ML20081H052
| ML20081H052 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/03/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17195A002 | List: |
| References | |
| NUDOCS 8311070291 | |
| Download: ML20081H052 (3) | |
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o UNITED STATES g
NUCLEAR REGULATORY COMMISSION o
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,I WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 77 TO PROVISIONAL OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. DPR-25 l
I COMMONWEALTH EDIS0N COMPANY DRESDEN NUCLEAR POWER STATION, UNIT NOS. 2 AND 3 DOCKET NOS. 50-237 AND 50-249
1.0 INTRODUCTION
By letter dated May 24, 1978, as sup(plemented July15, 1981, and May 2, 1983, Comonwealth Edison Company (Ceco) the licensee) proposed amendments to Appendix A of Operating License Nos. DPR-19 and DPR-25. The subject change involves Section 4.3.B.1.b of the Technical Specifications for Dresden Unit i
1
-Nos. 2 and 3.
The licensee has proposed to amend Section 4.3.B.1.b, i
Surveillance Requirements for Control Rod Coupling Integrity, to modify the wording for better understanding.
A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal Register on September 21, 1983 (48 FR 43131).
No request for hearing was received and no comments were received, t
2.0 EVALUATION Because of an early history of occasional control rod uncoupling in the i
l Dresden reactors, Technical Specification surveillance for uncoupling verifications in addition to those now required in the Standard Technical Specifications for General Electric Boiling Water Reactors, NUREG-0123, were included as specification 4.3.B.1.b.
This required confirmation of coupling using nuclear instrument response during a rod notch withdrawal.
When no instrument response was discernable at lower power, the response should be verified when the reactor is operating at power levels about 20%.
The intent of the specification was to provide a general check for all control rods but specifically for those rods with an uncoupling history.
The existing Technical Specifications 4.3.B.1.b wording implies that all rods be reverified at power levels above 20%.
To clarify the intent of surveillance requirement 4.3.B.1.b, CECO proposed a wording change in a letter dated May 24, 1978.
In a letter dated May 2, 1983, the wording was modified to further clarify the intent and consistency with the description and analysis in the Final Safety Analysis Report (FSAR).
i 8311070291 831103 PDR ADOCK 05000237 P
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., l The Final Safety Analysis Report (FSAR) Section 3.5.4, Surveillance and Testing for Control Rods, states in part "...During reactor operation individual control rod drive mechanisms can be actuated to demonstrate functional performance.
Each time a control rod is withdrawn a notch, the operator will observe the in-core monitors' indications to verify that the control rod is following the drive mechanism.
When the operator withdraws a control rod full out of the core, he tests the coupling integrity by trying to withdraw the rod drive mechanism to the overtravel position.
Failure of the drive to overtravel demonstrates rod to drive coupling integrity."
FSAR, Section 6.5.1, Design Basis for Control Rod Velocity Limiters, states in part "...The purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core."
i FSAR Section 14.2.1, Control Rod Drop Accident, shows that the analysis is based upon a fully inserted control rod assumed to fall out of the core after becoming disconnected from its drive and after the drive has been removed to the fully withdrawn position.
In order to assure that the control rod remains connected to its drive, and in the interest of good operating practices, the licensee's proposed change to Technical Specification 4.3.B.1.b, reaffirms the FSAR.
Further, when the nuclear instrumentation does not pro-vide evidence of the control rod movement, e.g., during a startup, the proof of coupling integrity for rods with uncoupling history will be conducted at a power level in excess of 20% where local power range monitors will give the necessary indication.
In addition, the licensee submittal of July 15, 1931 references General Electric Service Information Letter, SIL #52, Supplement 2, July 31,1974, which shows that improper installation of the control rod drive inner filter had resulted in causing control rods to become uncoupled when they reach position 48 (fully withdrawn). The licensee has implemented the improved i
GE overhaul procedure and test to assure proper installation of the inner i
filter. This has resulted in a significant reduction in events of uncoupled control rods.
Based on the foregoing, the staff finds the licensee's proposal to improve the wording of Technical Specification 4.3.B.1.b to be acceptable.
3.0 ENVIRONMENTAL QUALIFICATION The staff has determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this deter-mination, the staff further concludes that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5.(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal, need not be prepared in connection with the issuance of these amendments.
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4.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
S.0 ACKNOWLEDGEMENT The following staff members contributed to this evaluation:
T. M. Tongue K. R. Ridgway T. N. Tambling Dated: November 3, 1983
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