ML20116F356
| ML20116F356 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1996 |
| From: | Terry C TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20116F358 | List: |
| References | |
| TXX-96433, NUDOCS 9608060324 | |
| Download: ML20116F356 (12) | |
Text
!
E Log # TXX-96433 File # 916 (2.1) 916 (2.2) 3 --
916 (3/4 2.5) 10010 7UELECTRIC' Ref. # 10CFR50.90 10CFR50.36 C. I.ance Terry v eransam am e aa y,y 9y U. S. Nuclear Regulatory Commission Attn: Document Control Desk i Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N05. 50-445 AND 50 446 SUBMITTAL 0F LICENSE AMENDHENT REQUEST 96 007 REVISION TO UNIT 1 RX TRIP SETPOINT l Gentlemen:
Pursuant to 10CFR50.90. TU Electric hereby requests an amendment to the CPSES Unit 1 Operating License (NPF 87) and CPSES Unit 2 Operating License (NPF 89) by incorporating the attached changes into the CPSES Units 1 and 2
{
Technical Specifications. These changes apply equally to CPSES Units 1 and j 2 except where a specific unit is indicated. l Based on analyses of the core configuration and expected operation for CPSES Unit 1, Cycle 6. revised core safety limit curves and revised Overtempereture N 16 reactor trip setpoints are required. In addition, the TU Electric Small Break LOCA Topical Report on the Core Operating Limits Report Technical Specification is incorporated. The topical report change is applicable to both Units.
Attachment 1 is the required affidavit. Attachment 2 provides a detailed description of the proposed changes, a safety analysis of the proposed changes and TV Electric's determination that the proposed changes do not involve a significant hazard consideration. Attachment 3 provides the affected Technical Specification pages marked up to reflect the proposed changes.
TV Electric requests approval of this proposed license amendment by October 25, 1996, with implementation of the Technical Specification changes to occur within 30 days after NRC approval.
In accordance with 10CFR50.91(b) TV Electric is providing the State of Texas with a copy of this proposed amendment.
l DOI 9608060324 960731 PDR ADOCK 05000445 p PDR Energy Plaza 1601 Bryan Street Dallas, Texas 7520134II
TXX 96433 Page 2 of 2 Should you have any questions, please contact Mr. Jimmy Seawright at (214) 812 4375.
Sincerely,
, e. ap.
f
! C. L. Ter r,y By. M l
Roger't. Walker' Regulatory Affairs Manager
! JDS/grp Attachments: 1. Affidavit i l 2. Description and Assessment Affected Technical Specification pages as 3.
revised by all approved license amendments 1
c- Mr. L. J. Callan, Region IV !
Mr. T. J. Polich NRR '
Ms. L. J. Smith, Region IV l Resident Inspectors, CPSES Mr. Arthur C. Tate Bureau of Radiation Control.
Texas Department of Public Health i 1100 West 49th Street i i Austin, Texas 78704 I
l i
Attachment 1 to TXX 96433 Page 1 of 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I i
In the Matter of )
)
Texas Utilities Electric Company ) Docket Nos. 50 445
) 50 446 i (Comanche Peak Steam Electric ) License Nos. NPF-87 Station, Units 1 & 2) ) NPF-89 AFFIDAVIT Roger D. Walker being duly sworn, hereby deposes and says that he is Regulatory Affairs Manager for TV Electric, the licensee herein: that he is duly authorized to sign and file with the Nuclear Regulatory Commission this License Amendment Request 96-007: that he is familiar with the content thereof: and that the matters set forth therein are true and correct to the i best of his knowledge, information and belief. j W -
1/
- ~
It694r n/ Walker Regulatory Affairs Manager STATE OF TEXAS )
)
COUNTY OF DALLAS )
., Subscribed and sworn to before me, on this 30th day of July , 1996.
, Notary gblic Gayle R. Peck s .
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ATTACHMENT 2 to TXX 96433 DESCRIPTION AND ASSESSMENT
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Attachment 2 to TXX 96433 Page 1 of 8 DESCRIPTION AND ASSESSMENT
.I. BACKGROUND Based on analyses of the core configuration for CPSES Unit 1, Cycle 6.
revised core safety limit curves and Overtemperature N 16 reactor trip setpoints are required. The core safety limits are the loci of points of THERMAL POWER, Reactor Coolant System (RCS) pressure and average temperature below which either the calculated Departure from Nucleate Boiling Ratio (DNBR) is no less than the safety analysis limit value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. Important parameters used to establish these lines include the RCS flow rate, the design nuclear enthalpy rise hot channel factor (Fu), the cycle-specific reference axial power shape, and the cycle-specific. core configuration. The core safety limits are calculated for pressures between the Pressurizer Pressure Low and Pressurizer Pressure -
High reactor trip setpoints and for powers up to the overpower reactor trip setpoint.
The Overtemperature N-16 reactor trip setpoint is calculated such that a reactor trip will be initiated before the core safety limits are exceeded.
The 0vertemperature N 16 reactor trip function provides DNB protection from events which result in changes in power pressure, temperature, or axial power shape. As for the core safety limits, the range over which the overtemperature reactor trip setpoint is calculated is bounded by the Pressurizer Pressure Low and Pressurizer Pressure High reactor trip setpoints and the overpower reactor trip setpoint. The operation of the Main Steam Safety Valves also limit the power / temperature range over which the overtemperature trip setpoint must provide DNB protection. If an event results in an axial power shape whir,h is more severe, from a DNB standpoint, than the reference axial power shape,-the Overtemperature N 16 trip setpoint is automatically reduced in order to assure that the DNB protection afforded by the trip'setpoint remains adequate. The axial power shapes are " recognized" by the plant instrumentation in terms of delta-flux (AI): 1.e., the difference between the currents generated by the top two excore power range neutron flux detector sections and the currents generated by the bottom two detector sections. These instruments are spanned to detect axial flux imbalances in the range of i 75% RTP.
TU Electric has submitted the topical report "Small Break Loss of Coolant Accident Analysis Methodology," RXE 95 001-P via Letter logged TXX 95315 Dated December 19, 1995. It is anticipated that the NRC staff will issue a Safety Evaluation Report approving the use of the methodologies described in this report with the expectation that TU Electric perform a Small Break LOCA analysis using this methodology in order to support Unit 1 Cycle.6 operation. In order to accomplish this activity, it is necessary to include the topical report in the list of NRC approved methodologies in Technical Specification 6.9.1.6b.
L Attachment 2 to TXX-96433 >
Page 2 of 8 '
II. DESCRIPTION OF TECHNICAL SPECIFICATIONS CHANGE REQUEST The core safety limits of Figure 2.1 la and the Overtemperature N 16 reactor trip function setpoints are revised to reflect the analyses performed to support operation with the Unit 1 Cycle 6 core configuration.
This configuration consists of 192 fuel assemblies manufactured by Siemens Power Company and one fuel assembly manufactured by Westinghouse Electric Company. The mixed core penalty is significantly less than in previous core designs due to the greatly reduced number of Westinghouse fuel assemblies. In addition, the Unit 1 Cycle 6 axial power distributions are calculated to be slightly more skewed toward the top half of the core than those calculated for Unit 1 Cycle 5. The changes in the axial power distributions require a revision to the f(AI) trip reset function of the Overtemperatu:e N 16 trip setpoint. For this cycle, it was determined that !
no trip setpoint reduction was required for those bottom-skewed axial power !
differences (AI < 0.0) over the span of the AI instrumentation. In i addition, the core safety limits and the Overtemperature N 16 trip setpoint l are revised to reflect the Unit 1 Cycle 6 core configuration with the j l reduced mixed core penalty by: ;
A. Using the methodologies specified in Technical Specification 6.9.1.6b, calculations and analyses have been performed to identify i the new reactor core safety limit curves for Unit 1. Technical Specification Figure 2.1 la have been revised to replace the old curves with the new reactor core safety limit curves.
B. Using the new reactor core safety limit curves from "A" above, calculations and analyses have been performed to determine new N-16 related setpoint values and parameters for Unit 1 as noted below:
In Technical Spectfication Table 2.21, Note 1 for the Overtemperature N-16 Trip Setpoint, the following Terms will be
- changed as noted
I o K, from 0.0134/*F to 0.0173/*F
= K from 0.000719/psig to 0.000890/psig 3
a qt q, range from 65% and +4% to -65% and +4.6%
= Overtemperature N-16 setpoint reduction from 1.81% to 0.0% for each percent that the magnitude of q q, exceeds 65%
a Overtemperature N 16 setpoint reduction from 2.26% to 3.04% for I each percent that the magnitude of qt q exceeds +4.6% (current value +4%)
l o Footnote identifying that no setpoint reduction is required for the span of the AI indication.
In Technical Specification Table 2.2-1, Note 2, for the
- Overtemperature N 16 Allowable Value, the maximum amount by which the l Trip Setpoint is allowed to exceed the computed Trip Setpoint, is
! increased from 3.51% to 3.64%.
C. In addition to the analysis of the reactor core safety limits and the
[
DNB related parameters for the Unit 1 Cycle 6 reactor core
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Attachment 2 to TXX 96433 Page 3 of 8 configuration (including revised Overtemperature N 16 setpoint equation coefficients), Technical Specification 6.9.1.6b is revised to allow the use of the methodology presented in the TU Electric topical report, "Small Break Loss of Coolant Accident Analysis Methodology," RXE 95 001 P. TU Electric intends to use the methodology described in this Topical Report to analyze the effects of the small break loss of coolant accident.
SUMMARY
To summarize. TU Electric proposes changing the reactor core safety limits in CPSES Unit 1, Cycle. 6. As a result of the new reactor core safety limits, the Overtempersture N 16 trip setpoints have been recalculated.
The Unit 1, Cycle 6 61alyses have been performed using methodologies which are NRC approved and satisfy the applicable safety analyses limits.
Finally, Small Break LOCA Analysis will be used to analyze the effects of a small break loss of coolant accident. These changes are consistent with the Westinghouse Improved Standard Technical Specifications (NUREG 1431.
Revision 1).
III. ANALYSIS The core safety limits were recalculated based on the Unit 1 Cycle 6 core configuration. The changes relative to the limits in the current Technical Specification are due to the reduced allowance required for the effects of the mixed core penalty. The methodologies used by TV Electric to determine the reactor core safety limits are wholly consistent with and represent no change to the Technical Specification 2.1 BASES for Safety Limits.
The Reactor Trip System setpoint limits specified in Technical Specification 2.2. Table 2.21 are the nominal values at which the reactor trips are set for each functional trip. The trip setpoints have been selected to ensure that the core and Reactor Coolant System (RCS) are ,
prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences. The Overtemperature N 16 trip function initiates a reactor trip which helps protect the core and RCS from exceeding their safety limits.
The Overtemperature N 16 trip provides core protection to prevent DNB and vessel exit saturation for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that: the transient is slow with respect to piping delays from the core to the N 16 detectors: the pressure is within the range between the Pressurizer High and Low pressure reactor trip setpoints; and the power is less than the Overpower N 16 trip setpoint. The Overtemperature N-16 reactor trip setpoint is automatically varied with coolant temperature, pressurizer pressure, and axial power distribution.
With a normal operations axial power distribution, the Overtemperature N 16 reactor trip limit is always below the reactor core safety limits. If the axial flux difference is greater than that of the reference distribution,
Attachment 2 to TXX 96433 l Page 4 of 8 l
as indicated by the difference between top and bottom power range neutron flux detectors, the Overtemperature N 16 reactor trip setpoint is automatically reduced according to the notations (Note 1) in Technical Specification 2.2, Table 2.21. This reduction provides protection consistent with the reactor core safety limits.
The Overtemperature N-16 reactor trip setpoint was recalculated in order to optimize the operating margin while continuing to protect the core safety limits. The change in the value of the K coefficient primarily reflects 2
the reduced mixed core penalty,. which, on the core safety limits figure, shifts the intersection of the DNB limit and vessel exit boiling limit to a higher power. l l The f(AI) overtemperature trip reset function is designed to automatically reduce the Overtemperature N 16 reactor trip setpoint if an axial power >
shape is detected which is more severe, with respect to DNB, than the '
l reference shape used in the development of the core safety limits. The slope of the trip reset function required for Unit 1 Cycle 6 is steeper than the slope contained in the current Technical Specification. The range l for which no trip reset is required is extended out slightly from a AI of l 4.0% RTP to one of +4.6% RTP. In addition, no trip reset is required for j the most negative flux imbalances.
i
- The FSAR Chapter 15 event most affected by the change in the l Overtemperature N 16 trip setpoint'is the rod withdrawal at power event l presented in FSAR Section 15.4.2. This event was reanalyzed using the L revised Overtemperature N 16 trip setpoint. All relevant event _ acceptance l
criteria were determined to be satisfied. Therefore, this change does not result in any reduction in any margin of safety. The other, less limiting l events, in which the Overtemperature N 16 setpoint is used will be l evaluated prior to Unit 1. Cycle 6 operation through the Reload Safety l Evaluation performed in accordance with 10CFR50.59. This safety evaluation will be contingent upon the NRC approval of the Overtemperature N 16 setpoint change.
All analyses of the core safety limits, the overtemperature N 16 trip setpoint, and the Rod Withdrawal at Power event were performed in accordance with the NRC approved methodologies listed in the Technical Specification 6.9.1.6b, Items 9, 10, 11, 12, 13 and 14.
TU Electric has submitted the topical report "Small Break L(ss of Coolant Accident Analysis Methodology," RXE 95-001 P. It is antici nted that the NRC staff will issue a Safety Evaluation Report approving t u. use of the methodologies described in this report. In order to use the methodology in this report to support Unit 1 Cycle 6, it is necessary to include the topical report in the list of NRC approved methodologies in Technical l Specification 6.9.1.6b.
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Attachment 2 to TXX-96433 Page 5 of 8 l
IV SIGNIFICANT HAZARDS CONSIDERATI0iS ANALYSIS i
TV Electric has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10CFR50.92(c) as discussed below:
- 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
A. Revision to the Unit 1 Core Safety Limits l Analyses of reactor core safety limits are required as part of reload calculations for each cycle. TV Electric has performed the analyses I of the Unit 1, Cycle 6 core configuration to determine the reactor core safety limits. The methodologies and safety analysis values result in new operating curves which, in general, permit plant operation over a similar range of acceptable conditions. This change means that if a transient were to occur with the plant operating at the limits of the new curve, a different temperature and power level might be attained than if the plant were operating within the bounds of the old curves. However, since the new curves were developed using NRC approved methodologies which are wholly consistent with and do not represent a change in the Technical Specification BASES for l safety limits, all applicable postulated transients will continue to be properly mitigated. As a result, there will be no significant increase in the consequences, as determined by accident analyses, of any accident previously evaluated.
B. Revision to Unit 1 Overtemperature N 16 Reactor Trip Setpoints, Parameters and Coefficients As a result of changes discussed, the Overtemperature N 16 reactor trip setpoint has been recalculated. These trip setpoints help ensure that the core safety limits are maintained and that all applicable limits of the safety analysis are met.
Based on the calculations performed, the safety analysis value for Overtemperature N 16 reactor trip setpoint has changed. This essentially means if a transient were to occur, the actual temperature and power level achievable prior to initiating a reactor trip could be slightly higher. However, the analyses performed show that, using the TV Electric methodologies, all applicable limits of the safety analysis are met. This setpoint provides a trip function which allows the mitigation of postulated accidents and has no impact on accident initiation. Therefore, the changes in safety analysis l values do not involve an increase in the probability of an accident j and, based on satisfying all applicable safety analysis limits, there is no significant increase in the consequences of any accident previously evaluated.
In addition, sufficient operating margin has been maintained in the i
1 Attachment 2 to TXX 96433 !
Page 6 of 8 j
overtemperature setpoint such that the risk of turbine runbacks or i j
reactor trips due to upper plenum flow anomalies or other operational ,
transients will be minimized thus reducing potential challenges to l the plant safety systems. ;
C. Incorporation of TU Electric Small Break LOCA Topical Report.
RXE 95-0001 P.
TU Electric has submitted the topical report "Small Break Loss of Coolant Accident Analysis Methodology." RXE 95 001 P and plans to use
, the. report to support Unit 1, Cycle 6. In order to accomplish this activity it is necessary to include the topical report in the list of NRC approved methodologies in Technical Specification 6.9.1.6b.
Use of this topical report is contingent upon NRC approval:
therefore inclusion of this report in Section 6 of the Technical i Specifications is administrative in nature and does not change the i probability or consequences of an accident. j
SUMMARY
l The changes in the amendment request applies NRC approved
! methodologies, changes in safety analysis values, new core safety limits and new N 16 setpoint and parameter values to assure that all applicable safety analysis limits have been met. The potential for an operational transient to occur has been reduced and there has been i no significant impact on the consequences of any accident previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes involve the use of revised safety analysis values and the calculation of new reactor core safety limits and reactor trip setpoints. As such. the changes play an important role in the analysis of postulated accidents but none of the changes effect plant hardware or the operation of plant systems in a way that could initiate an accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from
! any accident previously evaluated.
- 3. Do the proposed changes involve a significant reduction in a margin of safety?
In reviewing and approving the methods used for safety analyses and calculations, the NRC has approved the safety analysis limits which establish the margin of safety to be maintained. While the actual impact.on safety is discussed in response to question 1. the impact
- on margin of safety is discussed below
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( Attachment 2 to TXX 96433 l Page 7 of 8 I A. Revision to the Unit 1 Reactor Core Safety Limits The TV Electric reload analysis methods have been used to determine new reactor core safety limits. All applicable safety analysis limits have been met. The methods used are wholly consistent with Technical Specification BASES 2.1 which is the bases for the safety limits. In particular, the curves assure that for Unit 1. Cycle 6 the calculated DNBR is no less than the safety analysis limit and the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. The acceptance criteria remains valid and continues to be satisfied; therefore, no change in a margin of safety occurs.
l B. Revision to Unit 1 Overtemperature N 16 Reactor Trip Setpoints, l Parameters and Coefficients l
l Because the reactor core safety limits for CPSES Unit 1. Cycle 6 are recalculated, the Reactor Trip System instrumentation setpoint values for the Overtemperature N 16 reactor trip setpoint which protect the reactor core safety limits must also be recalculated. The Overtemperature N 16 reactor trip setpoint helps prevent the core and Reactor Coolant System from exceeding their safety limits during
, normal operation and design basis anticipated operational .
! occurrences. The most relevant design basis analysis in Chapter 15 l cf the CPSES Final Safety Analysis Report (FSAR) which is affected by i the change in the safety analysis value for the CPSES Unit I l l Overtemperature N 16 reactor trip setpoint is the Uncontrolled Rod ;
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section 15.4.2). This event has been re-analyzed with the revised safety analysis value for the Overtemperature N 16 reactor trip setpoint to demonstrate compliance with event specific acceptance criteria.
Because all event acceptance criteria are satisfied, there is no l
degradation in a margin of safety.
The nominal Reactor Trip System instrumentation setpoints values for the Overtemperature N 16 reactor trip setpoint (Technical Specification Table 2.21) are determined based on a statistical l combination of all of the uncertainties in the channels to arrive at a total uncertainty. The total uncertainty plus additional margin is applied in a conservative direction to the safety analysis trip setpoint value to arrive at the nominal and allowable values l presented in Technical Specification Table 2.21. Meeting the l requirements of Technical Specification Table 2.21 assures that the Overtemperature N 16 reactor trip setpoint assumed in the safety analyses remains valid. The CPSES Unit 1. Cycle 6 Overtemperature N 16 reactor trip setpoint is different from previous cycles which provides more operational flexibility to withstand mild transients
, without initiating automatic protective actions. Although the setpoint is different, the Reactor Trip System instrumentation
- setpoint values for the Overtemperature N 16 reactor trip setpoint i are consistent with the safety analysis assumption which has been l analytically demonstrated to be adequate to meet the applicable event i
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Attachment 2 to TXX 96433 I
Page 8 of 8
! acceptance criteria. Thus, there is no reduction in a margin of safety.
C. Revise 6.9.1.6b to include Topical Report RXE-95-001 P, "Small Break Loss of Coolant Accident Methodology"
- TV Electric has submitted the topical report "Small Break Loss of l i Coolant Accident Analysis Methodology," RXE 95 001 P and plans to use i the report to support Unit 1 Cycle 6. In order to accomplish this l activity, it is necessary to include the topical report in the list of NRC approved methodologies in Technical Specification 6.9.1.6b.
Use of this topical report is contingent upon NRC approval; therefore, inclusion of this report in Section 6 of the Technical l Specifications is administrative in nature and does not reduce the margin of safety.
l Using the NRC approved TV Electric methods, the reactor core safety limits are determined such that all applicable limits of the safety analyses are met. Because the applicable event acceptance criteria continue to be met, there is no significant reduction in the margin 1 of safety.
Based on the above evaluations. TU Electric concludes that the activities associated with the above described changes present no significant hazards consideration under the standards set forth in 10CFR50.92(c) and, accordingly, a finding by the NRC of no significant hazards consideration is justified. ,
V. ENVIRONMENTAL EVALUATION TU Electric has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10CFR20 or would change an inspection or surveillance requirement. TV Electric has evaluated the proposed changes and has determined that the changes do not involve (I) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), an environmental assessment of proposed change is not required.