ML20195J713
| ML20195J713 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/19/1998 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20195J716 | List: |
| References | |
| NUDOCS 9811250051 | |
| Download: ML20195J713 (13) | |
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ia CHART,Es H. Crese Baltimore Gas and Electric Company Vice President Calven Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495 4 455 November 19,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 License Amendment Request: Revision to Service Water System Technical Soecification
REFERENCES:
(a)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated July 20,1998, License Amendment Request: Service Water Heat Exchangers Replacement (b)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 16,1997, License Amendment Request: Service Water Heat Exchangers Replacement (c)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated November 14, 1997, Response to Request for Additional Information - License Amendment Request: Service Water Heat Exchangers Replacement (d)
Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE),
dated February 10, 1998, Issuance of Amendment for Calvert Cliffs Nuclear Power Plant, Unit No. I (TAC No. M98784)
Pursuant to 10 CFR 50.90, Baltimore Gas and Electric Company (BGE) hereby requests an amendment to Operating License Nos. DPR-53 and DPR-69 to incorporate the changes described below into the Technical Specifications for Calvert Cliffs Units 1 and 2.
The proposed amendment revises Technical Specification 3.7.6, " Service Water (SRW) System" to allow operation of Calvert Cliffs with one SRW plate and frame heat exchanger (PHE) secured for maintenance or other reasons, and removing one containment air cooler (CAC) from service to enable the affected subsystem to remain operable. Specifically, the proposed change adds "One SRW heat
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exchanger inoperable" as a new condition for Limiting Condition for Operation (LCO) 3.7.6.
The required actions for the new condition are to secure one CAC within one hour and restore the heat exchanger to operable condition within 7 days, or be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This limits the effect of one inoperable PHE to only one containment cooling train made inoperable by the 9811250051'981119' PDR ADOCK 05000317 P
PD_R
Document Control Desk 1
November 19,1998 i
Page 2 f
PHE. Consequently, the new action statement introduced in the SRW LCO for an inoperable PHE is similar to the one that already exists in the CAC LCO for one inoperable containment cooling train.
The proposed change is facilitated by BGE's ongoing project that is replacing each shell and tube SRW heat exchanger with a pair of new PHEs having increased thermal performance capability. Calvert Cliffs Unit 1 SRW heat exchangers were replaced during the 1998 refueling outage, and the replacement for F
Unit 2 is scheduled to take place during the 1999 refueling outage (Reference a). Hence, the proposed change will not be applicable to Unit 2 until the completion of the replacement activity in 1999.
The new SRW System configuration allows independent isolation of each PHE, which permits one PHE j
to remain in service while the other PHE on the subsystem is isolated for maintenance or other reasons.
Since a single PHE cannot remove the full accident heat load associated with the SRW subsystem, the current licensing basis requires the affected SRW subsystem to be declared inoperable while a single PHE is isolated (see Attachment 1 for a detailed discussion). Consequently, all associated safety-related equipment is removed from service. However, isolating one CAC associated with the isolated PHE will reduce the maximum accident heat load on the system sufficiently to ensure that SRW supply
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temperature would be maintained within design limits. This would allow continued operation of the I
affected cooling weer subsystems and all loads cooled by these subsystems. This mode of SRW System operation was originally proposed in Reference (b) and was denied by the NRC (References e and d) because a condition for an inoperable PHE and the associated required actions were not specifically addressed in the Technical Specification for the SRW System.
The proposed revision will provide greater availability of safety-related equipment during PHE maintenance or other activities. It ensures that the safety features provided by the saltwater and SRW, except for the isolated containment air cooler, are maintained, i.e., the availability of safety-related equipment required to mitigate the radiological consequences of an accident described in the Updated j
Final Safety Analysis Report is enhanced by the Dexibility provided by this technical specification revision.
I REQUESTED CHANGES Revise Technical Specification 3.7.6 as shown in the marked-up Technical Specification page in
. Attachment (3).
SCHEDULE As mentioned above, our plan calls for the replacement of Unit 2 SRW heat exchangers during the 1999 refueling outage, which is scheduled to begin on March 12,1999, and end on April 26,1999. To take full advantage of the new SRW heat exchanger configuration for both Units, we request that you review
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and approve our application by April 26,1999.
ASSESSMENT AND REVIEW We have evaluated the significant hazards considerations associated with this proposed modification, as required by 10 CFR 50.92, and have determined that there are none (see Attachment 2 for a complete discussion). We have also determined that operation with the proposed modification will not result in i
any significant change in the types or significant increases in the amounts of any effluents that may be J
released offsite, and no significant increases in individual or cumulative occupational radiation exposure.
Therefore, the proposed amendment is eligible for categorical exclusion as set forth in 1
Document Control Desk November 19,1998 Page 3 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed modification. The Plant Operations and Safety Review Committee and the Offsite Safety Review Committee have reviewed this proposed modification and concur that operation with the proposed modification will not result in an undue risk to the health and safety of the public.
Should you have questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours, STATE OF MARYLAND
- TO WIT:
COUNTY OF CALVERT I, Charles H. Cruse,.being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this License Amendment Request on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my persona 1 knowledge, they are based upon information provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to j
be reliable.
,n
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Subscrp' ed and worn before me a Notary Public in nd for the State of Maryland and County of
( 'al
,this / h day of '
.1998.
WITNESS my Hand and Notarial Seal:
Notary Public j
! dM My Commission Expires:
hte' GT/gt/ dim Attachments: (1)
Summary Description and Safety Analysis (2)
Determination of Significant Hazards (3)
Technical Specification Marked-up Page cc:
' R. S. Fleishman, Esquire H. J. Miller, NRC
]
. J. E. Silberg, Esquire Resident Inspector, NRC S. S. Bajwa, NRC R. I. McLean, DNR -
A. W. Dromerick, NRC J. H. Walter, PSC
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ATTACHMENT (1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS Baltimore Gas & Electric Company Docket Nos. 50-317 & 50-318 November 19,1998
ATTACHMENT (1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS BACKGIdOUND By Reference (1), Baltimore Gas and Electric Company (BGE) submitted a license amendment request for Calvert Cliffs Unit I to implement a modification that constituted an unreviewed safety question.
The modification involved replacing the two shell and tube service water (SRW) heat exchangers with four new plate and frame heat exchangers (PIIEs) having increased thermal performance capability.
l The new SRW System configuration, with two PHE per train, allows independent isolation of each PHE, which permits one PHE to remain in service while the other PHE on a given train is isolated for maintenance or other reasons. Therefore, as part of that submittal, BGE introduced a new operational mode for the SRW System," Normal Operation with One PHE Secured," which limits the effect of one inoperable PHE to only one containment cooling train made inoperable by the PHE. A single PHE cannot remove the full accident heat load associated with the SRW subsystem. However, isolating one containment air cooler (CAC) associated with the isolated PHE will reduce the maximum accident heat load on the system sufficiently to ensure that SRW supply temperature would be maintained within design limits. This would allow continued operation of the affected cooling water subsystems and all loads cooled by these subsystems. The NRC approved the license amendment request, but denied this mode of SRW System operation (References 2 and 3) because a condition for an inoperable PHE and the associated required actions were not specifically addressed in the Technical Specification for the SRW System. By Reference (4), BGE submitted a similar license amendment request for Unit 2 and was approved by Reference (5).
This license amendment request is being submitted to realize the full benefit of the new SRW System configuration by revising the Technical Specification to limit the effect of one inoperable PHE to only one containment cooling train made inoperable by the PHE.
DESCRIPTION OF AFFECTED SYSTEMS A.
SRW SYSTEM The SRW System is a closed loop system that uses plant demineralized water treated with a corrosion inhibitor. The system removes heat from various turbine plant components, a blowdown recovery heat exchanger, CACs, a spent fuel pool cooling heat exchanger, and diesel generator (DG) heat exchangers.
The system is divided into two subsystems in the Auxiliary Building to provide adequate redundancy in the design basis accidents. The non-safety-related portion of the system, located in the Turbine Building, is divided into two separate supply headers, but combine downstream for Unit 1 and upstream for Unit 2 of the Turbine Building loads into a common return header. The SRW Turbine Building piping is automatically isolated upon Safety injection Actuation Signal (SIAS). While in normal operation, both subsystems are usually in service and are independent to the degree necessary to assure the safe operation and shut down of the plant, assuming a single failure. The SRW supply temperature is maintained at or below 95 F. During shutdown, operation of the SRW System is essentially the same as during normal operation, except that the heat loads are reduced. During a loss-of-coolant accident (LOCA), each subsystem supplies two CACs and one DG (except 11 SRW which has no DG). During this event, the SRW System design temperature initially increases to 115 F (see response to Question Nos. 2 and 3 in Reference 2), and subsequently decreases belcw 105 F within 35 minutes. The SRW System supply temperature is maintained at or below 105 F for the remainder of the event.
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MHMENT (1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS The design safety function of the SRW System is to supply cooling water to the CACs to support cooldown of containment, and to DGs (except No. I A DG) to ensure continued reliable operation of the DGs as an emergency power supply.
B.
SW SYSTEM The SW System consists of two subsystems. Each subsystem provides SW to heat exchangers cooling the SRW System, Component Cooling (CC) System and the Emergency Core Cooling System (ECCS) pump room. These heat exchangers transfer heat from the associated systems to the Chesapeake Bay. The SW System has three pumps. The pumps provide the driving head to move SW from the Intake Structure, through the system, and back to the circulating water discharge conduits.
During normal operation, both subsystems are in operation with one pump running on each header, and a third pump in standby. If needed, the standby pump can be lined-up to either supply header.
The SW flow through the SRW and CC heat exchangers is throttled to provide sufficient cooling to the heat exchangers, while maintaining total subsystem How of a value that satisfies both pump runout, and minimum flow requirements.
Operation following a LOCA has two phases - pre-and post-Recirculation Actuation Signal (RAS). One subsystem can satisfy the design heat removal requirements during both phases of the accident. After a LOCA and before a RAS, saltwater flow to the CC heat exchangers is automatically isolated on a SIAS. During this phase, the ECCS pump room air coolers may also be cooled by SW if the ECCS pump room temperature exceeds its preset limit. Upon initiation of a RAS, the SW isolation valves on the CC heat exchanger return to their pre-accident positions and the operator can throttle flow to maintain CC temperatures. The SRW heat exchangers and the ECCS pump room air coolers continue to operate during this phase. If the PHE SW outlet control valves are fully open prior to the accident, they will be placed in automatic, by the operator, after a RAS.
Should a rupture or blockage occur in the common SW discharge piping downstream of the heat exchangers and air coolers, an alternate flow path may be employed so that the function of the components on the No.12(22) SW Subsystem will not be impaired. This alternate flow path may also be used to support maintenance during Modes 5 (Cold Shutdown),6 (Refueling), and defueled.
C.
CACs The CACs e.re provided with SRW to remove heat from the Containment during normal plant operation and in the event of a LOCA or main steam line break. Flow through the coolers is controlled by air-operated control valves located in the containment penetration rooms. The CAC SRW inlet control valves are normally open with the valve operator vented to atmosphere. During a LOCA, SRW flow to each CAC is controlled by an air-operated control valve. Upon receipt of a SIAS, the control valve is automatically throttled to the predetermined flow setpoint or throttle position. The resulting flow is based on satisfying the post-LOCA heat removal requirements while limiting the system heat load and preventing two phase flow. Upon receipt of a RAS, the valve will return to its fully open position.
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ATTACHMENT (1)
SUMMARY
DESCRHTION AND SAFETY ANALYSIS DESCRIPTION OF SYSTEM OPERATION A.
NORMAL OPERATION WITH TWO PHEs OPERATING ON EACH SW SUBSYSTEM Normally, SW flow to each SRW PHE will be maintained by a flow indicating controller at the pre-established setpoint. The SW strainers will flush automatically at the selected frequency or if an abnormally high strainer differential pressure is experienced prior to the normal flush. The SW bypass lines will be throttled by pressure indicating controllers to help maintain SW header pressure within the prescribed band.
During normal system operation with both SW headers in service, both Nos.11(21) and 12(22) trains can remove the design heat !oad with SW temperatures up to 90 F.
At higher bay temperatures or system heat loads, e.g., while testing the DG, the outlet control valves may be fully opened to maintain SRW temperatures. Increasing SW flow to the PHEs will increase their heat removal capachy. The valves can be fully opened by moving the associated hand switch from
" AUTO" to "OPEN." This mode will support plant operation in all modes and conditions (except post-RA3) with SW temperatures up to 90 F.
B.
NORMAL OPERATIONS WITH ONE SW SUBSYSTEM SECURED, TWO PHEs ON THE OPERATING SUBSYSTEM Either SW subsystem can be secured for maintenance by cross-connecting the SRW System to the remaining pair of PHEs. The two operating PHEs can remove the heat load during normal operations or a design basis accident with SW temperatures up to 90 F. In this configuration, the affected SW, SRE and CC subsystems and all supported heat loads are considered to be operable.
C.
PROPOSED NORMAL OPERATION WITH ONE PHE SECURED Valves are provided in the SW and SRW Systems to allow isolation of a single strainer-PHE combination while continuing to operate the other PHE, the CC heat exchanger, and the ECCS pump room air cooler on the affected SW subsystem. Under most conditior.2, a single PHE cannot remove the full LOCA heat load while maintaining SRW temperature within its design limits.
However, if one CAC on the affected subsystem is isolated and removed from operation, the single PHE can remove the remaining accident heat load on that subsystem. This would allow the DG (not applicable on No. I1 SRW Subsystem), the remaining CAC, the CC heat euhangen, and the ECCS pump room cooler on the affected subsystem to remain operable while the one PHE is out-of-l service. The other SW and SRW subsystems would still have two operable PHEs and would satisfy j
the assumptions in the acci6nt analysis.
The SRW heat loads during normal operations differ significantly from those experienced during accident and are not equally divided between the two subsystems. However, as discussed above, two PHEs can remove the full normal heat load from the system. Some system realignment may be necessary at higher Bay temperatures to balance the heat load between the operating heat exchangers.
The current licensing basis, however, requires the affected SRW subsystem to be declarul inoperable while a single PHE is isolated (Reference 3). Consequently, all associated safety-related components are removed from service and the Technical Specification Limiting Condition for Operation (LCO) for each affected component must be met. To permit the proposed mode of 3
e ATTACIIMENT (1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS operation, we are proposing a Technical Specification revision (see Attachment 3) which will limit
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the effect of one inoperable PHE to only one containment cooling train made inoperable by the PHE. The proposed revision to the Technical Specifications adds an action statement in the SRW LCO similar to the one that already exists in the CAC LCO. The following table summarizes operation with one PHE secured under the current licensing bas!= and under the proposed license amend.wnt.
Under Current Licensing Basis Under the Proposed License Amendment 1 PHE removed from service.
1 PHE removed from service.
The associated SRW subsystem is declared The unaffected PHE on the subsystem remains inoperable.
Action Statement for an operable. Action Statements for an inoperable inoperable SRW subsystem is entered.
PHE is entered.
Completion time - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion time - 7 days l
Both CACs on the associated containment The associated CAC is removed from service.
cooling train are declared inoperable. Action The unaffecteo CAC on the associated Statement for an inoperable containment containment cooling train remains operable.
cooling train is entered.
Action Statement for an inoperable containment c 1 ng train is entered.
Completion time - 7 days Completion time - 7 days All other associated safety-related All other associated safety-related components components are declared inoperable and the remain operable.
Technical Specification LCO for each affected equipment must be met. (Example: ECCS Subsystem - Completion time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; CC Subsystem - Completion time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.)
SAFETY ANALYS[S The proposed activity does not involve any physical modifications or change to the operating limits, i.e., pressure and temperature, of the existing SW or SRW Systems. Isolating one CAC while the PHE is out-of-service will reduce the SRW subsystem heat load sufficiently to ensure that the remaining PHE can maintain SRW subsystem temperatures within their design limits. This will allow for continued l
operation of the affected DG, one CAC on the affected subsystem, and the associated CC and ECCS l
cooling train. The existing Technical Specifications allow isolation of individual SW or SRW l
components without rendering the subsystem inoperable. The proposed Technical %ecification revision l
would allow the isolation of a PHE and an associated CAC while maintaining the remainder of the j
subsystem components and the SW system operable. All other equipment would continue to operate as currently described in the licensing basis.
Under both the current licensing basis and the proposed license amendment request, the Technical Specification LCO for an inoperable containment cooling train is entered (see summary table above).
Therefore, the proposed isolation of one CAC to enable the remaining subsystems to remain operational, has no operational impact on the affected containment cooling train subsystem. The proposed seven-day completion time to restore the inoperable PHE to operable status is consistent with the completion time for an inoperable containment cooling train, which is the only subsystem that is declared inoperable 4
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i ATTACHMENT (1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS under the proposed license amendment. The seven-day completion time was developed taking into account the redundant heat removal capabilitas afforded by combinations of the containment spray and cooling systems, and the low probability of a design basis accident occurring during this period. The proposed one-hour completion time to isolate a CAC for an inoperable PHE is based on low probability of a design basis accident occurring during this period.
The consequences of a single failure of the SW or SRW piping or components are identical to those already considered in the Updated Final Safety Analysis Report (UFSAR). Any such failure would be enveloped by the UFSAR accident analyses, which assume a single failure results in the loss of an entire SW and SRW subsystem. The remaining subsystem is capable of mitigating the effects of the accident.
The proposed Technical Specification change would maintain the design redundancy feature of all safety-related equipment except the affected CAC train during heat exchanger maintenance or other activities. By rwiding the second DG and the CC and ECCS cooling equipment, the consequences of many malfunctions during the period required for PHE maintenance will be reduced since redundant components will still be available.
The current licensing basis requires that when one PHE is taken out-of-service, the associated SRW subsystem be considered inoperable. In this condition, the remaining operable SRW subsystem is adequate to perform the system heat removal functions. However, the overall reliability of the system is reduced because a single failure in the SRW System could result in loss of SRW function. This, in turn, could result in the loss of the one ; perable DG and all DG-backed electrical loads. In the proposed configuration, during PHE maint 3nce, the containment cooling function would still require the unaffected SRW subsystem; a single C%re on that subsystem could result in its loss. The loss of the operable containment cooling train can result in increased containment temperatures, which in turn would increase the heat load on the remaining SRW subsystem. The single operable PHE on that subsystem might not be capable of maintaining SRW temperature within its design limits with the increased containment temperature. This could result in the loss of the second DG and the associated SRW subsystem. The end result of this malfunction, the loss of all SRW functions, would be identical to the consequences of a malfunction of the operable SRW subsystem while operating under the existing Technical Specification with the second system inoperable However, many malfunctions that could result in the complete loss of a safety function while one SRW l
tram is inoperable would not necessarily result in the loss of the associated function under the proposed operating configuration. For example, during PHE maintenance, the associated DG is currently j
l inoperable. This eliminates the redundancy of many safety-related components in the event of a loss of l
cifsite power. With implementation of the proposed technical specification change, the affected DG would remain operable. In the event of a malfunction of any component powered from one DG, the redundant component powered from the other DG would still be available.
CONCLUSION The proposed revision will provide greater availability of safety-related equipment during PHE maintenance or other activities. It ensures that the safety features provided by the SW and SRW, except for the isolated CAC, are maintained, i.e., the availabil;ty of safety-related equipment required to mitigate the radiological consequences of an accident described in the UFSAR is enhanced by the flexibility provided by this Technical Specification revision.
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4 ATTACHMENT {1)
SUMMARY
DESCRIPTION AND SAFETY ANALYSIS REFERENCES (1)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 16,1997, License Amendment Request: Service Water Heat Exchangers Replacement (2)
Letter from Mr. C.
H. Cruse (BGE) to NRC Document Control Desk, dated November 14,1997, Response to Request for Additional Information -- License Amendment Request: Service Water Heat Exchangers Replacement (3)
Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated February 10,1998, Issuance of Amendment for Calvert Cliffs Nuclear Power Plant, Unit No.1 (TAC No.
M98784)
(4)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated July 20,1998, License Amendment Request: Service Water Heat Exchangers Replacement (5)
Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated November 5,1998, Issuance of Amendment for Calvert Cliffs Nuclear Power Plant, Unit No. 2 (TAC No.
MA2332) l i
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l ATTACHMENT (2) i l
i DETERMINATION OF SIGNIFICANT HAZARDS i
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Baltimore Gas & Electric Company Docket Nos. 50-317 & 50-318 l
November 19,1998
l ATTACHMENT (2) q DETERMINATION OF SIGNIFICANT HAZARDS The proposed amendment revises Technical Specification 3.7.6, " Service Water (SRW) System" to allow operation of Calvert Cliffs with one SRW plate and frame heat exchanger (PHE) secured for maintenance or other reasons, and isolating one containment air cooler (CAC) to enable the affected subsystem to remain operable. Specifically, the proposed change adds "One SRW heat exchanger inoperable" as a new condition for Limiting Condition for Operation (LCO) 3.7.6. The required action for the new condition is to secure one CAC and restore the heat exchanger to operable condition within 7 days, or be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This limits the effect of one inoperable PHE to only one containment cooling train made inoperable by the PHE by introducing an action statement in the SRW LCO similar to the one that already exists in the CAC LCO.
The proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendments:
l.
Would not involve a significant increase in the probability or consequences of an accident previously evalaated.
None of the systems associated with the proposed revision to the Calvert Cliffs Technical Speci6 cations are accident initiators. The Saltwater (SW) and SRW Systems are used to mitigate the effects of accidents analyzed in the Updated Final Safety Analysis Report (UFSAR).
The SW and SRW Systems provide cooling to safety-related equipment following an accident.
The CACs are provided with SRW to remove heat from the Containment in the event of an accident. They support accident mitigation functions; therefore, the proposed modification does not increase the probability of an accident previously evaluated.
The proposed revision will provide greater availability of safety-related equipment during PHE maintenance activities. It ensures that the safety features provided by the SW and SRW, except for the isolated CAC, are maintained, i.e., the availability of safety-related equipment required to mitigate the radiological consequences of an accident described in the UFSAR is enhanced by the flexibility provided by this Technical Specification revision.
Furthermore, the proposed revision will not change, degrade, or prevent actions described or assumed in any accident described in the UFSAR. Se proposed activity will not alter any assumptions previously made in evaluating the raciological consequences of any accident described in the UFSAR.
Therefore, the proposed modification does 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 di[ferent 13pe of accident from any accident p eviouslyerufu sted.
None of the systems associated with this modification are identified as accident initiators in the UFSAR. The SW and SRW Systems and the CACs are used to mitigate the effects of accidents analyzed in the UFSAR. None of the functions required of these systems have been changed by the proposed revision to the Technical Specifications. This activity does not modify any system, structure, or component such that it could become accident initictor, as opposed to its current role as an accident mitigator.
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ATTACIIMENT m l
DETERMINATION OF SIGNIFICANT IIAZARDS l
l Therefore, the proposed change does not create the oossibility of a new or different type of accident from any accident previously evaluated.
3.
Would not involve a sigmficant reduction in a margin ofsafety.
i The safety design basis for the SW and SRW Systems is the availability of sufficient cooling l
capacity to ensure continued operation of equipment during normal and accident conditions. The i
redundant cooling capacity of these systems, assuming a single failure, is consistent with assumptions used in the accident analysis.
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With one SRW subsystem inoperable, the remaining SRW subsystem is adequate to perform the heat removal function. However, the reliability is reduced because a single failure in the operable SRW subsystem could result in loss of SRW function. The proposed change will allow continued operation of some SRW-cooled components while a PHE is being out-of-service. The second SRW subsystem will still be available to perform the SRW function. In addition, the reliability of many diesel generator-backed components will be improved since the second diesel generator will remain operable while in this action statement.
During a design basis accident, a minimum of one containment cooling train (two of the four CACs) and one containment spray train, is required to maintain the containment peak pressure and temperature, below the design limits.
Under the existing Technical Specification j
requirement, with one containment cooling train inoperable, the inoperable containment cooling train must be returned to operable status within seven days. He remaining operable containment l
spray and cooling units provide iodine removal capabilities and are capabic of removing at least i
100% of the heat removal needs after an accident. The seven-day completion time was developed taking into account the redundant heat removal capabilities afforded by combinations of the containment spray and cooling systems, and the low probability of a design basis accident occurring during this period. He proposed change to Technical Specification 3.7.6 would allow three CACs to remain operable during maintenance on a PIIE, instead of the two that are maintained under the current Technical Specification requirement.
For the above reasons, the margin of safety has been preserved, and in some cases increased, by the proposed revision to the Technical Specifications.
Therefore, this proposed modification does not significantly reduce the margin of safety, i
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