ML20126B264
| ML20126B264 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/14/1992 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-92611, NUDOCS 9212220030 | |
| Download: ML20126B264 (7) | |
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!llll$l'D Log # TXX-92611 File # 916
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C 10010 r
r Ref. # 10CFR50.90
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6 TUELECTRIC December 14, 1992 w imam J. cahm. Jr.
Group t'we l'residem U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N05. 50-445 AND 50-446 SUBMITTAL LAR 92-002, SUPPLEMENTAL 1 - ADDITIONAL INFORMATION COMBINED UNIT 1 AND 2 TECHNICAL SPECIFICATIONS REF:
Letter logged TXX-92537, dateu October 29, 1992, from William J. Cahill, Jr. to the USNRC Gentlemen:
Via the referenced letter. TV Electric submitted a supplement to License Amendment Request (LAR)92-002.
The NRC staff has requested an expanded no significant hazards determination evaluation for the changes in Enclosure (1) of the referenced letter.
The expanded evaluation is attached.
Should you have any questions, please contact Mr. Manu Patel of Nuclear Licensing at (214) 812-8298.
Sincerely,
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William J. Gnhill, Jr.
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MCP/grp Attachments:
- 1. Affidavit
- 2. No Significant Hazards Determination Evaluation c-Mr. J. L. Milhoan, Region IV Resident Inspectors CPSES (2)
Mr. T. A.
Bergman, NRR Hr. B. E. Holian, NRR Mr. D.
K.
Lacker Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street s
9212220030 921214 PDR ADOCK 05000445 hI su) N. Olive Street LB. 81 Dallas. Texas 75201 l>
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, to TXX-92611 Page 1 of 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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Texas Utilities Electric Company
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Docket Nos. 50-445
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and 50-446 (Comanche Peak Steam Electric
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Station, Unit 1 & 2)
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William J. Cahill, Jr. being duly sworn, hereby deposes and says that he is Group Vice President. Nuclear of TV Electric, the lead Applicant herein; that he is duly authorized to sign and file with the Nuclear Regulatory Commission this Additional Information to the Supplement to license Amendment Request (LAR)92-002 to the CPSES Unit 1 Operating License (NFP-87); that he is f amiliar with the content thereof; and that the matters i
set forth therein are true and correct to the best_of his knowledge, information and belief.
2/d.
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WilliamJ.Cah[T1,Jr.
/7' Group Vice President, Nuclear 5,TATEdFTEXAS
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i COUNTY OF DALLAS)
Subscribed and sworn to before me, on this 14th day of December 1992.
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. to TXX 92611
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Page 1 of 4 1)
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
i The following changes affect the pressure / temperature limits which define the acceptable regions of operation:
o the 50 psi decrease in heatup/cooldown curve i
o the 39 psi decrease in a portion of the PORV setpoint curve.
o the addition of the 20 F/ hour heatup curve and o
the l'F increase in the criticality limit.
j As such, these changes do not affect the probability that overpressure events (the events of concern) would occur.
These changes only affect j
the conditions from which events could be initiated, e
The consequences of overpressure events are limited by assuring that the applicable stress limits for the Reactor Coolant System pressure boundary (e.g., ASME Boiler and Pressure Vessel Code.Section III.
Appendix G and 10CFR50, Appendix G) are not exceeded.
Because the revised acceptable regions of operation still assure that these limits are not exceeded, these changes have no impact on the consequences of an overpressure event.
4 The change of terminology to use the adjusted reference temperature (ART) is editorial only.
The changes in the ART values are updates based on the results from the reactor vessel material irradiation surveillance program.
The ART
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values on these figures are informational only and do not directly affect plant operations or performance.
The margin for instrumentation error described in the text above the heatup and cooldown curves is changed to reflect the additional 50 psi of measurement uncertainty which is incorporated into the curves.
The margin for instrumentation error as noted on these figures is informational only and does not directly affect plant operation or performance.
In summary, these changes are either editorial or descriptive or only affect the limits which' define the acceptable range for operation.
As such, these changes do not change the probability or consequences of an accident previously evaluated, i
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I to TXX-92611 i
Page 2 of 4 a
2)
Does the proposed change create the possibility of a new or different i
kind of accident from any accident previously evaluated?
L The proposed changes are either editorial or descriptive or only affect the limits which define the acceptable regions for operation.
No changes are proposed which could result in a new or different kind of 1
i accident from any accident previously evaluated.
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3)
Does the proposed change involve a significant reduction in a margin of l
safety?
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The margin of safety is determined by the failure point for a particular j
system, structure, or component and the acceptance criteria which are j
established to ensure that the failure _ point is not reached during the i
events of concern.
For these specifications, the failure points of
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concern are the points at which brittle fracture failures could occur in the Reactor Coolant System (RCS) pressure boundary.
The acceptance s
criteria are, in part. the pressure / temperature limits provided by Figures 3.4-2 and 3.4-3 and the PORV setpoint limits in Figure 3.4-4.
For the Reactor Coolant System, the severity of the stresses which can exist are determined by the actual temperature, pressure, and j
heatup/cooldown rates which are allowed.
The method of determining the i
pressure / temperature limit curves for various heatup and cooldown rates is based on approved calculational methodologies which establish an acceptable margin between the actual stresses and the failure point of the materials.
An allowance for measurement uncertainties of the instruments is then combined with the actual stresses to produce the j
heatup/cooldown limit curves.
The different heatup/cooldown rate curves are calculated by the same j
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methodologies and the same instrument uncertainties apply.
Therefore, the curve for each heatup/cooldown rate provides essentially-l the same margin of safety.
For example, for each heatup curve, the j
maximum stress, and therefore the minimum margin of safety, exists at the heatup rate designated for that curve.
Consequently, the margin of 4
safety for a 60*F/ hour heatup while operating on the 60 F/ hour curve is essentially the same as the margin of safety for a 100 F/ hour heatup while operating on the 100 F/ hour curve.
The new 20*F/ hour heatup curve was determined using the same calculational methodologies as the new 60 F/ hour and 100 F/ hour heatup curves.
The same instrument uncertainties were applied to develop all of these curves and therefore each provides essentially the same margin 3
of safety.
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' to TXX-92611 Page 3 of 4 1
The instrumentation used to assure operation within the allowed range for these curves has not been changed and therefore the uncertainty of the instrumentaton has not changed.
However, a previously unrecognized 50 psi uncertainty has been incorporated into all of the heatup/cooldown and PORV setpoint curves.
Because the cctual instrument uncertainty is unchanged but the existing curves are being lowered to incorporate an allowance for the additional 50 psi uncertainty, the maximum stress conditions allowed to exist have been reduced and the margin of safety has been increased.
As with the heatup and cooldown curves discussed above, the maximum allowable PORV setpoints specified in Figure 3.4 4 are selected to assure that pressure / temperature limits are not exceeded.
The transients of concern are analyzed including factors such as equipment time delays, instrumentation uncertainties, valve opening times, etc.,
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to ensure that the pressure overshoot does not exceed the limits established by Appendix G of Section 111 of the ASME Boiler and Pressure Vessel code.
The PORV setpoint curve is determined by combining the results of the analysis with the limits determined by the ASME code.
To account for the additional 50 psi uncertainty, the analysis was revised.
When the analyses was revised, some overly conservative 3ssumptions were replaced with acceptable but less conservative assumptions.
As a result, the PORV setpoint curve has decreased when below 237*F by 39 psi e
instead of 50 psi.
The other 11 psi was absorbed by the revised l
assumptions.
No changes were made to the installed plant hardware or instrumentation.
Thus, an actual plant transient will progress in the same manner following this change in PORV setpoint curve except for the initial conditions.
In other words, the actual pressure overshoot for the limiting transient is not expected to change. The pressure limits per the ASME code remains essentially unchanged.
Therefore, the 39 psi i
reduction in the setpoint limit (i.e., initial conditions) increases the margin of safety.
The change in the terminology to use " ART" and the descriptive change in i
the pressure margin have no direct.effect on either plant operations or on the actual margin.
The changes in the ART values are the results of an update of e11 sting-l calculations and are based on the actual neutron fluence obtained from the reactor vessel material irradiation surveillance program.
The revised ART values are used in the calculations of the revised heatup and cooldown curves.
The effects of the changes to the ART values are reflected in the heatup and cooldown curves, but do not directly affect any margins or plant operations.
The fact that the ART values decreased indicates that the reactor vessel is more resistant to brittle fracture; however, the change is so small as to be inconsequential.
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( to TXX-92611 Page 4 of 4 i
The revised criticality limits reflected on Figure 3.4-2 specify pressure / temperature limits for critical core operation in order to provide additional margin during actual power production, i n accordance j
with 10CFR50. Appendix G.
The curves are applicable for RCS temperatures below approximately 350"F.
However, because criticality 4
below an average RCS temperature of-551'F is prohibited by Technical Specification 3.1.1.4, the changes to these limits are descriptive, and j
have no effect on margin or plant operations.
I In summary, the proposed changes are either editorial or descriptive in nature witn no effect on margin, or represent an increase in the actual j
margin of safety.
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File No. : 10002 COMMITMENT TRACKING Page No. :
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OUTGOING CORRESPONDENCE EVALUATION 4
SHEET i
DOCUMENT IDENTIFICATION Title / Subject : SUBMITTAL LAR 92-002.SUPPLEhENT 1-ADDITIONAL INFORMATION COMBINED UNIT 1 AND 2 TECH SPEC Log No. : TXX 92611 j
File
- 916 10010 Date
- 12/11/1992 I ITEM TRACKING l
Tracking Description Document /No.
No new commitments identified in letter : TXX-92611 j
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-* Item Types : EA - External Action OB - Operational Basis DB - Design Basis IA - Internal Action DO - Description Only C
_ Committed-Date P - Proposed Date CDMPLETED BY :
REVIEWED BY : \\
Initials MCP Date l'A S h Initials' Date b
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