ML16138A771
| ML16138A771 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/13/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A770 | List: |
| References | |
| NUDOCS 9401180066 | |
| Download: ML16138A771 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.203 TO FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO. 203 TO FACILITY OPERATING LICENSE DPR-47 AND AMENDMENT NO.200 TO FACILITY OPERATING LICENSE DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated May 3, 1993, as supplemented August 11, 1993, Duke Power Company (the licensee) submitted a request for changes to the Oconee Nuclear Station, Units 1, 2, and 3 Technical Specifications (TS).
The requested changes would revise the limiting conditions of operation and surveillance requirements related to the Low Pressure Service Water (LPSW), the Low Pressure Injection (LPI), the High Pressure Injection (HPI), the Reactor Building Cooling Unit (RBCU), and the Reactor Building Spray (RBS) systems.
Administrative changes are included to delete redundant requirements, correct a misspelling and update the table of contents.
2.0 EVALUATION The staff's review and evaluation of each proposed change is given below.
Administrative changes to the table of contents and introduction (pages iv, viii, ix, x, and xi):
The licensee proposes to revise the table of contents to add new surveillance requirement TS 4.5.3, "Containment Heat Removal Capability," delete references to Figures 4.5.1-1, 4.5.1-2, and 4.5.2-1, delete blank pages and revise associated page and section numbers. These changes to the table of contents are purely administrative in nature, do not affect the substance of the TS changes themselves, and are acceptable.
TS 3.3.2 - Extension of the allowable outage time for one Low Pressure Injection (LPI) train inoperable from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
The licensee proposes to revise TS 3.3.2.a(2) to extend the allowable outage time for one LPI train inoperable from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. During this time period, the remaining operable LPI train would be capable of mitigating the consequences of a design basis accident. This change is consistent with the requirements of NUREG-1430, "Standard Technical Specifications for B&W Plants," LCO 3.5.2 (ECCS - Operating), Required Action A.1. The 72-hour 9401180~c066 940113 PR ADoCIL 05000269
completion time is reasonable, based on the redundant capabilities afforded by the operable LPI train, and the low probability of a Design Basis Accident (DBA) occurring during this period.
TS 3.3.5 - Relocation of TS requirement to lock open valve LPSW-108:
Valve LPSW-108 is the LPSW isolation valve on the discharge side of the cooler in each Oconee unit's Reactor Building Cooling Unit (RBCU). Currently, TS 3.3.7 "Low Pressure Service Water System" requires that valve LPSW-108 be locked open. In the event this valve were to close, the associated RBCU would be inoperable. However, the operability of the entire LPSW system would not be affected. Therefore, since the requirement to lock open valve LPSW-108 pertains more directly to the RBCU system than to the LPSW system, the licensee proposes to relocate the requirement to TS 3.3.5 "Reactor Building Cooling (RBC) System."
The Bases on page 3.3-6 have been revised to describe this change.
We find that the proposed relocation of the requirement to lock open valve LPSW-108 from TS 3.3.7 to TS 3.3.5 is administrative in nature and does not change the substance or effect of the requirement. Therefore, the change is acceptable.
TS 3.3.7.a - Requirement for the third LPSW pump in the shared Unit 1 and Unit 2 LPSW system to be operable:
For the shared Unit 1 and Unit 2 LPSW system, TS 3.3.7a currently requires only two of the three LPSW pumps to be operable at all times. However, calculations of the consequences of severe Loss-of-Coolant Accidents (LOCAs) are based on the assumption that two LPSW pumps are operating. Two operating LPSW pumps are required to provide adequate post-LOCA cooling of the reactor building. If only two LPSW pumps are operable and one of them fails, then the requirement for two operating pumps could not be met. To provide for the possible single failure of one LPSW pump, all three LPSW pumps must be operable. Therefore, the proposed revision of TS 3.3.7.a to require all three LPSW pumps to be operable in the shared Unit 1 and Unit 2 LPSW system is acceptable.
TS 3.3.7.b - Extension of the allowable outage time for one LPSW train inoperable from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
For Unit 3, the LPSW is supplied by either of the two LPSW trains, each containing a LPSW pump required to be operable by TS 3.3.7.a. In normal or post-accident operation, one operating pump supplies the service water needs of the unit. In the event of a failure of the operating pump or its associated train, the redundant operable LPSW pump and train would be capable of providing the required service water.
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-3 For Units 1 and 2, the LPSW is supplied to each unit by either of the two LPSW trains from a shared LPSW system containing three LPSW pumps required to be operable by TS 3.3.7.a. In normal or accident operation, two of the three pumps are operating, so that the flow of one pump is available to supply the service water needs of each unit. The third operable LPSW pump could supply these needs in the event of a failure of one of the operating pumps or a component of its train.
The licensee proposes to extend the allowable outage time for one LPSW train inoperable from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. During this time period, the remaining operable LPSW train would be capable of mitigating the consequences of a design basis accident. This change is consistent with the requirements of NUREG-1430, "Standard Technical Specifications for B&W Plants," LCO 3.7.7 (Component Cooling Water System), Required Action A.1.
The 72-hour completion time is reasonable, based on the redundant capabilities afforded by the operable LPSW train, and the low probability of a DBA occurring during this period. The extended 72-hour outage also provides a more adequate time period for the repair or replacement of an inoperable LPSW system component.
Therefore, extending the allowable outage time for one LPSW train inoperable to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is acceptable. Therefore, this TS change is acceptable.
TS 4.5.1.1.1 Bases - Revision to High Pressure Injection (HPI) testing Bases:
The licensee proposes to revise the Bases associated with TS 4.5.1.1.1 to make it clear that the intent of this HPI testing requirement is to verify proper response to actuation of Engineered Safeguards (ES) channels 1 and 2 by the HPI system, as indicated by control room instrumentation. The test is not intended to verify HPI pump performance, which is tested in accordance with American Society of Mechanical Engineers (ASME) Code requirements (Section XI IWP) as required by 10 CFR Part 50.55a(f).
This revision to the Bases of TS 4.5.1.1.1 is acceptable because it merely clarifies the intent of the specification.
TS 4.5.1.1.2 - Revision to Bases for testing Low Pressure Injection (LPI) and related LPSW Systems:
The Bases associated with TS 4.5.1.1.2 have been revised to clarify the intent of the LPI testing requirements of TS 4.5.1.1.2. The intent is to verify proper response, as indicated by control room instrumentation, to actuation of Engineered Safeguards (ES) channels 3 and 4 by the LPI system, as well as by LPSW system components which support the LPI system (e.g., valves LPSW-4 and LPSW-5).
The test is not intended to verify containment heat removal capability of the LPI coolers; this is accomplished by testing in accordance with proposed Specification 4.5.3. The test is not intended to verify LPSW pump performance, which is tested in accordance with ASME Section XI IWP requirements.
This revision to the Bases of TS 4.5.1.1.2 is acceptable because it serves to clarify the intent of the specification.
-4 TS 4.5.1.2.1 - Deletion of redundant testing requirements:
Currently, TS 4.5.1.2.1 specifies that the HPI and LPI pumps be tested.in accordance with TS 4.0.4. In addition, this TS requires verification of initial pump startup and operation for 15 minutes with discharge pressure and flow within +/- 10 percent of a point on the generic pump head curves in Figures 4.5.1-1 and 4.5.1-2 as the acceptance criteria. This requirement is redundant to both TS 4.0.4 and 10 CFR 50.55a, which specify testing of safety-related pumps in accordance with the ASME Boiler and Pressure Vessel Code. The Code requires measurement of pressure head and flow.at a single selected point of operation. When compared with previous measurements of flow and pressure head, this testing will monitor pump degradation and will indicate a shift in the pump head-flow curve. Pump curves are available in the Final Safety Analysis Report for indicating design basis/safety limits for these pumps.
The deletion of TS 4.5.1.2.1 is acceptable because its requirements are redundant to those of TS 4.0.4 and 10 CFR 50.55a. The proposed deletion of Figures 4.5.1-1 and 4.5.1-2 is acceptable because these pump curves are available in the Final Safety Analysis Report.
TS 4.5.2.1.1 - Extension of the test interval for the Reactor Building Spray (RBS) system spray nozzle flow test from 5 years to 10 years:
Currently, TS 4.5.2.1. requires testing for obstruction of the spray nozzles every five years by blowing compressed air or fog through the spray headers and nozzles. After additional review of the required frequency of spray nozzle testing, and consideration of the passive nature of the design of the spray nozzles, the NRC staff concluded that a test at 10-year intervals is adequate to detect obstruction of the spray nozzles. The 10-year test interval is specified in Surveillance Requirement 3.6.6.8. in NUREG-1430, "Standard Technical Specifications for Babcock and Wilcox (B&W) Plants,"
September 28, 1992. Also, in NUREG-1366, "Improvements to Technical Specification Surveillance Requirements," September 28, 1992, the NRC staff recommends that the test interval for testing spray nozzles be extended to 10 years. The 10-year surveillance interval for nozzle plugging in spray systems constructed of stainless steel tubing was also included in Generic Letter (GL) 93-05, "Line-Item TS Improvements to Reduce Surveillance Requirements Testing during Power Operation," dated September 27, 1993.
For spray system piping constructed of carbon steel, any change of the surveillance interval must be justified. By telephone on November 1, 1993, the licensee confirmed that the spray system piping in the three Oconee units is constructed of stainless steel.
In light of these generic recommendations for B&W plants, we find the proposed revision to TS 4.5.2.1.1, extending the test interval to 10 years, acceptable.
The licensee has relocated TS 4.5.2.1.1.c to become a subsection of TS 4.5.2.1.1.a to make it clear that these test acceptance criteria (visual observation and control board indication of response to the actuation signal) apply to the test of TS 4.5.2.1.1.a but do not apply to the nozzle air flow test of TS 4.5.2.1.1.b. We find this revision is administrative in nature and clarifies the intent of TS 5.4.5.2.1.1; therefore, the revision is acceptable.
-5 TS 4.5.2.1.1.a Bases -
Revision to Bases for testing the RBS System:
The Bases of TS 4.5.2.1.1.a have been revised to provide additional clarification that the intent of revised TS 4.5.2.1.1.a is to verify proper response to actuation of Engineered Safeguards (ES) channels 7 and 8 by the Reactor Building Spray system.
Also, the reference in the Bases to a spray pump flow acceptance criterion (1000 gpm) has been deleted, as well as a reference to the monthly rotation of LPSW pumps. These revisions to the Bases are acceptable because the LPSW pumps are tested per the requirements of the ASME Code and 10 CFR Part 50.55a.
These tests do not verify the capability of the pumps to meet design basis but do monitor the pumps for degradation.
TS 4.5.2.1.2.b - Revision of RBCU testing requirements to show acceptance criteria in terms of response to an ES signal:
The current RBCU test acceptance criteria in TS 4.5.2.1.2.b include specifications of LPSW flow through each cooler (greater than 1400 gpm) and air flow through each fan (greater than 40,000 CFM). The licensee proposes to revise the surveillance to delete the LPSW flow and air flow test requirements, and to state acceptance criteria in terms of component response to an ES signal.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly, the appropriate valves have completed their travel and fans are running at half speed.
The testing of the RBCU system pumps and valves is performed in accordance with the ASME Code, which does not require the special LPSW flow and air flow tests. However, the cooling capability of the RBCU system will be directly verified by the requirements of the proposed new TS 4.5.3 (see below).
Therefore, the deletion of the LPSW flow and air flow test requirements from TS 4.5.2.1.2.b is acceptable.
TS 4.5.2.1.2 Bases - Revision of RBCU/LPSW testing requirements:
The licensee proposes to revise the Bases of TS 4.5.2.1.2 to make clear that the intent of its testing requirements is to verify response to activation of Engineered Safeguards (ES) channels 5 and 6 by the RBCUs, as well as by LPSW system components which support the RBCUs. The intent is not to verify containment heat removal capability of the RBCUs; this is accomplished by additional testing in accordance with proposed new TS 4.5.3. This revision is acceptable because it helps to clarify the intent of TS 4.5.2.1.2.
TS 4.5.2.2 - Deletion of redundant testing requirements:
Currently, TS 4.5.2.2 specifies that RBS system pumps and valves be tested in accordance with TS 4.0.4. In addition, this TS references a generic pump head curve in Figure 4.5.2-1 as the acceptance criterion. This requirement is redundant to both TS 4.0.4 and 10 CFR 50.55a, which require testing of safety related pumps and valves in accordance with the ASME Boiler and Pressure Vessel Code. The Code requirements specify testing at reference values of pressure head and flow. These tests do not verify the capability of the pumps to meet design basis but do monitor the pumps for degradation.
-6 The proposed deletion of TS 4.5.2.2 is acceptable because the testing of RBS system pumps and valves in accordance with the ASME Code is already required by TS 4.0.4 and 10 CFR 50.55a. The deletion of Figure 4.5.2-1 is also acceptable because it is available in the Final Safety Analysis Report for indicating design basis/safety limits for these pumps.
TS 4.5.2 Bases - Editorial change:
In the first sentence of the Bases relating to TS 4.5.2, the phrase "Reactor Building Coolant Systems," has been corrected to read "Reactor Building Cooling Systems." This change revising incorrect terminology is acceptable.
New TS 4.5.3 - Containment Heat Removal Capability:
The TS relating to the LPSW, LPI, RBS, and RBCU testing requirements do not specifically verify that these systems are capable of performing the intended safety function of maintaining containment pressure and temperature below design limits following an accident. It was presumed that the LPSW flow and air flow through each RBCU specified in the current TS 4.5.2.1.2, for example, would ensure an adequate post-accident containment heat removal capability for each RBCU.
These specifications did not provide for the possible loss of heat removal capability by service-induced fouling of the heat exchangers (coolers) in the LPSW, LPI, and RBCU systems. Therefore, the licensee proposes to add the new TS 4.5.3 requiring the specific surveillance of containment heat removal capability on a refueling frequency (TS 4.5.3.1.a). In addition, TS 4.5.3.1.b requires the determination of the fouling rate of the LPI and RBCU coolers so that the frequency of surveillance may be modified, if required to ensure that containment heat removal capability remains sufficient to maintain post accident conditions within design limits.
With the addition of the proposed new TS 4.5.3, an additional justification is provided for the deletion of the LPSW pump flow and RBCU fan air flow requirements of TS 4.5.2.1.2, from which containment heat removal capability is currently inferred. This capability is more reliably determined from the surveillance of the proposed TS 4.5.3. As discussed above, the surveillance in the proposed new TS 4.5.3 would provide increased assurance that the containment heat removal systems would maintain post-accident conditions in the containment within design limits. Therefore, the proposed new TS 4.5.3 is acceptable.
New Bases for the new TS 4.5.3 have been added to clarify the need for the new specification.
Administrative changes to renumber the technical specifications:
Current TS 4.5.3, "Penetration Room Ventilation System," and TS 4.5.4, "Low Pressure Injection System Leakage," have been renumbered due to the addition of new TS 4.5.3, "Containment Heat Removal Capability." These changes are purely administrative in nature, and are, therefore, acceptable.
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3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (58 FR 52983 dated October 13, 1993). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: S. Kirslis Date: January 13, 1994