3F1006-02, License Amendment Request 293, Revision 0 Revised Improved Technical Specification (ITS) 3.7.7, Nuclear Services Closed Cycle Cooling Water (SW) System
| ML062920567 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/11/2006 |
| From: | Young D Progress Energy Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F1006-02 | |
| Download: ML062920567 (31) | |
Text
Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.90 October 11, 2006 3F1006-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Crystal River Unit 3 - License Amendment Request #293, Revision 0 Revised Improved Technical Specifications (ITS) 3.7.7, Nuclear Services Closed Cycle Cooling Water (SW) System
Dear Sir:
Florida Power Corporation, doing business as Progress Energy Florida, Inc. (PEF), hereby submits License Amendment Request (LAR) #293, Revision 0, which requests a change to the Crystal River Unit 3 (CR-3) Facility Operating License in accordance with 10 CFR 50.90.
This LAR is proposing to revise CR-3 Improved Technical Specifications (ITS) 3.7.7, Nuclear Services Closed Cycle Cooling Water (SW) System. The proposed revision will implement a more conservative Technical Specification for the system.
This letter is being submitted pursuant to the guidelines of NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety." The CR-3 ITS Bases (Attachment E) and plant procedure (OP-408, Nuclear Services Cooling System) have been revised to impose administrative controls to assure CR-3 is operated in a manner consistent with the plant safety analysis.
This letter establishes no new regulatory commitments.
The CR-3 Plant Nuclear Safety Committee has reviewed this request and recommended it for approval.
Progress Energy Florida, Inc.
Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428
U.S. Nuclear Regulatory Commission Page 2 of 3 3F1006-02 If you have any questions regarding this submittal, please contact Mr. Paul Infanger, Supervisor, Licensing and Regulatory Programs at (352) 563-4796.
Sincerely, Dale E. Young Vice President Crystal River Nuclear Plant DEY/seb Attachments:
A. Background, Description of the Proposed License Amendment Request, Technical Analysis B. Regulatory Analysis (No Significant Hazards Consideration Determination, Applicable Regulatory Requirements and Environmental Impact Evaluation)
C. Proposed Revised Improved Technical Specifications Pages - Strikeout/Shadowed Format D. Proposed Revised Improved Technical Specifications Pages - Revision Bar Format E. Existing Improved Technical Specifications Bases - Revision Bar Format F. Proposed Revised Improved Technical Specifications Bases - Strikeout/Shadowed Format xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector
U.S. Nuclear Regulatory Commission Page 3 of 3 3F1006-02 STATE OF FLORIDA COUNTY OF CITRUS Dale E. Young states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this 11 day of Oct4___, 2006, by Dale E. Young.
Signature of Notary Prh+i_
State of Florida FORK".. EMD 34UM,0 (Print, type, or stamp Commissioned Name of Notary Public)
Personally Produced Known -OR- Identification
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT A LICENSE AMENDMENT REQUEST #293, REVISION 0 Background, Description of the Proposed License Amendment Request, Technical Analysis
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 1 of 7
Background
The Nuclear Services Closed Cycle Cooling Water (SW) system removes process and operating heat from safety-related components during normal operations as well as during a transient or accident conditions. As shown in Figure 1, the system consists primarily of one normal service pump, two emergency pumps, a surge tank pressurized with nitrogen for volume and pressure control and four heat exchangers which reject heat to the Ultimate Heat Sink (UHS) via the Nuclear Services Seawater (RW) system. The two emergency pumps, in conjunction with adequate heat removal ability by the heat exchangers, provide the necessary capability for cooling the containment fan assembly cooling coils and fan motors, spent fuel pool, SW and RW pump motors, and other equipment which must function following an accident.
This proposed amendment will implement a more conservative Improved Technical Specification (ITS) 3.7.7 SW system. The current Crystal River Unit 3 (CR-3) ITS Limiting Condition for Operation (LCO) 3.7.7 requires two emergency SW pumps and three SW heat exchangers for the system to be OPERABLE. The current ITS Action 3.7.7 Condition A, allows the plant to operate for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with one emergency SW pump inoperable or one required SW heat exchanger inoperable. The ITS Bases state that the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action is appropriate for the loss of redundancy.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Technical Specification requirement was approved as part of CR-3's initial license.
However, a CR-3 calculation (M97-0133) and informal case studies (Action Requests 119442, 131522) have revealed the loss of a SW heat exchanger to be not a loss of redundancy, but a potential loss of safety function. The loss of safety function is based on the inability of two SW heat exchangers, at certain UHS temperatures, to perform their function of adequately removing heat from safety-related components during worst case post accident conditions. The proposed revision to ITS 3.7.7 will allow only an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ACTION Completion Time for a condition with one required SW heat exchanger inoperable.
As per guidance in NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," the CR-3 ITS Bases, plant procedure (OP-408, Nuclear Services Cooling System) and Design Basis Documents have been revised to impose administrative controls to prevent operation with less than three SW heat exchangers.
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 2 of 7 RB Figure 1. Nuclear Services Closed Cycle Cooling Water (SW) System
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 3 of 7 Description of the Proposed License Amendment Request The proposed License Amendment Request (LAR) #293, Revision 0, will revise the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) 3.7.7, "Nuclear Services Closed Cycle Cooling Water (SW) System."
The current ITS:
CONDITION REQUIRED ACTION COMPLETION TIME A. One emergency SW A.1 Restore SW system to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pump inoperable. OPERABLE status OR One required SW heat exchanger inoperable.
B. Required Action and B.1 Be in Mode 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in Mode 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> The requested revision:
CONDITION REQUIRED ACTION COMPLETION TIME A. One emergency SW A.1 Restore SW system to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pump inoperable. OPERABLE status B. One required SW heat B.1 Restore SW system to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exchanger inoperable. OPERABLE status C. Required Action and C.1 Be in Mode 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND C.2 Be in Mode 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> The requested change will remove allowance from the CR-3 Technical Specifications that may permit a potential loss of safety function for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 4 of 7 Technical Analysis The current Crystal River Unit 3 (CR-3) Improved Technical Specification (ITS) LCO 3.7.7 requires two OPERABLE emergency Nuclear Services Closed Cycle Cooling (SW) system pumps and three OPERABLE SW heat exchangers. If there is a loss of one emergency SW pump or one SW heat exchanger, the current allowed ACTION Completion Time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A CR-3 calculation and informal case studies have found the current ITS 3.7.7 CONDITION A to be insufficient to assure plant safety under all conditions. At a design basis Ultimate Heat Sink (UHS) temperature of 95°F, the Raw Water system can maintain the SW system below its maximum post accident temperature of 110°F with three heat exchangers in service. Each SW heat exchanger is rated at approximately one third of the total required heat transfer rate for emergency operations without significant additional margin. Therefore, it can be inferred that at a UHS temperature of 95°F, two SW heat exchangers are not able to remove adequate process and operating heat from required components during the limiting Design Basis Accident. If the SW system exceeds 110F, equipment required for accident mitigation, including the High Pressure Injection system and control complex chillers may not be able to perform their required safety functions.
Analysis to determine SW heat loads during a Large Break Loss of Coolant Accident (LBLOCA) has been developed with conservative assumptions. During a Design Basis Accident, the SW system must be capable of removing heat from safety-related components assuming the worst-case single active failure occurs. The duties of the SW system include removing normal spent fuel pool heat load, peak post-LOCA heat from the Reactor Building via the containment fan assembly cooling coils and fan motors and cooling to safety related pumps and motors. The most limiting analysis is the overheating of the SW system under accident conditions.
When determining the ability of the SW system to perform its function during the Design Basis Accident, it is assumed the single active failure is the loss of one Raw Water (RW) pump. This failure will result in a reduction in the heat transfer rate from SW to the UHS, negatively impacting the cooling ability of SW while maintaining the most conservatively large heat load. In this case, the loss of an entire Emergency Core Cooling (ECCS) system train is less conservative because maintaining the second SW pump, and other equipment in service, maximizes the heat transfer into the SW system resulting in the highest SW temperature. Conservative assumptions are made when determining the heat input to SW by vital loads such as the Reactor Building Cooling Unit (RBCU) and spent fuel pool cooling system heat exchangers. The spent fuel pool's heat load contribution is based on a conservatively high fuel assembly discharge (84 assemblies) pool load 26 days after shutdown. This is considered a bounding heat load for future outages since typically approximately 73 fuel assemblies are off loaded during an outage. In addition, no credit is given for the decay of this heat load over time. To determine the RBCU heat load to SW, the RBCU SW inlet temperature is assumed to be 100°F. At this assumed temperature, the heat transfer rate from the Reactor Building to SW is greater than at the maximum SW design temperature of 110OF during a Design Basis Accident.
The proposed amendment will limit the time the safety function of the SW system can be compromised. The proposed revised ITS ACTION Completion Time for one inoperable SW heat exchanger will reduce the time the plant can operate with only two OPERABLE SW heat exchangers from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This timeframe is more comparable to other ITS Conditions where accident analysis assumptions are compromised. ACTION conditions allowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are typically utilized for a loss of one train in a system having two trains. An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 5 of 7 is more appropriate for a condition where the safety function could be compromised. Other CR-3 ITS ACTIONS that permit 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when an analysis assumption limit is exceeded or a condition compromises safety function include:
3.2.5 ACTION A Power Peaking Factors (FQ(Z) not within limits) 3.5.4 ACTION A Borated Water Storage tank (BWST) (boron concentration or water temperature not within limits) 3.6.5 ACTION A Containment Air Temperature (air temperature not within limit) 3.7.2 ACTION A Main Steam Isolation Valves (MSIVs) (one or more MSIV inoperable on one steam generator) 3.7.3 ACTION C Main Feedwater Isolation Valves (MFIVs) (one or more MFW flow paths with two MFIVs inoperable)
The request for a revision to the ACTION Completion Time is comparable to the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> action limits listed above. The action to restore water temperature or boron concentration of the Borated Water Storage Tank (BWST) to within limits is ITS 3.5.4 ACTION A. In the case of the BWST, the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ACTION Completion Time is based on engineering judgment considering the time required to correct the boron concentration or water temperature and prepare it for use. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limit to restore the SW heat exchangers to OPERABLE is acceptable, based on engineering judgment, considering the time required to isolate a SW heat exchanger, perform work to return it to compliance and restore it to service under most circumstances. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> timeframe is also adequate to perform cleaning or a number of other maintenance activities that could restore the SW heat exchangers to OPERABLE status. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ACTION Completion Time is comparable to the expectations of LCO 3.0.3 which requires the plant to be placed in Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after initiating the ACTION and is preferable to reactor shutdown.
The proposed 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time limit for restoration of an inoperable SW heat exchanger is acceptable based on the low probability of an accident that would deliver a high heat load to containment occurring coincident with the loss of a passive component, the SW heat exchanger, and conservatively high UHS temperature of 95*F. The probability of failure of passive components is low since no operator or control actions are required in order for them to perform their function.
CR-3's Licensing Basis does not assume the single failure of passive components such as the SW heat exchangers.
Emergency Safeguards (ES) and available Operator actions provide defense in depth to prevent overheating of the SW system. Engineered Safeguards actuation of Building Spray (BS) relieves the heat load input to SW by cooling the Reactor Building. Upon an ES signal, both trains of BS are actuated and work in conjunction with one RBCU to furnish Reactor Building cooling. The post-accident containment cooling analysis assumes either one of two trains of RBCUs and one of two trains of BS actuates OR both trains of BS actuate. Either option provides adequate containment cooling. Each train of BS is independent of the other train and the entire BS system is independent of RBCUs. If the SW system does not perform its design function, the RBCUs will not be able to perform their cooling function. In the limiting case (RW pump single failure), the two trains of BS will be available for cooling (BS pumps are cooled by Decay Heat Closed Cycle Cooling (DC)).
The BS system is sized to provide more than 100% of the required Reactor Building cooling with both trains in operation. Both Decay Heat (DH) pumps which provide Low Pressure Injection (LPI) are also cooled by DC. This supplies another means of cooling the Reactor Building should SW not be available.
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 6 of 7 CR-3 Abnormal Operating Procedure AP-330, Loss of Nuclear Service Cooling, directs operators to align alternate cooling for SW cooled components where available. Operator actions that can be performed to prevent overheating of the SW system include the ability to suspend Spent Fuel Pool Cooling and align two Makeup Pump (MUP-1A and MUP-IC) motors to DC instead of SW for cooling (one normally aligned to DC). Spent fuel pool cooling is one of the largest heat loads on the SW system. Spent fuel pool cooling can be suspended for certain time periods without causing damage to the fuel. MUP-IA and MUP-1C serve an important role as High Pressure Injection pumps. Control Complex Chillers can be aligned to the non-safety-related Secondary Cooling system if it is available. Also, the RBCUs can be aligned to the non-safety-related Industrial Cooling system. All of these actions are in current plant procedures. High Pressure Injection is actuated on an ES signal and serves to cool the reactor core.
As per guidance in NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," CR-3 ITS Bases, Design Basis Documents and a plant procedure (OP-408, Nuclear Services Cooling System) have been revised to impose administrative controls to prevent operation with less than three heat exchangers. B3.7.7 LCOs, APPLICABILITY and ACTIONS have been modified to clarify the need for three SW heat exchangers unless previously qualified by engineering analysis. The CR-3 ITS Bases revisions were fully implemented on August 7, 2006. Subsequent to the approval of this license amendment request, the CR-3 ITS Bases will be updated to reflect the revised ITS 3.7.7.
The LCO statement in the Bases has been expanded to make clear the reason for requiring three SW heat exchangers during normal operation. "Each heat exchanger is rated at one-third the total required heat transfer rate for both normal and emergency operations with an Ultimate Heat Sink temperature of 95°F. Therefore, normal operations will include three of four SW heat exchangers in service which will allow the fourth SW heat exchanger to be removed from the system for maintenance."
The APPLICABILITY statement in the Bases has been supplemented with an explanation of the reasoning for three out of four operating SW heat exchangers for normal and emergency operations.
"The number of heat exchangers required is based on meeting the design basis heat transfer rate for both normal and emergency operations.. .Therefore; normal operations will have three of the four heat exchangers in service..." This addition emphasizes the need for three heat exchangers for normal operation and ensures the plant is operated in a manner consistent with the plant safety analysis.
CR-3 plant procedure OP-408, Nuclear Services Cooling System, has been revised to include multiple Notes to emphasize the need for three SW heat exchangers (SWHIEs) for all operating conditions, "NOTE: At least three SWHEs must remain in service during Modes 1-4 to meet SW design bases under most UHS conditions. Refer to Technical Specifications 3.7.7 and 3.7.9, if required." The Note has been placed strategically in Section 4.11, "SW Heat Exchangers Operations (Continuous)." This Note serves as a reminder of the need for three SW heat exchangers for SW system operability. Prior to conducting maintenance, procedures systematically valve in the standby SW heat exchanger before removing another SW heat exchanger for routine maintenance. The maintenance procedures ensure three SW heat exchangers are in service at all times.
U.S. Nuclear Regulatory Commission Attachment A 3F1006-02 Page 7 of 7 In addition, Design Basis Documents were reviewed to ensure the clarity of the requirement for three SW heat exchangers. The CR-3 Final Safety Analysis Report (FSAR) Revision 30, Section 9.5.2.1.2.d states, "Four identical nuclear services heat exchangers are installed. These are sized to provide the emergency cooling requirements with three heat exchangers in operation and one in reserve." Also, under Section 9.5.2.1, "Three of the four nuclear service heat exchangers supply the full normal and emergency cooling requirements, with the fourth unit on reserve." Enhanced Design Basis Document (EDBD) Tab 6/11, for the Nuclear Services Closed Cycle Cooling Water System, Section 2.0 states, "Three of the four (33% capacity) heat exchangers are required to handle total heat load."
These administrative controls (revision of the ITS Bases, OP-408 and Design Basis Documents) provide assurance that CR-3 will operate within the constraints of the safety analysis until such time that this license amendment request is approved.
Finally, the impact of SW heat exchanger tube blockage on performance is considered. Tube blockage negatively impacts the SW system's ability to remove heat released to the Reactor Building. Regularly scheduled preventive maintenance, including cleaning of the tubes, is performed on the SW heat exchangers to remove debris blockage of the tubes and ensure safety function is maintained.
The proposed 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ACTION Completion Time is reasonable, considering conservative assumptions in the determination of maximum SW temperature following a LBLOCA, the low probability of an accident that would deliver a high heat load to containment occurring during the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time interval and available actions that provide defense in depth to aid in the mitigation of the SW heat load during a design basis accident.
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT B LICENSE AMENDMENT REQUEST #293, Revision 0 Regulatory Analysis No Significant Hazards Consideration Determination Applicable Regulatory Requirements Environmental Impact Evaluation
U.S. Nuclear Regulatory Commission Attachment B 3FI006-02 Page 1 of 2 No Significant Hazards Consideration Determnination The proposed License Amendment Request (LAR) #293, Revision 0, will revise the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) 3.7.7, "Nuclear Services Closed Cycle Cooling Water (SW) System." The amendment will revise the Allowed ACTION Completion Time for one inoperable SW heat exchanger from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 1. Does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The limiting design basis accident for CR-3 includes, as an assumption, adequate heat removal capability by the SW system. The amendment is being proposed to ensure the SW system performs its design basis function. Adequate heat removal is provided by three OPERABLE SW heat exchangers.
The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time will reduce the window that the plant can operate with only two SW heat exchangers before a shutdown is required. The proposed change does not increase the probability of an accident previously evaluated since the amendment is not a modification to plant systems, nor a change to plant operation that could initiate an accident. Therefore, granting the LAR does not involve a significant increase in the probability or consequences of an accident previously evaluated. The dose consequences of all design basis accidents are unchanged by this proposed amendment.
- 2. Does not create the possibility of a new or different type of accidentfrom any accident previously evaluated.
The function of the SW system considered in the design basis is to remove process and operating heat from safety-related components during normal as well as transient conditions. The proposed amendment to limit the allowed ACTION Completion Time to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> will ensure the function of the SW system is consistent with the design basis and will not result in changes to the design, physical configuration of the plant or the assumptions made in the safety analysis. The requirement does not change the function of the system nor its ability to perform its design function. No alteration to plant configuration or operation is proposed. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does not involve a significant reduction in the margin of safety.
CR-3's design basis considers adequate heat removal by the SW system to cool the containment fan assembly cooling coils and fan motors, spent fuel pool, SW pump motors and other equipment which must function following an accident. This proposed amendment will not alter the current design basis.
By limiting the allowed ACTION Completion Time to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the proposed amendment to ITS 3.7.7 will limit the time the safety function of the SW system can be compromised. Therefore, the amendment does not result in a reduction of the margin of safety.
Based on the above, CR-3 concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.
U.S. Nuclear Regulatory Commission Attachment B 3F1006-02 Page 2 of 2 Applicable Regulatory Requirements/Criteria The proposed amendment is not a risk-informed change. The operation of the system will be the same as is currently considered in the current Crystal River Unit 3 Probabilistic Risk Analysis.
Environmental Impact Evaluation 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) result in a significant increase in individual or cumulative occupational radiation exposure.
Progress Energy Florida, Inc. has reviewed proposed License Amendment Request #293, Revision 0, and concludes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with this request.
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT C LICENSE AMENDMENT REQUEST #293, Revision 0 Proposed Revised Improved Technical Specifications Pages Strikeout/Shadowed Format Strikeout Text Indicates Deleted Text fid7owed T~ex Indicates Added Text
SW SYSTEM LCO 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Nuclear Services Closed Cycle Cooling Water (SW) System LCO 3.7.7 The SW System shall be OPERABLE with:
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One emergency SW pump A.1 Restore SW system to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status OR One required SW heat exehanger inoperablTe B. Required Action and B.3 Be in MODE 30. 6 hou-rs assoei ated Completion Timie not met. AND B.2 Be in MODE 5. 36 houtrs xchanger inoperable ~ PERABLE status' re d~Rqi oMODE72 UF Bii _
Sssoci ated Cornpletiý ýq line not met.! D rUzl* i3-nI 7I Crystal River Unit 3 3.7-15 Amendment No. IR I
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT D LICENSE AMENDMENT REQUEST #293, Revision 0 Proposed Revised Improved Technical Specifications Pages Revision Bar Format
SW SYSTEM LCO 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Nuclear Services Closed Cycle Cooling Water (SW) System LCO 3.7.7 The SW System shall be OPERABLE with:
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One emergency SW pump A.1 Restore SW system to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status B. One required SW heat B.1 Restore SW system to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exchanger inoperable. OPERABLE status C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Crystal River Unit 3 3.7-15 Amendment No.
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT E LICENSE AMENDMENT REQUEST #293, Revision 0 Existing Improved Technical Specifications Bases Revision Bar Format
SW System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Nuclear Services Closed Cycle Cooling Water System (SW)
BASES BACKGROUND The Nuclear Services Closed Cycle Cooling Water (SW) System removes process and operating heat from safety related components during normal operations as well as during a transient or accident conditions. The system consists primarily of one normal service pump, two emergency pumps, a surge tank pressurized with nitrogen for volume and pressure control, and four heat exchangers which reject heat to the Nuclear Services Seawater System. During normal operations, the normal duty (i.e., non-engineered safeguards (ES)) SW pump provides cooling to various essential and non-essential components. Additionally, the SW System provides cooling to the spent fuel pool cooling system heat exchangers during all operating conditions.
The design and operation of the SW System, along with a list of the components supplied during normal and emergency conditions, can be found in FSAR Section 9.5 (Ref. 1). On an Engineered Safeguards Actuation System (ESAS) signal, SW System components realign to provide cooling to ES equipment required post-accident. To ensure the system can accommodate these additional heat loads, non-essential loads are isolated.
The SW System also supplies cooling to the containment fan assembly cooling coils and fan motors for containment heat removal following a loss of coolant accident (LOCA). Other vital heat loads supplied by the SW System during emergency operations are the motor-driven EFW pump, air coolers for the SW and Nuclear Services Seawater System ES pump motors one or two make-up and purification pumps and motors (two pumps are normally aligned for SW cooling),spent fuel coolers, and spent fuel cooling pump air handing units.
As a closed system, the SW System also serves as an intermediate barrier to radioactivity releases to the environment from potential leaks in associated systems. To prevent in-leakage from the containment atmosphere, the SW System is operated at a pressure higher than that which would be present in the containment following a LOCA or steam line break (SLB) (Ref. 2).
(conti nued)
Crystal River Unit 3 B 3.7-36 Amendment No. 149
SW System B 3.7.7 BASES APPLICABLE The SW System provides cooling for components essential for SAFETY ANALYSIS the mitigation of design basis accidents. An ESAS signal will start both emergency SW pumps (each pump is actually two pump assemblies driven by a single motor), transfer cooling of the containment fan assembly cooling coils and fan motors from the CI System to the SW System, and isolate various non-essential loads. The two emergency pumps (100 percent capacity each), in conjunction with adequate heat removal ability by the heat exchangers provide the necessary ca ility for cooling the motor-driven EFW pump, containment fan assembly cooling coils and fan motors, spent fuel pool, SW and Nuclear Services Seawater System pump motors, and other equipment which must function following an accident.
By supplying the containment fan assembly cooling coils and fan motors following a LOCA, the SW System and the Reactor Building Spray System act in conjunction to ensure the pressure and temperature in containment are maintained less than the design limits. The OPERABILITY of the Reactor Building Spray System is addressed by LCO 3.6.6.
The Nuclear Services Closed Cycle Cooling Water System satisfies Criterion 3 of the NRC Policy Statement.
LCOs The requirement for OPERABILITY of both emergency SW pumps and adequate heat removal ability by the SW heat exchangers in MODES 1 through 4 provides sufficient capacity to ensure adequate postaccident heat removal, considering a worst case single active failure. Each emergency SW pump is powered from a separate 4160 V ES bus. Each of the two sets of emergency SW pumps is capable of supplying 100 percent of the required system flow. Each heat exchanger is rated at one-third the total required heat transfer rate for both normal and emergency operations with an Ultimate Heat Sink (UHS) temperature of 95 0 F. Therefore, normal operations will include three of four heat exchangers in service which will allow the fourth heat exchanger to be removed from the system for maintenance (Ref. 2).
APPLICABILITY In MODES 1, 2, 3, and 4, the SW system is a normally operating system that must be capable of performing its post-accident safety functions, which in ude providing cooling water to components required for Reactor Coolant System (RCS) and containment heat removal, equipment essential to safely shutdown the plant, and equipment required for adequate spent fuel pool cooling.
(continued)
Crystal River Unit 3 B 3.7-37 Revision No. 62
SW System B 3.7.7 BASES APPLICABILITY The number of heat exchangers required is based on meeting (continued) the design basis heat transfer rate for both normal and emergency operations. With 0a design basis Ultimate Heat Sink (UHS) temperature of 95 F, three heat exchangers are required to be in service (Ref. 4). Therefore, normal operations will have three of the four heat exchangers in service. Operation with less than three heat exchangers in service per Action A, could only be allowed with an engineering evaluation that ensures the design basis heat transfer rate is available based on the number of heat exchangers, expected debris blocked tubes, and the UHS temperature.
In MODES 5 and 6, the SW System is not required to be OPERABLE due to the limitations on RCS temperature and pressure in these MODES. Additionally, there are no other echnical Specification LCOs supported by SW which are applicable during these plant conditions.
ACTIONS A.1 If one of the emergency SW pumps is inoperable, action must I be taken to restore the affected component(s) to OPERABLE I status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time for restoring full SW System OPERABILITY is consistent with other ECCS Specifications for a loss of redundancy Condition and, has been shown to maintain a suitable limit on risk. As such, this Completion Time is based on engineering judgment and is consistent with i ndustry-accepted practi ce.
Note: The following is an administrative control per NRC Administrative Letter 98-10. If only two SWHE are OPERABLE, the SW system may not have adequate heat removal ability.
If an engineering evaluation demonstrates that adequate heat removal is available based on the number of heat exchangers, expected debris blocked tubes, and the UHS temperature, then the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ACTION may be utilized. If the evaluation does not demonstrate adequate heat removal, then LCO 3.0.3 applies.
B.1 and B.2 If the inoperable SW component(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
Crystal River Unit 3 B 3.7-38 Revision No. 62
SW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the SW flow path provides assurance that the proper flow paths exist for SW operation. The isolation of the SW flow to individual components may render these components inoperable, but does not affect the operability of the SW system. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing.
These valves include valves in the main flow paths and the first normally closed valve in a branch line. In lieu of the first normally closed valve in the branch line, credit may be taken for verifying valve position of another valve downstream, providing the isolation of the flow path is achieved. Verifying correct valve alignment of valves immediately downstream of an unsecured valve still assures isolation of the flow path. There are several exceptions for valve position verification due to the low potential for these types of valves to be mispositioned. The valve types which are not verified as part of this SR include vent or drain valves, relief valves, instrumentation valves, check valves and sample line valves. A valve that receives an actuation signal is allowed to be in a non-accident position provided the valve will automatically reposition within the proper stroke time. For a power operated valve to be considered "locked, sealed, or otherwise secured," the component must be electrically and physically restrained. This surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in their correct position.
The 31 day frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.
SR 3.7.7.2 This SR verifies proper automatic operation of the SW valves on an actual or simulated actuation signal. The SW System is a normally operating system that cannot be fully actuated as part of routine testing during at-power operation. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was considered acceptable from a reliability standpoint.
(continued)
Crystal River Unit 3 B 3.7-39 Revision No. 62
SW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.2 (continued)
REQUIREMENTS The SR is modified by a note indicating the SR is not applicable in the identified MODE. This is necessary in order to make the requirements for automatic system response consistent with those for the actuation instrumentation.
SR 3.7.7.3 This SR verifies proper automatic operation of the SW emergency pumps on an actual or simulated actuation signal.
The SW System is a normally operating system that cannot be fully actuated as part of routine testing during at-power operation. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was considered acceptable from a reliability standpoint.
The SR is modified by a note indicating the SR is not applicable in the identified MODE. This is necessary in order to make the requirements for automatic system response consistent with those for the actuation instrumentation.
REFERENCES 1. FSAR, Section 9.5.
- 2. Enhanced Design Basis Document for Nuclear Services Closed Cycle Cooling Water System.
- 3. FSAR, Section 9.3.
- 4. Calculation M97-0133, "SW Heat Loads During LBLOCA and SW Temperature Decay Time" Crystal River Unit 3 B 3.7-40 Revision No.62
PROGRESS ENERGY FLORIDA, INC.
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT F LICENSE AMENDMENT REQUEST #293, Revision 0 Proposed Revised Improved Technical Specifications Bases Strikeout/Shadowed Format Stik*cout Text Indicates Deleted Text
- hadowed Text Indicates Added text
SW System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Nuclear Services Closed Cycle Cooling Water System (SW)
BASES BACKGROUND The Nuclear Services Closed Cycle Cooling Water (SW) System removes process and operating heat from safety related components during normal operations as well as during a transient or accident conditions. The system consists primarily of one normal service pump, two emergency pumps, a surge tank pressurized with nitrogen for volume and pressure control, and four heat exchangers which reject heat to the Nuclear Services Seawater System. During normal operations, the normal duty (i.e., non-engineered safeguards (ES)) SW pump provides cooling to various essential and non-essential components. Additionally, the SW System provides cooling to the spent fuel pool cooling system heat exchangers during all operating conditions.
The design and operation of the SW System, along with a list of the components supplied during normal and emergency conditions, can be found in FSAR Section 9.5 (Ref. 1). On an Engineered Safeguards Actuation System (ESAS) signal, SW System components realign to provide cooling to ES equipment required post-accident. To ensure the system can accommodate these additional heat loads, non-essential loads are isolated.
The SW System also supplies cooling to the containment fan assembly cooling coils and fan motors for containment heat removal following a loss of coolant accident (LOCA). Other vital heat loads supplied by the SW System during emergency operations are the motor-driven EFW pump, air coolers for the SW and Nuclear Services Seawater System ES pump motors one or two make-up and purification pumps and motors (two pumps are normally aligned for SW cooling),spent fuel coolers, and spent fuel cooling pump air handing units.
As a closed system, the SW System also serves as an intermediate barrier to radioactivity releases to the environment from potential leaks in associated systems. To prevent in-leakage from the containment atmosphere, the SW System is operated at a pressure higher than that which would be present in the containment following a LOCA or steam line break (SLB) (Ref. 2).
(continued)
Crystal River Unit 3 B 3.7-36 Amendment No. 149
SW System B 3.7.7 BASES I APPLICABLE The SW System provides cooling for components essential for SAFETY ANALYSIS the mitigation of design basis accidents. An ESAS signal will start both emergency SW pumps (each pump is actually two pump assemblies driven by a single motor), transfer cooling of the containment fan assembly cooling coils and fan motors from the CI System to the SW System, and isolate various non-essential loads. The two emergency pumps (100 percent capacity each), in conjunction with ade-uate heat r,,oval ability by t Mhe exchangers heat rxchAn provide the necessary capability for cooliing the motor-driven EFW pump, contalnment fan assembly cooling coils and fan motors, spent fuel pool, SW and Nuclear Services Seawater System pump motors, and other equipment which must function following an accident.
By supplying the containment fan assembly cooling coils and fan motors following a LOCA, the SW System and the Reactor Building Spray System act in conjunction to ensure the pressure and temperature in containment are maintained less than the design limits. The OPERABILITY of the Reactor Building Spray System ,is addressed by LCO 3.6.6.
The Nuclear Services Closed Cycle Cooling Water System satisfies Criterion 3 of the NRC Policy Statement.
LCOs and requirement The hre. . adquate removal of OPERABILITY for heat both emergency ability heatpumps by the SW SW I
exchangers in MODES 1 through 4 provides sufficient capacity to ensure adequate postaccident heat removal, considering a worst case single active failure. Each emergency SW pump is powered from a separate 4160 V ES bus.
Each of the two sets of emergency SW pumps is capable of supplying 100 percent of the required system flow. Each heat exchanger is rated at one-third the total required heat transfer rate for both normal and emergency operations with an Ultimate Heat Sink (UHS) temperature of 95 0 F.
Therefore, normal operations will include three of four heat exchangers in service which will allow the fourth heat exchanger to be removed from the system for maintenance (Ref. 2).
APPLICABILITY In MODES 1, 2, 3, and 4, the SW system is a normally operating system that must be capable of performing its post-accident safety functions, which in ude providing cooling water to components required for Reactor Coolant System (RCS) and containment heat removal, equipment essential to safely shutdown the plant, and equipment required for adequate spent fuel pool cooling.
(continued)
Crystal River Unit 3 B 3.7-37 Revision No. CrR
SW System B 3.7.7 BASES APPLICABILITY The number of heat exchangers required is based on meeting (continued) the design basis heat transfer rate for both normal and emergency operations. With 0a design basis Ultimate Heat Sink (UHS) temperature of 95 F, three heat exchangers are required to be in service (Ref. 4). Therefore, normal operations will have three of the four heat exchangers in service. Operation with less than thre heat Pexhanesi servire per Acir ,cd alwdwith-nyb an engieerng ealutiontha cn1SUre the desi?! basis heat transfer rateý is ;4ail ;l based on the nubrof heat exchangers, expeeted debris blocked tubes, and the UIIS In MODES 5 and 6, the SW System is not required to be OPERABLE due to the limitations on RCS temperature and ressure in these MODES. Additionally, there are no other Technical Specification LCOs supported by SW which are applicable during these plant conditions.
ACTIONS A.1 If one of the emergency SW pumps is inoperable, action must I be taken to restore the affected component(s) to OPERABLE I status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time for restoring full SW System OPERABILITY is consistent with other ECCS Specifications for a loss of redundancy Condition and, has been shown to maintain a suitable limit on risk. As such, this Completion Time is based on engineering judgment and is consistent with industry-accepted practice.
Note: The followin isandmi nistrati ve control pe~r MrC tLh eI
- s1I aI ipTay -W not U .& ai L II have
- r l A; 11* .
If:an eineigealuatioln demonstrates that adeut heat rem.val I aIv-a*lae I basd oI the number of heat- exchnl.,a
!exected debris blocked tubes andý the$.In- temeraturete th M2hur ACTI3ON may be?ti .... d. If1 the evaluation does
-not demonstrate adequatte heat removal, then LCO 3.0.)
applies.
B.1 ftnd-B--Ž lf-dfd6ftbth rýFee Ceq0-e h ta rs -i s NOPERABLE, action must be taken to restore-the SW-s1sferifi o OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion ime for restori ng full SW-System OPERABILITY is acceptabl-ased on engineering analysis considering conservative-'-
ssumptions in the determination of maximum SW temperatufre ollowing a Large Break LOCA (LBLOCA), the low probabi~lity f an accident that would deliver a high heat load to" 1-ontAi-nm-ent occurrin -duri~ng_ the- 8..hour Completion Tiff-(continued)
Crystal River Unit 3 B 3.7-38 Revision No. 62
SW System B 3.7.7 BASES ACTIONS Tf u U-,1r R-con, I Y
[ t-F-va and --avai*I b tio-6-nW That p6roVi dee--fe-ns idi epth to aid in the !mitigatijon o the SW het load duri ng B.-:an~d-B-.-i F(71 a-nad-C7Z If the inoperable SW component(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
Crystal River Unit 3 B 3.7-38 Revision No. f&
SW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the SW flow path provides assurance that the proper flow paths exist for SW operation. The isolation of the SW flow to individual components may render these components inoperable, but does not affect the operability of the SW system. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing.
These valves include valves in the main flow paths and the first normally closed valve in a branch line. In lieu of the first normally closed valve in the branch line, credit may be taken for verifying valve position of another valve downstream, providing the isolation of the flow path is achieved. Verifying correct valve alignment of valves immediately downstream of an unsecured valve still assures isolation of the flow path. There are several exceptions for valve position verification due to the low potential for these types of valves to be mispositioned. The valve types which are not verified as part of this SR include vent or drain valves, relief valves, instrumentation valves, check valves and sample line valves. A valve that receives an actuation signal is allowed to be in a non-accident position provided the valve will automatically reposition within the proper stroke time. For a power operated valve to be considered "locked, sealed, or otherwise secured," the component must be electrically and physically restrained. This surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in their correct position.
The 31 day frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.
SR 3.7.7.2 This SR verifies proper automatic operation of the SW valves on an actual or simulated actuation signal. The SW System is a normally operating system that cannot be fully actuated as part of routine testing during at-power operation. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was considered acceptable from a reliability standpoint.
(continued)
Crystal River Unit 3 B 3.7-39 Revision No. 62
SW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.2 (continued)
REQUIREMENTS The SR is modified by a note indicating the SR is not applicable in the identified MODE. This is necessary in order to make the requirements for automatic system response consistent with those for the actuation instrumentation.
SR 3.7.7.3 This SR verifies proper automatic operation of the SW emergency pumps on an actual or simulated actuation signal.
The SW System is a normally operating system that cannot be fully actuated as part of routine testing during at-power operation. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was considered acceptable from a reliability standpoint.
The SR is modified by a note indicating the SR is not applicable in the identified MODE. This is necessary in order to make the requirements for automatic system response consistent with those for the actuation instrumentation.
REFERENCES 1. FSAR, Section 9.5.
- 2. Enhanced Design Basis Document for Nuclear Services Closed Cycle Cooling Water System.
- 3. FSAR, Section 9.3.
Crystal River Unit 3 B 3.7-40 Revision No. 62