ML20092A525
| ML20092A525 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/29/1992 |
| From: | Mcdonald D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20092A530 | List: |
| References | |
| NUDOCS 9202100158 | |
| Download: ML20092A525 (5) | |
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1 UNITED STATES o
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.167 TO FACILITY OPERATING LICENSE NO. DPR-53 gA(TMQ3LGAS,ANDE([CTRICCOMPANY CALVERT CLIFFS NUCLEAR POWER PLANT UN T NO. 1 n
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1.0 INTRODUCTION
By letter dated December 31, 1991, the Baltimore Gas and Electric Company (BG&E or the licensee) submitted a request for a change to the Calvert Cliffs Nuclear Power Plant, Unit No.1, Technical Specifications (TS). The re change would revise TS 4.5.1.a.2, Emergency Core Cooling System (ECCS) quested Safety Injection Tanks (SITS)-Surveillance Requirements on an exigent basis.
Specifically.-a footnote is to be added which would be effective from the date the requested amendment is issued to exempt only motor operated valve, 1-MOV-644, from the requirement to verify that the SIT isolation valves are in the open position at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The footnote would expire prior to entering Mode 3 subsequent tt refueling outage 10, which is currently scheduled for the spring of 1992.
The verification for 1-MOV-644 is not needed during this period because a temporary modification to secure the valve in the open position has been implemented. The valve stem has been welded to the valve yoke until the valve can be repaired or replaced during the upcoming refueling outage. The bent-valve stem and temporary modification have resulted in the remote valve position indicators, which are normally used to verify the valve position, not being available.
The surveillance requirements for 1-MOV-644 were being met, on ar interim basis, by. making containment entries at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during power operation.
The licensee also requested a Temporary Waiver of Compliance (TWOC) while the NRC staff reviewed the exigent TS amendment request. 'The
.TWOC was granted by NRC letter dated December 13, 1991, which relieves BG8E from making containment entries based on the temporary modification which secured 1-MOV-644 in its open position.
2.0 EVALUATION On December 22,1991, MOV-644 was used to isolate check valve SI-245, which is located downstream of SIT 12B, to perform repairs on the leaking check' valve during an unplanned outage. -Subsequent to the repairs, 1-MOV-644 failed to L
open. The valve was inspected and the valve stem was found to be bent and was I
causing the motor-operator to bind. The motor-operator was disable to allow the valve to be manually opened.
This ac+1on resulted in the loss of the position indication and the valve was w aified by welding the valve stem to t.
i the valve yoke to assure the isolation valve maintained its required open position for Mode 1, 2 and 3 operation.
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As previously noted, the modification is temporary until the upcoming refueling outage in the spring of 1992.
BG&E considered repairing 1-MOV-644 which would outage, require cold shutdown with a partial drained necessitate a seven-day (mid-loop operation), or possible use of a freeze plug.
reactor coolant system However, BG&E desired to avoid any unnecessary entry into a partially drained reactor coolant system condition, due to increased plant vulnerability to unexpected transients while in this condition.
Therefore, BG&E opted to make the temporary modification described above.
BG&E considers that the welded open valve provides adequate assurance that the valve will remain open and that verification of valve position is unneces:ary until the valve is replaced or repaired during the upcoming refuelina outage.
The NRC staff has reviewed the BG&E safety analysis and justification provided in its December 31, 1991, request.
The surveillance requirements of TS 4.5.1.a.2 are to provide reasonable assurance that all of the SIT isolation valves remain in their open position during Mode 1, 2 and 3 operation to allow the SITS to perform their safety function of reflooding the reactor core during).certain conditions following a design bases Loss-of-Coolant Accident (LOCA The first consideration is to assure that the valve, 1-MOV-644, is in the open position.
BG&E used several indications to assure themselves that the valve was open.
These included:
- 1) performing a test which required partial flow of water from the SIT through the repaired check valve and 1-MOV-644 to a drain line, during which the SIT level was monitored ar.d a decrease noted indicating the isolation valve was open, 2) the ex when 1-MOV-644 was manually opened, 3) pected arrount of stem travel w 's measure the valve stem moved freely once the stem nut was moved past the binding portion, 4) a weak link analy:,is indicates that the stem is the most likely location of failure, and 5) the force required to open the valve once unseated was consistent with that of a properly operating valve.
Although none of these indications provided absolute assurance that 1-MOV-644 is in the full open position, collectively they provide reasonable assurance that the valve is open.
In addition to the weak link analysis indicating that the stem is the most likely portion of the valve to fail, the next likely failure would be the bottom of the wedge at the valve seating surface. When 1-MOV-644 was c6 sed during the check valve repair no bypass leakage was observed indicating no damage in the wedge / valve seating surface area.
Thus the NRC staff has concluded that there is reasonable assurance that1-MOV$44isintheopenposition.
The second consideration was to repair 1-M0V-644 or provide a means of temporarily securing the valve in its open position.
The decision and supporting rationale for performing a temporary modification in lieu of repairing the valve until the upcoming refueling outage (approximately 3 months) was previously discussed.
The NRC staff agrees that a seven-day outage requiring cold shutdown and mid-loop operation would result in undue hardship, the unit would be vulnerable to unexpected transients while in mid-loop operation, and that no corresponding safety benefit would be gained for the short period of operation until the upcoming outage. Therefore, the NRC staff has determined that the use of a temporary modification for the short time frame is acceptable.
The temporary modification used by BG&E to secure the valve in the open position was to perform a 360* weld around the unthreaded portion of the 1-MOV-644 stem and the valve yoke.
Calculations were performed to determine the design margin available using the weld. The initial results indicated a 40% margin, however, more detailed calculations were performed incorporating explicit modelling of dead weight, thermal, seismic, and packing loads resulted in a 22% margin. BG&E also noted that eeisting welding procedures were used and the finished weld was subjected to dye penetrant and visual inspections.
The NRC staff has determined that, the use of a 360* weld between the valve stem and valve yoke supported by the detailed analysis summarized above and the subsequent weld inspections is an acceptable means of securing 1-MOV-644 in the open position.
The weld, which was completed on December 28, 1991, was subsequently visually inspected after reaching normal operating conditions on December 30, 1991. This additional visual inspection of the weld at normal operating conditions indicates that the thermal stresses during operation have not resulted in any damage to tne weld. BG&E has also indicated that all necessary procedures are in place to operate 1-MOV-644 in the welded open position including the actions necessary for depressurizing the SIT nitrogen pretsure during a unit shutdown.
Therefore, the NRC staff has concluded that the valve,1-MOV-644 is adequately secured in the open position and that subsequent verification is unnecessary until the valve is repaired or replaced during the opcoming refueling outage 10.
3.0 STATEMENT OF EXIGENT CIRCUMSTANCES BG&E states that exigent circu'nstances pursuant to 10 CFR 50.91 exist with respect to the need for consideration of the proposed amendment. The need for this change could not have been foreseen in that it resulted from corrective maintenance activities being performed during an unplanned outage to repair a leaking check valve. Application for an exigent amendment was made as soon as possible following the determination of the appropriate course of action.
personnel are required to enter the containment during power (Mode 1) operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to verify that 1-MOV-644 is in the open position, even though it i
is welded open, to meet the current TS. This results in unnecessary radiation exposure with no corresponding safety benefit and is inconsistent with the l
objectives of maintaining occupational radiation exposures to as low as reasonably l
achievable (ALARA). As previously noted, the December 31, 1991, exigent amendment request also requested a TWOC while the TS amendment request is being reviewed by the NRC.
By NRC letter dated December 31, 1991, the NRC granted the requested TWOC from the requirements of TS 4.5.1.a.2 for 1-MOV-644 to be in effect until the NRC staff completes its review for the requested exigent amendment.
Based on the above, the NRC staff has determined that the licensee has used best efforts to make a timely application and that exigent circumstance are present which warrant processing the requested amendment pursuant to 10 CFR l
50.91(a)(6).
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4.0 FINAL NO SIGNIFICANT HAZARD CONSIDERATION The Comission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
(1) involve a significant increase in the probability or consequences of and accident previously evaluation; or (2) create the possibility of a new or different kind of accident from an accident previously evaluated; or (3) involved a significant reduction in a margin of safety.
The following evaluation, by the licensee and with which we agree, demonstrates that the proposed amendment does not involve a significant hazards consideration.
Operation of the Calvert Cliffs Nuclear Power Plant, Unit 1, in accordance with the proposed amendment will not involve a significant hazards consideration as defined in 10 CFR 50.92, since it does not:
1.
Would n7t involve a significant increase in the probability or consequences of an accident previously evaluated.
The open or closed position of Safety Injection Tank isolation valves are not considered an initiator for any accidents previously evaluated. Therefore, the probability of previously evaluated accidents would not be increased by the requested change.
Previously evaluated accident analyses assume that Safety Injection Tank isolation valves are open. The requested change eliminates the verification of that condition for one valve but the valve has been welded in the open position. Therefore,ithasbeen[ reasonably) assured that the valve will function as required during any previously analyzed accident and that there will be no increase in consequences due to the request;d change.
Therefore, this change would not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Would not create the possibility of a new or different type of accident from any accident previously evaluated.
Welding the valve stem to the valve yoke on 1-MOV-644 ensures that the valve will remain open during Modes 1, 2 and 3, thus eliminating the need for periodic verification of valve position in those Modes. The only new or different type of accident that could be created by failing to verify the isolation valve position would be the unknown closure of the valvo. However, this possibility is precluded by welding the valve in the open
)osition. This change in surveillance requirements does not affect tie design and tunction of the isolation valve, nor the operation of the isolation valve as the l
valve's design and function is to remain open in Modes 1, 2 and 3 and the valve is not allowed to be operated in those Modes.
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Therefore, [there is reasonable assurance that] the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluatad.
3.
Would not involve a significant reduction in a nargin of safety.
The margin of safety provided by this surveillance reauirement is the assurance that the isolation valve is open.
The need to verify t8at the valve is open has been eliminated by welding the valve in the open position. Therefore, [there is reasonable assurance that]
the proposed change would not involve a significant reduction in a margin of safety.
Based on the foregoirg, the Comission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore, the Comission has made a final determination that the proposed amendment does not involve a significant hazards consideration.
5.0 STATE CONSULTATION
In accordance with the Comission's regulations, the Maryland State official was notified of the proposed issuance of the amendment. The State official had no coments.
6.0 ENVIPONMENTAL CONSIDERATION The amendnent changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFP Part 20 and changes to the surveillance requirements. The NRC staff has determined that the amendaent involves no significant increase in th. 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 cumuletive occupational radiation exposure. The Comission has previously issued a proposed finding-that the amendment involves no significant hazards consideration, and there has been no pubite comment on such finding (57 FR 938). Accordingly, the' amendment meets the eli forth in 10'CFR 51.22(c)(9)gibility criteria for categorical exclusion set Pursuant to 10 CFR 51.22(b)-no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
7.0 CONCLUSION
The Comission has concluded}e assurance that the health based on the considerations discussed above, that:
(1) there is reasonab public will not be endangered by operation in the proposed manner, 2) such activities will be conducted in compliance with -the Comission's regulations, and (3) the issuance of the amendment will not-be inimical to the common d u.nse and security or to the health and safety of the public.
Principal Contributor:
Daniel Mcdonald Date: January 29, 1992
I Mr. G. C. Creel January 29, 1992 A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Conrnission's next regular biweekly Federal Register notice.
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Sincerely, Daniel 0. Mcdonald, Senior Project Manager y-Project Directorate I-1 Division of Re3ctor Pro.iects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Amencment No. 167to DPR-53 2.
Safety Evaluation cc w/ enclosures:
See next page
______m____
tir. G. C. Creel January 29, 1992 A copy of the related Safety Evaluation is enclosed. A Notice of 1ssuance will be included in the Commission's next regular biweekly federal Register notice.
Sincerely, Original Sigrad By:
Daniel G. Mcdonald, Senior Project flanager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Enclosures:
1.
Amenoment No.167 to DPR-53 2.
Safety Evaluation cc w/ enclosures:
See next page
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- i/d/92
- 01/d/92
- 01 /14/92
- 01/16/92
- 1 /a /92 lC
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DFFICIAERECORD COPY Document Name:
ISS EXIG AMEND
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