ML20033G518
| ML20033G518 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/06/1990 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20033G519 | List: |
| References | |
| ULNRC-2170, NUDOCS 9004100161 | |
| Download: ML20033G518 (7) | |
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'ddl U.S. Nuclear Regulatory Commission
____ ATTH:
Document Control Desk
' ", Mail Station F1-137 Washington, D.C.
20555 ULNRC-2170 Gentlemen:
DOCKET NUMBER 50-483 CALLAWAY PLANT REMOVAL OF POWER RANGE, NEUTRON FLUX, HIGil NEGATIVE BATE TJGP _ FROM_TECHN!QAL SPECIFlGMLO_lM O
Union Electric herewith transmits an application for amendment to Facility Operating License No. NPF-30 for Callaway Plant.
This amendment application revises the Technical Specifications by deleting the power range, neutron flux, high negative rate trip.
This change request is consistent with Westinghouse WCAP-11394(P) which was approved by NRC on October 23, 1989.
Plant modifications to remove the trip circuitry will be completed in Refuel 4, which is scheduled for Fall 1990.
Your approval is requested by September 1, 1990.
Attachments 1, 2 and 3 contain the Safety Evaluation, the Significant Hazards Evaluation, and the Proposed Technical Specification Changes in support of this amendment.
The Callaway Onsite Review Committee and the Nuclear Safety Review Board have approved this amendment request.
The proposed amendment will be implemented upon NRC approval and completion of plant modifications during Refuel 4.
Very truly yours, Donald F.
Schnell DS/dvd
'l' Attachments 9004100161 900306 ADOCK0500ggg3 p ygp/
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r STATE OF MISSOURI )
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Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Senior Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
/
By
/
5 Donald F.
Schnell Senior Vice President Nuclear SUBSC(IBED and sworn to before me this JMLf4 day of 7MA4<A>
1990.
Ais 0 A&u CARBARA JF fiDI ARY PUBLIC, STAf( OF I,'istopRt fu COMMISSION EXPIRES APRIL 22, 1993 l.
ST. LOUIS COUNTY l
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4 cci Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.
Washington, D.C. 20037 Dr. J. O. Cermak CFA, Inc.
4 Professional Drive (Suite 110)
Gaithersburg, MD 20879 R. C. Knop Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RR#1 Steedman, Missouri 65077 S. V. Athavale (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural Resources P.O. Box 176 Jefferson City, Mo 65102
es.. eet o bec D. Shafer/A160.761
/QA Record (CA-758)
Huclear Date E210.01 DFS/ Chrono D. F. Schnell J. E. Birk J. V. Laux M. A. Stiller G. L. Randolph R. J.
Irwin H. Wuertenbaecher W. R. Campbell A. C. Passwater R. P. Wendling D. E. Shafer D. J. Walker O. Maynard (WCNOC)
N. P. Goel (Bechtel)
T. P. Sharkey NSRB (Sandra Auston)
A. 4A (iv) c.s.y.n
Att0 chm:nt 1 o
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ULNRC-2170
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Page 1 of 2 SAFETY EVALMb. TION j
This amendment request revises the Technical Specifications to remove the power range, neutron flux, high negative rate trip or j
negative flux rate trip (NFRT) f rom the Nuclear Instrumentation System.
Technical Specifications affected by this change include Tables 2.2-1, 3.3-1, 3.3-2 and 4.3-1 and their associated bases.
In May, 1987, the Westinghouse Owners Group submitted Topical Report WCAP-11394(P), " Methodology for the Analysis of the Dropped Rod i
Event."
NRC found this methodology acceptable for referencing in license applications by letter dated October 23, 1989.
I The NFRT will be removed after the completion of a plant-specific, cycle-specific evaluation, using the approved Westinghouse methodology, which demonstrates that the DNB design basis r
is not exceeded during the course of the dropped Rod Cluster Control Assembly (RCCA) transient.
In the evaluation, three analyses provide
- 1) the statepoints, i.e.,
the reactor power, temperature and pressure l
at the most limiting time in the transient (transient analysis), 2) the hot channel factor at the most limiting conditions, consistent with primary system conditions and reactor power, in the transient (nuclear analysis), and 3) the DNB analysis at the conditions determined by Items 1 and 2 (thermal-hydraulic analysis).
All three analyses are performed using a parametric approach so that cycle-dependent conditions can be determined from the available data.
No credit is taken for any direct trip due to the dropped rods, and no automatic power reduction is aseamed to occur due to the dropped rods.
An analysis, using the methodology from WCAP-11394(P), will be completed as part of the Reload Safety Evaluation for each Callaway operating cycle.
The analysis will be performed using the cycle-dependent control rod worth at the full power insertion limits and the moderator temperature coefficient as input.
This analysis l
produces a hot channel factor which is used to calculate the limiting DNBR during the transient and ensure that the DNB design basis is met.
The removal of the NFRT will not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
Cycle-specific evaluations will be performed to demonstrate that the DNB design basis is met during the course of a dropped rod event, without relying on a direct reactor trip or automatic power reduction.
The proposed change to remove the NFRT will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report.
No reliance is placed on a direct reactor trip or automatic power reduction due to dropped RCCA(s).
Attcchm:nt 1 ULNRc-2170 I
Page 2 of 2 l
The removal of the NFRT will not affect the margin of safety as defined in the basis for any technical specification.
Cycle-specific evaluations will be performed to demonstrate that the DNB design basis l
is met during the course of a dropped rod event, without relying on a direct reactor trip or automatic power reduction.
t Based on the above, this amendment request does not adversely
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effect or endanger the health or safety of the general public and does i
not involve an unreviewed safety question.
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Attcchm:nt 2 ULNRC-2170 Page 1 of 1 SI GNI FI_CNiT_,_ HAZARD EVALUJTI ON This amendment request revises the Technical Specifications to remove the power range, neutron flux, high negative rate trip or negative flux rate trip (NFRT) from the Nuclear Instrumentation System.
Technical Specifications affected by this change include Tables 2.2-1, 3.3-1, 3.3-2 and 4.3-1 and their associated bases.
NFRT removal is based on completion of a cycle-specific analysis using the methodology of WCAP-11394(P), which was approved by NRC on October 23, 1989.
The analysis will be performed using the cycle-dependent control rod worth at the full power insertion limits and the moderator temperature coefficient as input.
This analysis produces a hot channel factor which is used to calculate the limiting DNBR during the transient and ensure that the DNB design basis is met.
The removal.of the NFRT does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Cycle-specific evaluations using an NRC approved methodology will be performed to demonstrate that the DNB design basis is met during the course of a dropped rod event, without relying on a direct reactor trip or automatic power reduction.
The method and manner of plant operation will be unchanged other than the deletion of unnecessary surveillances.
The proposed change to remove the NFRT does not create the possibility of a new or different kind of accident from any previously evaluated.
No reliance is placed on a direct reactor trip or automatic power reduction due to dropped RCCA(s).
The deleted circuitry has been shown to be non-essential.
The removal of the NFRT does not involve a significant reduction in a margin of safety.
Cycle-specific evaluations using an NRC approved methodology will be performed to demonstrate that the DNB design basis is met during the course of a dropped rod event, without relying on a direct reactor trip or automatic power reduction.
Based on the above discussions and those presented in Attachment 1,
it has been determined that the requested Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident or condition over previous evaluations; or involve a significant reduction in a margin of safety.
Therefore, the requested license amendment does not involve a significant hazards consideration.
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