ML20058N887
| ML20058N887 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/16/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058N873 | List: |
| References | |
| NUDOCS 9312220318 | |
| Download: ML20058N887 (4) | |
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UNITED STATES
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y-J NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D.C. 20E464001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
RELATED TO AMENDMENT NO.
182 TQ FACILITY OPERATING LICENSE NO. NoF-3
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TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY i
AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. I DOCKET NO. 50-346 1.0 JNTRODUCTION l
By letter dated November 9, 1992, as. supplemented on November 22,.1993, Toledo Edison Company (the licensee) proposed to change the Technical Specifications (TS) for.the Davis-Besse Nuclear Power Station, Unit 1.
The proposed TS' change reflects de-energization of the. Borated Water Storage Tank.(BWST) 1 outlet isolation valves DH-7A'and DH-78 in their open position during-t operational modes 1, 2, 3, and 4.
The licensee will revise the following TS-t to reflect the proposed change-(1)
TS 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - T,,;t 280 'F;
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(2)
TS Bases 3/4.5.2 and 3/4.5.3 Emergency Core Cooling Systems - ECCS Subsystems; l
(3)
TS Bases 3/4.6-2-1, Containment Systems - Depressurization and Cooling Systems - Containment Spray System i
1 (4)
Surveillance Requirement (SR) 4.5.2.d.2.b.
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.The supplemental letter of November 22, 1993, provided additional information and did not change the NRC. staff's proposed no significant hazards ~
determination.
l 9312220318 931216 "
DR ADOCM 05000346.,
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" P 2.0 EVALUATION BWST outlet isolation valves DH-7A and DH-78 are de-energized during modes 1, 2, 3, and 4 to praclude postulated inadvertent closure of the valves in the event of a fire, which could result in a loss of the availability of the BWST during normal plant shutdown.
Containment emergency sump recirculation valves DH-9A and DH-9B are de-energized during modes 1, 2, 3, and 4 to preclude postulated inadvertent opening of the valves in the event of a fire, which could result in draining the BWST to the containment emergency sump and the loss of this water source for normal plant shutdown.
The power removal from valves DH-7A/78 during modes 1, 2, 3, and 4 is accomplished by locally opening the breakers for the valve operators at their respective motor control centers (MCCs). This requires manual action outside the control room. Action to close the breakers is necessary during the change from emergency core cooling system (ECCS) injection to recirculation following a postulated loss-of-coolant accident (LOCA) in order to manipulate the valves.
When the valve operators are energized (breakers closed), an interlock is provided between BWST outlet isolation valves DH -7A/78 and containment emergency sump recirculation valves DH - 9A/9B so that all valves cannot be opened at the same time. However, when the breakers are open, this interlock function will be maintained administrative 1y because the control room operators are prevented by emergency procedure from shifting pump suction during a LOCA from the BWST to the containment emergency sump ttntil the BWST low-low level safety features actuation system (SFAS) level 5 signal is received.
The BWST capacity is about 482,778 gallons which permits the cooldown operation up to 37 minutes in the case of a LOCA.
It takes an operator only 3 minutes to get all four breakers closed once the BWST low-low level is received.
Thus, the emergency core cooling function will be accomplished in the case of a LOCA and it is acceptable.
The interlock will maintain its function once the valve operator's breakers are closed. The operators will open valves DH-9A/98.
The interlock will automatically close valves DH-7A/78 thereby, realigning pump suction from the BWST to the containment emergency sump. The valves will take approximately 15 seconds to reach their new positions.
Current plant emergency procedures for a large break LOCA require the operators to close the breakers for valves DH-9A/98.
Closure of the breakers for valves DH-7A/78 will be added in the emergency procedure.
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. The NRC staff questioned the risk impact of the additional operator actions required for a LOCA and the operator training of the new procedure. The staff had a telecon on September 2, 1993, with the licensee to resolve these issues.
l The licensee stated that LOCAs were not dominant contributors to core-damage for Davis-Besse. However, to assess the impact of this proposed change on the core-damage frequency, the operator actions associated with going to i
recirculation were revised to account for closing the breakers for DH-7A/7B.
All operator actions associated with going to recirculation following-a LOCA were modified.
frequency (approximately a half percent).The results showed a slight increas In addition, appropriate operator training will be provided prior to implementation of the license amendment' The licensee subsequently submitted this information in o letter dated November 22, 1993. The NRC staff has reviewed this inform 0 tion and finds it acceptable.
Therefore, its concerns are resolved.
In the case of a fire, de-energization of valves DH-7A/7B and DH-9A/9B will ensure that water from the BWST is available for plant shutdown. A manual operator action will maintain the interlock function in the case of a LOCA.
In addition, the licensee will revise the emergency procedure to include the closure of the breakers for valves DH-7A/7B in the case of a LOCA.
Thus, the LOCA consequences remain acceptable based on the BWST capacity and the operator action to close the breakers.
Based on the above, the NRC staff finds that the proposed TS changes for Davis-Besse are acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.
The State official had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves 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 issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (58 FR 28061). Accordingly, the amend.r.unt meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need t
be prepared in connection with the issuance of the amendment.
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5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense-and security or to the health and safety of the public.
Principal Contributor:
K. Desai Date:
December 16, 1993
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