= :=j==:==hr: *r
_ .=u i : c_ :.
. . .. =_. t ..=ux - N x S(L ; =% /y/, : = =-:=21==:=- ::=r: =.
+
t z m : :. = c .= j = = r=. O
- a. ,
-r-
=u.< u. .===.
. . . . . _ . . . - -..m - . . -
.= =,
t- L
=.
;*7
.__-- 1=_= =._ :, =. =._.:.=_.{.__
. _ *P (f 1_.-.
m 1-----.-: =._4.. . _ ._ i ::pEEj=EMj - E' E~7^EE=t :-}: _ J ggg.E=jgg J g, =. =__ ;1_ =.,t x a:_. -
- 12. .- ._: :; u_ n. _u.;_. t=. a.__r t =. @ . . . , .
3 2 .
- 1 ' : 21= = ;; - .
r : :t :- n- *
- q. -:-r. :h,N. :: .
-i: 2. z .:
=.. . p. .:7 : . N. ,
.j
., ; . .a ; u.;
.. ..]-.- ,
.. ..g .
i 1.6 y. :,;p _
- q. a. . ..;. . .
! i T :iD Miis . F= = % .!
. h. .
-{ =l : E._.:.iiE_ E. x.- . --
- li . .'t= 1._
t xE.x. d '- .
- T;_u - " - -
---:= -+-~1--"- -j
.t.
n t u.. =j c. J
.- : = t : = =1 n-t ' ~
- =:( =.n:J=t=trn t n :
1.5 =:t;;j=- =-- 2:=g=4==:r2- _-.=- = .= = .: _ g
. . = , =, p=-
- = - a =tr . _ __ _ ==: --
2.:t r n:inc
- =:= =
=:1 = =rJ.r = :n -. _. i r r-
= 4 . . f_____._.i =_:==.c=:
i r =#. ; r= - + - - - -
- == =;= -
._. :=- ----+-:.--- =-
. =:==:= =.::=l_=t=. =:= . _.a . _ i .__ t__ L _ ._ . =
+--+
=:= =t= --- -e_ ----t-- _
, 1,4 .
= : = ; = -: = :reni_t== _= - -- - - - : .+ - - 1_2_ ): -
i
-0.0 0.2 0.4 0.6 0.8 1.0 Fraction of Rated Power t
21
FIGURE 3 670 ' - j j t - l , 4 , .
.___t.._
l , ., ,.-. , _ . . - - .. 7 . -. I, . .
. ;- i,
. ( .- j ... :! - -- j : T 650 N l: ! N' unn ne<2= 4 + -l-225D_p
]; dacceptabic-l l
- Opedatica
.. . . . j _ - {.=;..
. . . _ .q . .. .p .
N 2QOO p sia ' U2 I- --I : I'l ~ I 630 W :.is: =!dr r - ' !Nr !? 7:b
.li j ,j--
4pn Naby : : ti _= i --- : i. N!- :ij-- g .=y u. gy;. _t 3. _
#l#~
o g..t
+i:
610 -i - i il77 ; ndi~ - : 9'" ' e 2r bds =!!# :bla ;:P i . -
- t ..N . -i= :
- . f ~. : 3 j3
% il~. dl- . ;-if : .m .:I ~; .
4 l- : l_;;.
- :. .; n.- __j. . _. [
y -l 3"
l - "~l t:
- ' C
..z.. ,l.
~
590
~
"! . - -b i
uj.a: i-i:i _ y{ii y };. .
- d-i 4 :: 1
. . f; _ n N\g
- g. ..Ej..\l.\
7 _ .
~
- }c ACCEPTABLE, ; gj .i. j.;i- :: : :Ii- .:=l;9 - --
- -jf _gj- . ,
570 "-'I I ^" 0 *i "Elii 2 E = ~! =i3" *EI2+" 'i!* IT i- !. !- :li i I:i= =!:= Mkii- sti bli+ :h: @ - --
- b --p i.!:i :-j c; nin- :d j. " 2: p:2 ~ ill :-j ; E!: 7 -j-
.i I~ i' ! I:I i ihl-- El55 rr si ::jf E;I45[f .:j . I 550
. i: Ei :- ='
~
'" ::i:O ' =i ~
0 .2 .4 .6 .8 1.0 1.2 FRACTION OF RATED THERMAL POWER l FIGURE 2.1 Reactor Core Safety Limit - Four Loops in Operation l l f 22 __ .}}