ML20072L466
| ML20072L466 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/24/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072L457 | List: |
| References | |
| NUDOCS 9408310308 | |
| Download: ML20072L466 (7) | |
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UNITED STATES i
NUCLEAR REGULATORY COMMISSION
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WASHINGTON. D.C. 20555-0001 SAFETY. EVAL 0ATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 4
RELATED TO AMENDMENT N0.192 TO FACILITY OPERATING LICENSE NO. DPR-53 AND AMENDMENT NO.169 TO FACILITY OPERATING LICENSE NO. OPR-69 BALTIM0RE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 AND 2 DOCKET NOS. 50-317 AND 50-318
1.0 INTRODUCTION
By letter dated November 3, '1993, the Baltimore Gas and Electric Company (BG&E, the licensee) submitted a request for changes to the Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2, Technical Specifications (TSs). The' requested changes would modify portions of the surveillance requirements in TS 4.5.2.e.1 and its assoc 4 ted TS Pases.
The purpose of the requested TS changes is to support removal of u,e Autoclosure Interlock (ACI) which provides automatic isolation of the Shutdown Cooling (SDC) System suction line isolation valves.
In addition, the setpoint for the Open Permissive Interlock (OPI) would also be revised.
The associated TS Bases would be updated to reflect the proposed changes.
Removal of the ACI is expected to reduce the incidence of events involving loss of SDC System capability to cool during nonpower operations due to inadvertent closure of the suction valves.
The proposed TS changes are in accordance with the recommendations of Generic Letter (GL) 88-17, " Loss of Decay Heat Removal." GL 88-17 recommended changes to existing TSs which restrict or limit overall safety benefit.
Since the ACI is a contributor to the loss of SDC system events, its proposed removal is consistent with the recommendations in GL 88-17.
The OPI setpoint was reviewed by BG&E as part of this TS change request.
It was determined that the setpoint should be the actual reactor coolant system (RCS) pressure at the instrument tap location. The requested change for the OPI setpoint from 300 pounds per square inch absolute (psia) to 309 psia reflects the actual pressure at the instrument tap when the RCS pressure is 350 psia.
2.0 BACKGROUND
The SDC system is based on a design pressure of 350 psig while the RCS is designed for a maximum pressure of 2,500 psia.
During normal operating conditions, a double barrier between the high pressure RCS and the low pressure SDC system is provided by two motor-operated suction valves. These 9408310300 940824 PDR ADOCK 05000317 P
. valves are closed when the RCS is hot and pressurized (normal operating conditions) and open when.the SDC system i: in operation (cooldown or refueling).
Each of these suctio'n valves is provided with manual controls on the main control board.
There are two automatic interlocks associated with the control circuitry; the ACI and the OPI.
The OPI prevents the suction valves from being opened when RCS design suction pressure equals or exceeds 350 psia. The OPI function and the associated TSs are not affected by the proposed amendment except for a revision of the OPI setpoint at which the opening of the SDC suction isolation valves is allowed.
This revision corrects the setpoint of the actual RCS pressure at the instrument tap elevation from 300 psia to 309 psia which is the RCS pressure at the tap elevation when the SDC system suction pressure is 350 psia.
The purpose of the SDC system ACI is to ensure that the low pressure piping of the SDC system is properly isolated from the RCS pressure during startup operations. When the valves are in the open position, the ACI causes the valves to close automatically when RCS pressure increases to a value above the predetermined autoclosure setpoint. Although the ACI protects tne low pressure piping of the SDC system, spurious actuation could terminate decay heat removal during shutdown cooling operations.
The Commission and industry have previously recognized the safety benefits from removing the ACI circuitry from the SDC system. A disadvantage of the autoclosure feature is the possibility of an inadvertent valve closure during SDC system operation resulting in the loss of decay heat removal capability.
The safety benefits of removing the ACI circuitry were stated in the Commission's case study on long term decay heat removal, Case Study Report AE00/C503, " Decay Heat Removal Problems at U.S. Pressurized Water Reactors,"
December 1985 and also in a study performed for the Commission by Brookhaven National Laboratory, NUREG/CR-5015, " Improved Reliability of Residual Heat Removal Capability in PWRs as Related to Resolution of Generic Issue 99," May 1988.
In GL 88-17, " Loss of Decay Heat Removal," the Commission requested that TSs which restrict or limit the safety benefit of actions discussed in GL 88-17 should be identified and that appropriate changes should be submitted. One of the items listed by GL 88-17 that could limit such safety benefits was the ACI.
In parallel with the Commission's activities, Combustion Engineering (CE) completed a report, CE NSPD-550, " Risk Evaluation of Removal of Shutdown Cooling System Auto-Closure Interlock," September 1989, that documented the results of a generic analysis of the impact of removing the ACI from the SDC system. This report is generally applicable to Calvert Cliffs, Units 1 and 2, and was supplemented by a plant specific evaluation CE NSPD-548," Requirements for the Removal of the Shutdown Cooling Suction Valve Auto-Closure Interlock,"
September 1989.
The evaluation was performed to determine the change in an interfacing system loss-of-coolant accident (ISLOCA) frequency, the change in
. SDC system unavailabil'ity, and the impact on mitigating low-temperature overpressure events due to the re,moval of ACI.
3.0 EVALUATION In support of its requested TS changes, dG&E referenced the CE reports discussed in Section 2.0 above. The CE reports included a probabilistic risk analysis (PRA) regarding the removal of the SDC system ACI and a plant specific evaluation for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
BG&E described how improveme-is identified by the above reports will be implemented at the Calvert Ciiffs units.
These results take into account the impact of the removal of the ACI feature on the SDC system inlet isolation valves.
BG&E concluded that the implementation of their proposed design, TSs, administrative control, and procedure changes will reduce the frequency of a SDC system overpressurization event and increase the SDC system availability at the Calvert Cliffs units.
The staff reviewed BG&E's proposal against the recommendations and guidance in the CE NSPD-550 report and the plant specific evaluation in the CE NSPD-548 report. The hardware change proposed for the Calvert Cliffs units is the removal of the ACI function from the SDC system suction valves. The OPI will remain intact with a proposed setpoint change.
For Calvert Cliff, Units 1 and 2, Bq&E evaluated the following items to support the removal of the ACI.
1.
The means available to prevent Event V concerns:
Event V is a LOCA event outside of containment. The Calvert Cliff-design provides for a double barrier between the RCS and the SDC system.
The design provides a very high confidence that at least one barrier can be established and maintained under postulated conditions.
This is accomplished through the use of separate power supplies, independent valve position indication, and the separation of control and indication power sources.
Procedural controls, personnel training, autcmatic audible and visual alarms, and the OPI function minimize the potential for operator error when establishing double isolation or attempting to defeat it once it is established.
The OPI will prevent opening of the SDC system suction isolation valves when the RCS pressure exceeds the interlock setpoint.
2.
The alarms to alert the operator of an improperly positioned SDC system isolation valve:
Prior to the removal of ACI, automatic visual and audible alarms will be provided in the main control room to inform the operator if any one of the SDC system suction isolation valves is not fully closed when RCS pressure is above the alarm setpoint.
The alarms will be tested at each refueling outage.
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Verification of the adeqcacy of relief valve capacity:
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A review by the licensee of the original design basis of the SDC system relief valve indicated that the ACI has not been credited in the selection of the limiting events or mitigation of the resulting transients. The use of the SDC system relief valve along with administrative controls provide overpressure protection on the SDC system. As a result, the removal of the ACI will have no adverse impact on SDC system overpressure protection provisions.
4.
Means other than ACI to ensure that both isolation valves are closed:
Closure of both isolation valves prior to pressurizing the RCS remains primarily an operator function.
The alarms provide backup if the operator fails to close both valves.
In addition to the alarms described in Item 2 above, the proposed modification will use position indication, operating procedures, and personnel training to ensure that the double barrier is established when needed.
5.
Assurance that the OPI is not affected by ACI removal:
The OPI function will be maintained in its present form. Assurance that the OPI function is not affected will be confirmed by testing the operability of the OPI function after the ACI is removed.
6.
Assurance that valve position indication will remain available in the control room after ACI removal:
There is continuous valve position indication on the main control board.
The indication for the valve position utilizes DC control power.
Power for valve position indication has been separated from control power.
Circuit operability will be verified after the ACI is removed.
7.
Assessment of the effect of ACI removal on SDC system availability and low-temperature overpressure (LTOP) event:
The effect of ACI removal on LTOP protection for the reactor vessels was not assessed since no portion of the SDC system at Calvert Cliffs is used for this purpose.
The LTOP system at Calvert Cliffs makes use of a lowered setpoint on the Power-0perated Relief Valves (PORVs) located on the pressurizer, along with administrative controls, to provide protection against brittle fracture of the RCS pressure boundary. As a result, ACI removal has no effect on LTOP.
The use of the SDC system relief valve along with administrative controls provide protection against overpressurizing the SDC system.
A plant specific PRA evaluation was performed, as previously noted, to evaluate the affect of the proposed change on the probability of ISLOCA, SDC system availability, and potential mitigation of slow acting pressure transients.
The usa of ACI for isolation of the SDC system from the RCS
. during slow acti_ng pressure transients was not part of the original plant design bas.is However, considering that the potential for using the ACI-for this purpose exist, BG&E mvaluated the impact of ACI removal on this type of overpressure protection.
The results of the risk evaluation show that:
(1) the frequency of an interfacing system LOCA decreases with the removal of the ACI circuitry from the SDC system when accompanied by the addition of a control room alarm and procedural enhancements, (2) removal of the ACI increases SDC system availability, and (3) the net effect of ACI deletion from the SDC system is an overall improvement in safety.
The setpoint for the OPI function was reviewed as part of this proposed change by BG&E.
The surveillance requirement for the SDC system OPI provides assurance that the SCC system suction isolation valves are prevented from being remotely cpened when the RCS pressure is at or above the SDC system design suction pressure of 350 psia.
It was determined that the value in the TSs should be the actual RCS pressure at the instrument pressure tap location which is 309 psia when the SDC system suction pressure is 350 psia.
Revising the OPI actuation from 300 psia to 309 psia is a result of establishing a clear basis for this value and takes into effect the correct elevation at the point of the measurement for measuring the RCS pressure. The TS surveillance test procedure will contain the necessary compensation to be applied to this value to account for instrument uncertainties.
4.0 Summary The removal of the ACI from the SDC system and the other actions, as detailed above, is consistent with the recommendations of GL 88-17 and has an overall positive impact on safety. Therefore, BG&E's proposal to remove the ACI capability and to modify the TS surveillance associated with TS 4.5.2.e.1 by removing the requirement to verify tLat the ACI isolates the SDC system is acceptable.
Inclusion of the requirement in TS 4.5.2.e.I to verify that the SDC system OPI prevents the SDC system suction isolation valves from being opened with a simulated or actual RCS pressure equal to or greater than 309 psia is acceptable in that it provides reasonable assurance that the SDC system will not be over pressurized. Revising the OPI setpoint to 309 psia is acceptable sense it reflects the measured pressure at the instrument tap location when the SDC system pressure is 350 psia and the surveillance testing will take into account instrument uncertainties.
The revisions to TS Bases Section B 3/4.5 are acceptable in that they reflect the changes to the TSs detailed above.
5.0 STATE CONSULTATI0f]
In accordance with t'he Comrnission's regulations, the Maryland State official was notified of the proposed isstrance of the amendments. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requireme,t with respect to installation or use of a facility component located within the restricted area as defined ir. 10 CFR Part 20 and changes surveillance requirements. 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 issued a proposed finding that the amendments involve no significant hazards consideration, and there has beea no public comment on such finding (58 FR 64600). 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 environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 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 activities 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.
Principal Contributor:
H. Balukjian Date: August 24, 1994 l
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Docket Nos. 50-317 August 44, 1994 and 50-318-Mr. Robert E. Denton. ~
Vice President - Nuclear Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusty, Maryland 20657-4702
Dear Mr. Denton:
SUBJECT:
ISSUANCE OF AMENDMENTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO. 1 (TAC N0. M88164) AND UNIT NO. 2 (TAC N0. M88165)
The Commission has issued the enclosed Amendment No.192 to Facility Operating f.
License No. DPR-53 and Amendment No.169 to Facility Operating License No.
DPR-69 for the Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2, respec-tively. The amendments consist of changes to the Technical Specifications in response to your application transmitted by letter dated November 3, 1993.
The amendments modify the. surveillance requirements to reflect the removal of the auto-closure interlock from the shutdown cooling system and revises the setpoint for the open permissive interlock.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Reaister notice.
Sincerely, ORIGINAL SICidH2 EL Daniel G. McUonald, Senior Project Manager
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Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Amendment No.192 to DPR-53 2.
Amendment No.169 to DPR-69 3.
Safety Evaluation cc w/ enclosures:
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