ML20216H848
| ML20216H848 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 09/23/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216H845 | List: |
| References | |
| NUDOCS 9910040072 | |
| Download: ML20216H848 (3) | |
Text
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l' UNITED STATES s
g NUCLEAR REGULATORY COMMISSION t
'F WASHINGTON, D.C. 30806-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.190 TO FACILITY OPERATING LICENSE NO. DP AND AMENDMENT NO.187 TO FACILITY OPERATING LICENSE NO. DPR-30 COMMONWEALTH EDISON COMPANY 6NQ MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated June 29,1999, the Commonwealth Edison Company (Comed the licensee) submitted a request for changes to the Quad Cities Nuclear Power Station, Units 1 and 2, Technical Specifications (TSs). The requested amendment revises Surveillance Requirement (SR) 4.3.C.1 for control rod testing to increase the " notch" testing surveillance interval for partially withdrawn control rods from once per 7 days to once per 31 days. The change is J
consistent with the _ content of the improved Standard Technical Specifications (iSTS) j (NUREG-1433, Revision 1).
2.0 EVALUATION
]
SR 4.3.C.1 currently requires the following~:
o When above the low power setpoint of the RWM [ Rod Worth Minimizer), all withdrawn control rods not required to have their directional control valves disarmed electronically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
a.
At least once per 7 days, and The proposed change replaces SR 4.3.C.1.s with the following:
a.
At least once per 7 days *) for each fully withdrawn control rod, and at least once per 31 days *) for each partially withdrawn control rod, and A footnote, shown below, applicable to SR 4.3.C.1 is added to clarify the applicability of the requirement:
9910040072 990923 PDR ADOCK 05000254 P
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, b Not required to be performed until 7 days (for fully withdrawn) or 31 days (for partially withdrawn) after the control rod is withdrawn and above the low power setpoint of the RWM.
Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is frn to insert on a scram signal. These surveillances are not required when THERMAL POWER is less than or equal to the actuallow power setpoint of the RWM since the notch movement may not be compatible with the requirements of the control rod insert / withdraw sequence and the
. RWM. The 7-day frequency of SR 4.3.C.1 is based on operating experience related to the changes in control rod drive (CRD) performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control rods are tasted at a 31-day frequency, based on the potential power reduction required to allow the contrcl rod movement, and considering the large testing sample of SR 4.3.C.1. Furthermore, the 31-day frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
The above less restrictive requirements have been reviewed by the staff and have been found to be acceptable. The changes do not present a significant safety question in the operation of the plant because (1) at full power a large percentage of control rods (typically 80 - 90%) are fully withdrawn and will continue to be exercised each week. This is a significant sample size when looking for an unexpected random event, (2) the TS will continue to require at least 10 percent of the control rods on a rotating basis be scram time tested whether inserted, or partially withdrawn (current TS 4.3.D), and (3) operating experience has shown stuck control rods to be an extremely rare event while operating. The TS requirements that remain are l
consistent with current licensing practices, operating experience and p! ant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected. Therefore, this change is acceptable.
3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the lilinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a surveillance requirement. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously ismed a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (64 FR 40905). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmentalimpact statement or I
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3-environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activit!es will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
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Principal Contributor: Robert M. Pulsifer i
Date:- September 23, 1999 9
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