ML20072U884
| ML20072U884 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/08/1991 |
| From: | Schuster T COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| TAC-79724, TAC-79725, TAC-79726, TAC-79727, NUDOCS 9104190337 | |
| Download: ML20072U884 (4) | |
Text
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Comm:nwealth Edison
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> 1400 Opus Place
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O "J Downers Grove, Illinois 60515
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i April 8,1991 Dr. Thomas E. Murley, Director Office Of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk
Subject:
Byron Station Units 1 and 2 Braldwood Station Units 1 and 2 Supplement to Application for Amendment to Facility Operating Licenses NPF-37/66 & NPF 72/77 Appendix A, Technical Specifications TAC # 79724/79725 and 79726/79727 NBC_DockeLNosm60145_4L455_and_50-456L4.57
Reference:
(a)
January 26,1990 letter from T.K. Schuster to T.E. Murley s
Dear Dr. Murley:
Pursuant to 10 CFR 50.90 Commonwealth Edison Company (CECO) aro30 sed to amend Appendix A, Technical Specifications of Facility Operating Licenses WP :-37/66 and NPF-72/77 for Byron and Braldwood Stations respectively in the letter of Reference (a). The proposed amendment requested changes to the Action Statement requirements of Specification 3.1.3.1 and its associated Bases Section. In response to a request from your staff, two clarifications to the wording of Attachment 2 of Reference (a) are being provided. The first and second page of Attachment 2 have been revised and the revisions have been high lighted with a vertical bar in the right hand margin.
The first clarification added the word " Turbine" to clarifv that " Turbine Power Level" could be changed to maintain average core temperature (' rave) in accordance with its program value (Tref). Changing Turbine power level would initially and directly change Tref, but it would also ultimately change Tave. The second change to was an expansion of the description of the effect of a Rod Control Urgent Failure Alarm on the control rods. This change is self explanatory and is contained on the second revised page of Attachment 2. These clarifications have no effect on the previous finding that the proposed amendment involves no signif; cant hazards consideration.
9104190337 91o4o8 PDR ADOCK 03o00454 P
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pr. Thomas E. Murley April 8,1991 '
j Commonwealth Edison is notifying the State of Illinois of this supMement to an application for amendment by transmitting.a copy of this letter and its atlachment to the designated State Official.
To the b.<st of my knowledge and belief the statements contained herein are true and correct. In some respects, these statements are not based on my personal knowledge but upon information received from other Commonwealth Edison and l
contractor employees. Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
Please direct any questions you may have concerning this matter to this Respectfully,
% &/ w T.K. Schuster Nuclear Licensing Administrator State of-
. County of OCb Sy&;ga x av uswasg L s i k aall o
' YNN M. WLOD ARSKI Qi ARY POSUC. STATE OF ILUNOIS 2( COWISSION EXPIRES 6/25/94
Enclosure:
Attachment 2 revised pages (2) cc: W. Kropp-Byron -
S. Dupont Braidwood A. Hsia NRR-R. Pulsifer NRR W. Shafor Rill Office of Nuclear Facility Safety IDNS TKS/scl:lD867:2
ATTACHMENT 2 D E SCRIPllORAN D_B AS E SRElHELPJ30EOSED_GHANG ES t
The proposed changes revise the Action Statement for Technical Specification 3.1.3.1, Moveable Control Assemblies and the associated Bases Section.
The revision adds an Action Statement to address the condition when more than one i
fulllength rod is inoperable but still capable of insertion into the core upon receipt of a reactor trip signal. Under this condition, the Action Statement permits continued operation for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before a unit shutdown is requirec. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> aermits time for diagnosis and repair of the Inoperable but triapable rods. This time erension can potentially prevent an unnecessary transient on uhe plant required by a shutdown while still maintaining the safety of the unit since the reactor control and shutdown rods can perform their intended safety function of insertion into the core upon receipt of a reactor trip signal.
The purpose of the Control Rod Drive System (CRDS is two fold. The CRDS performs a control function which serves to insert or withd) raw rod cluster con assemblies within the reactor core to control average core temperature to a program value (Tref). During a temporary loss of the ability to adjust rod height, the function of maintaining average core temperature in accordance with its program value (Tref) can be accomplished by either boron changes or by changing Turbine power level. The l
CRDS performs its protection or safety function, reactor trip, by alacing the reactor in a subcritical condition when a safety system setting is approachec with any assumed credible failure of a single active component. The protection system (reactor trip function)is designed to be independent and isolated from the rod control system.
Therefore, a failure in the rod control system does not impact the ability of the protection portion of the CRDS to perform a reactor trip.
The oaerability of the shutdown and control rod banks are initial assumptions in all safety ana yses which assume rod insertion uaon reactor tria. This ensures the assumed reactivity is available for insertion, in adcition, operabit ty requires maintenance of proper bank withdrawal and overlap re distribution and control rod alignment are maintained. quirements so that correct pow Technical Specification 3.1.3.1 requires all shutdown and control rods to be operable and positioned within + 12 steps of their grou a step counter demand p(osition.
The moveable control assemblies Technical Specificat on 3.1.3.1 ensures that 1) acceptable power distribution limits are maintained, (2) the minimum shutdown margin is maintained and (3) the potential effects of rod misalignment on associated accident analyses are limited.
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ATTACHMENT 2 (CONTINUED)
DESC BlPllONANDEASES_O EIHEfBQEOS E D_CH ANG E S For one rod being inoperable, the Action Requirements vary significantly depending on whether the rod is immovable or untrippable or whether it is still trippable. For the rod that is immovable or untrippable the Unit must be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. For a rod that is Inoperable but still trippable, unit operation may continue indefinitely provided the Action Requirements of maintaining rod alignment and sequence and insertion limits are met; or power and the associated trip setpoints are reduced and the shutdown margin, power distribution, and reevaluation of certain accident analyses are performed per the Action Requirements. For individual rod inoperability, the current Technical Specification acknowledges the significance between the rod being immovable or known to be untrippable and the rod just being inoperable but trippable. However, the Technical Specification does not permit the same flexibility when more than one rod is inoperable but still trippable. Action b requires that with more than one full length rod inoperable or misaligned from the group step counter demand position by more than 112 steps than the Unit must be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The proposed amendment request provides a distinct Action Rec ulrement for more than one inoperable but trippable rod that is consistent with the signif cance of the malfunction, and the original bases of the specification. Having more than one rod inoperable due to being untrippable is more significant than having more than one rod that cannot be stepped due to an electrical malfunction, but remaining triapable.
Distinguishing between these types of malfunctions will allow an appropr ate time period to complete corrective action commensurate with the significance of the malfunction. Therefore, the proposed amendment allows continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the new action protects the original bases of the Moveable Control Assemblies Specifications by requiring the remainder of the rods in the group (s) with the inoperable rods be aligned to within + 12 steps of the inoperable rods while observing the other specifications of the section. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> interval permits a reasonable amount of time for diagnosis and repair of the inoperable rods. Thus, possibly eliminating a unit shutdown that can result in an unnecessary transient on the plant while the rods are still capable of perforrning their intended safety function, in most cases when more than one rod is found to be trippable (and aligned but inoperable, the malfunction can be traced to the Rod Control System. The typical) situation that has occurred at Byron and Braldwood Stations is that when multipIe rods are inoperable a Rod Control Urgent Failure alarm occurs. This alarm is indicative of an electronic / electric malfunction occurring within the logic or power sup aly portion of the CRDS. The inputs to the Rod Control Urgent Failure Alarm which innibit rod motion result from electrical failure in either the Logic or Power Cabinets of the Rod Control System. Failures causing the alarm could be loose or missing cards, component failures, or control logic errors in either the Logic or Power Cabinets. The effect of the interlock associated with the alarm is to energize both the stationary and moveable grippers thereby inhibiting control rod stepping motion. The conditions that can cause this alarm do not affect the ability to trip any control rods. This may result in a situation where the control rods cannot be stepped in or out of the reactor. However, the rods remain trippable and are therefore capable of performing their safety function. Since the majonty of CRDS malfunctions can be repaired without a reactor shutdown and since alant conditions are not outside any accident analysis assumptions, there is time avallote to locate the malfunction and restore the rods to operable status.
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