ML20106E941
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Catawba Nuclear Station UFSAR Appendix 14A. Tables Appendix 14A. Tables
Catawba Nuclear Station UFSAR Table 14-1 (Page 1 of 18)
Table 14-1. Compliance with Regulatory Guides (HISTORICAL INFORMATION - NOT REQUIRED TO BE REVISED)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification 1.68 Rev. 2 Partial C.3 The scope of testing will include, to the The inclusion of all control system and extent practicable, simulation of the effects equipment malfunctions which could of control system and equipment reasonably be expected to occur during malfunctions that have been identified in the plant lifetime is too broad. The scope of plant accident analysis as having a failures which could be expected to occur significant impact on the results of the should be limited to those failures which analysis. would have a significant impact on the analysis of events evaluated in 15.0.
App. A 1.b (2) Correct failure mode on loss of power to Only solenoid and pneumatically chemical control system valves will be actuated components are subject to limited to pneumatic and solenoid actuated change of state on loss of actuating valves and components. power.
App. A. 1.e Only those portions of power conversion Only power conversion system piping systems designated as Duke Safety Class B designated as Duke Safety Class B need are subject to thermal expansion and be subject to thermal expansion and restraint testing. The adequacy of power restraint testing. Portions of power conversion system piping designated Duke conversion systems designated as Duke Safety Class F or G is demonstrated Safety Class F or G can be adequately through visual verification during Hot verified by visual inspection for Functional Testing and Initial Startup indications of piping interference and Testing. abnormal stressing of restraints. All piping inside the main steam and feedwater isolation valves is Duke Safety Class B.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 2 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 1.g (1) Load carrying capability of breakers, motor Verification of all components by direct controllers, switchgear, transformers, and measurement is impractical due to the cables will be demonstrated by proper large number of cables, transformers, normal operation of the specified equipment switchgear, and load centers affected.
under conditions approximating full rated Selected component direct measurements load. Current measurements will be made under conditions approximating full rated at selected normal power system locations load combined with verification of normal to provide additional verification of proper operation during unit startup will be equipment sizing adequate to ensure proper sizing and selection of equipment.
1.68 Rev. 2 Partial App. A 1.g (2) Load - carrying capability of breakers, Verification of all components by direct motor controllers, switchgear, transformers, measurement is impractical due to the and cables will be demonstrated by proper large number of cables, transformers, normal operation of the equipment under switchgear, and load centers affected.
conditions approximating full rated load. Selected component direct measurements Current and temperature readings will be under conditions approximating full rated obtained at selected essential power system load combined with verification of normal locations to provide additional verification operation during unit startup will be of proper equipment selection and sizing. adequate to ensure proper sizing and selection of equipment.
App. A 1.g (2) The ability of essential loads to start under The conditions encountered during these minimum and maximum design voltage tests simulate the range of voltages under conditions will be verified during which essential loads would be required preoperational testing of the emergency to start and operate.
diesel generators and the offsite power system.
App. A 1.i (20) Sufficient measurements and/or Where adequate measurement methods observations will be made to ensure that do not exist, visual inspection procedures gross bypass leakage paths are not present are adequate to limit the bypass leakage for ice condenser containments. area present.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 3 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 1.j Instrumentation and control as well as The systems which will be required to be App. A 1.n auxiliary and miscellaneous systems to be tested by the criteria presented in Section tested will be selected based on the criteria C.1 vary from one plant design to for selection of systems to be tested given in another. The criteria for selection of Section C.1 of the regulatory guide. The systems to be tested given in Section C.1 extent to which the operability of these of the guide in combination with a graded systems is verified is based on a graded approach to the degree of testing to be approach determined by the system or performed provide a valid basis for the components importance to safety and development of tests to be conducted.
normal operation.
App. A 4.c Pseudo-ejected-rod measurements will not The calculational codes and analytical App. A 5.e be performed on Unit 2. methods used for nuclear analysis of the reactor core are presented in FSAR Section 4.3.3. The validity of these codes and safety analysis assumptions for ejected rod worth will be verified as part of the extensive startup testing on Unit 1.
The core design and control rods utilized on Unit 2 are identical to those for Unit
- 1. Control rod bank worths measurements should be sufficient to verify adequacy of ejected rod predictions. Therefore, without any gross errors in the measured bank rod worths, the Unit 2 pseudo ejected rod worth should be within the safety analysis assumptions.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 4 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 4.g Demonstration of proper process or effluent During initial startup testing historical A 5.z monitoring system response based on data has shown that process and effluent correlation with independent laboratory monitors may not experience levels in analysis will be conducted only for those excess of the minimum sensitivity of the monitors for which process or effluent levels monitor. A meaningful correlation with exceed the minimum sensitivity of the laboratory analysis is not possible for detector. these monitors.
App. A 4.h Demonstration of the operability of reactor Normal Station Chemistry procedures A.4.r coolant/secondary purification and clean up used during initial start-up maintain A.5.a.a. systems. Formal testing will not be water quality within the Technical performed. Specification limits. Normal Station procedures demonstrate Reactor Coolant System and Secondary System water quality. Systems needed to control water chemistry are functionally tested prior to power ascension. During power operation the reactor coolant and secondary side sampling is used to monitor water quality in accordance with Technical Specifications.
App. A.4.i Specific testing to demonstrate the The capability of the Rod Control System operability of control rod sequences and to function properly is demonstrated prior inhibit/blocking functions over the reactor to initial criticality by the Rod Control power level range during low power testing System Alignment Test (see Section 14.5).
will not be performed. Performing this test prior to criticality meets the intent of Regulatory Guide 1.68, Revision 2, Appendix A.4.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 5 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A.4.j Specific testing to demonstrate the The temperature of the upper and lower capability of primary containment containment is routinely monitored once ventilation during low power testing will not every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verified to be within be performed. Technical Specification limits during operation. This surveillance is sufficient to demonstrate the capability of the system to maintain the containment temperatures within the allowed limits during operation.
App. A.4.k Specific testing to demonstrate the The operability of the steam driven operability of steam driven ESF/plant equipment is demonstrated during the Hot auxiliaries and power conversion equipment Functional Testing. The testing of during low power testing will not be Auxiliary Feedwater System, Feedwater performed. and condensate system functional testing is described in the test abstracts (Section 14.4).
App. A.4.l Specific testing to demonstrate the The operability and stroke times of the operability and stroke times of main steam main steam isolation valves are verified line/branch line/bypass valves used for during Reactor Coolant System Hot protective functions during low power Functional Testing (FSAR, Section 14.4).
testing will not be performed.
App. A.4.n Specific testing to demonstrate the The operability of the plant process operability of control room computer computer is assured through extensive system will not be performed during low testing prior to initial criticality.
power testing.
App. A.4.o Specific testing to determine control rod The rod drop test is performed at scram times will not be performed during temperature and pressure prior to initial low power testing. criticality. Refer to the abstract for the Rod Cluster Control Assembly Drop Time Test (Section 14.5).
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 6 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 4.p Demonstration of the operability of Operability of the pressurize code relief pressurizer code relief valves at rated valves need not be conducted by means of temperature may be demonstrated by a an installed functional test due to the bench test verification, performed by the undesirable additional transient imposed valve vendor. Results of the vendor tests as on the valves and associated discharge well as copies of the test procedures will be piping.
available for review.
Demonstration of operability of main steam The open and closure set points of the safety valves will not be performed during main steam safety valves are verified low power testing. temperature during Hot Functional Testing (Section 14.4).
Demonstration of operability of Pressurizer PORV's tested during pressurizer/main steam PORV will not be precritical activities. Refer to Section performed during low power testing. 14.5.
App. A.4.q Demonstration of the operability of RHR The operability of the Residual Heat systems will not be performed during low Removal System up to its temperature and power testing. pressure limits is demonstrated by the Residual Heat Removal System Functional Test and the Reactor Coolant System Hot Functional Test, prior to criticality. The steam dump control operability is verified during the Reactor Coolant System Hot Functional Test.
App. A.4.s Vibration measurement of reactor vessel The reactor internals vibration testing and reactor coolant components will not be has been performed for similar plants in performed during low power testing. the past and a Catawba instrument vibration test program is not necessary.
The recommendations of the Regulatory Guide 1.20 position C.3 are satisfied by the inspection program discussed in FSAR Section 3.9.2.4.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 7 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A.4.u Specific testing to demonstrate major or Major plant control system operation at principal plant control system will not be temperature and pressure conditions has performed during low power testing. been demonstrated during the Reactor Coolant System Hot Functional Test (Abstract, Section 14.4). Other specific control system testing is performed at appropriate power levels during startup testing, as described in Section 14.5 and Figure 14-1.
1.68 Rev. 2 Partial App. A 5 Test and acceptance criteria will be Control system testing should verify developed to demonstrate the ability of proper control of process variables within major principal plant control systems to the design control deadband, not over the automatically control process variables range of design values of process within design limits around the nominal variables. Proper control of process reference value. variables will be demonstrated during power escalation over the range of 0 to 100% F.P.
Partial App. A 5.a Power coefficient measurements will not be NSSS vendor does not recommend performed at 100% power but will be performing this test at 100% power due to performed at 90% power instead (Unit 1 potential of violating axial flux difference Only). Power coefficient measurements will Technical Specification. Unit 2 has not be performed on Unit 2. essentially identical fuel and core loading as Unit 1. Errors between measured and predicted power coefficients at 30%,
50%, 75% and 90% on Unit 1 were less than the acceptance criterion value of
+/-0.5%. There is no reason for Unit 2 measurements to be different from their predicted values.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 8 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 5.b Departure from nucleate boiling ratio Axial, Radial, and Total Peaking will be (DNBR), maximum average planar linear directly measured and verified during heat generation rate (MAPLHGR), and power escalation testing and will be used minimum critical power ratio (MCPR) will to verify DNBR and linear heat rate not be directly verified during power margin by analysis.
escalation testing.
Partial App. A 5.f Core thermal and nuclear parameters will The reactor core will be under xenon not be demonstrated to be in accordance transient conditions at this time. There with predictions following a return of the would be insufficient time to gather data rod to its bank position. under transient conditions. There are no NSSS vendor predictions for this configuration.
App. A 5.g Special testing to demonstrate control rod The capability of the Rod Control System sequencers/withdrawal block functions to function properly is demonstrated prior operation will not be performed. to initial criticality by the Rod Control System Alignment Test (Abstract, Section 14.5). Performing this test prior to criticality meets the intent of Regulatory Guide 1.68, Revision 2, Appendix A.4.
App. A 5.h Rod drop times will not be measured at Measuring rod drop times at power power. would require disabling all position indication for the rods in violation of plant Technical Specifications.
App. A 5.i Test to demonstrate incore/excore From vendor predictions the Xenon and instrumentation sensitivity to detect rod power distributions at 50% and 100% are misalignment will not be performed at full similar. The performance of this test at power. 50% should adequately demonstrate the capability and sensitivity of incore/excore instrumentation to detect control rod misalignments equal to or less than Technical Specifications.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 9 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 5.k Special testing to demonstrate ECCS Testing to demonstrate this capability in operation will not be performed during low accordance with Regulatory Guide 1.79 power ascension testing. will be conducted during the preoperational testing program. Please refer to the Abstract for the Safety Injection System Functional Test (Section 14.4).
Partial App. A 5.l Specific testing to demonstrate capabilities The capability of systems and components of RHR systems will not be performed to remove decay heat will be during power ascension testing. demonstrated during the Station Blackout Test (Abstract, Section 14.5). Testing to demonstrate the capability of the Residual Heat Removal System has been discussed in the response to item A.4.g. Testing is also performed to demonstrate that damaging water hammer does not occur in the feedwater piping (see Section 14.5).
Partial App. A 5.m Differential pressure measurements will not Measured Reactor Coolant System loop be made across the core or major reactor flows will be compared with predicted coolant system components. Reactor Coolant System loop flows. Any gross deviation of actual loop or core pressure drops from predicted values will be identified by detection of the corresponding deviation of measured flow from prediction.
Idle loop flows will not be determined Tech. Specs. does not allow for less than during power ascension testing. full flow operation.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 10 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification Specific measurements for vibration levels The Reactor Coolant System operation of reactor coolant system components will will be shown to be as designed by the not be performed during power ascension performance of the Unit Load Steady-testing. State Test (Abstract, Section 14.5) at various power levels. Reactor Coolant System flow rates will be recorded during the NSSS Thermal Output Test and compared with predictions. Additional vibration testing is not necessary for Catawba reactor internals, since plants of similar design have been extensively tested and found satisfactory. The recommendations of Regulatory Guide 1.20 Position C.3 are satisfied by inspection program discussed in Section 3.9.2.4.
App. A 5.o Calibration and demonstration of the As a normal station operating procedure response of reactor coolant system leak the periodic surveillance test and Reactor detection systems will not be performed Coolant System Leak Test is run at no during power ascension. load conditions following fuel loading prior to initial criticality (see Section 14.5). In addition the Reactor Coolant Leakage detection Systems are calibrated and tested prior to initial criticality as required by the Technical Specifications.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 11 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A.5.p Vibration monitoring of reactor vessel The vibrational monitoring of reactor internals will not be performed during vessel internals is not necessary for power ascension testing. Catawba since many plants of essentially the same design have been extensively tested and found satisfactory. The recommendations of Regulatory Guide 1.20, Position C.3 are satisfied by the inspection program discussed in Section 3.9.2.4.
App. A.5.q Proper operation of failed fuel detection Failed fuel is detected through radio-systems will not be performed during power chemical analysis of reactor coolant ascension testing. samples. The Reactor Coolant System will be sampled and analyzed for specific activity as required by Technical Specifications.
App. A 5.r A verification of computer inputs and Inputs and calculations which do not performance calculations which are utilized serve to ensure compliance with to ensure compliance with the provisions of provisions of station operating license or the station operating license or accident accident analysis bases do not need to be analysis bases will be performed. verified.
App. A 5.t Capacities, set points, and reset pressures Vendor testing is adequate to ensure for the pressurizer mechanical code relief proper operation of the pressurizer code valves will be verified by vendor testing and relief valves. Transient test data obtained verification. during power escalation testing will be utilized to verify proper operation of the pressurizer mechanical code relief valves when reactor coolant system pressure transients of sufficient magnitude to verify proper operation are observed.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 12 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification 1.68 Rev. 2 Partial App. A 5.u Operability of main steam isolation valves Operability of the main steam isolation and branch steam isolation valves will not valves and branch steam isolation valves be verified during power escalation testing under full temperature and pressure at the 25% F.P. platean. conditions will be verified during hot functional testing.
App. A 5.w Demonstration of performance of The Catawba Nuclear Station design does penetration/shielding cooling system will not utilize shielding or penetration not be performed during power ascension cooling systems. Adequate design testing. performance for the main steam line penetrations is verified by concrete temperature measurements taken during Reactor Coolant System Hot Functional Testing (Abstract, Section 14.4).
App. A.5.c.c Specific testing for demonstration that The operability of Liquid and Gaseous gaseous/liquid waste processing, storage Waste Processing Systems is and release systems will not be performed demonstrated prior to the startup testing during power ascension testing. phase. For details see the Abstract for the Waste Gas System Functional Test, Section 14.4.
App.A.5.e.e Specific testing for demonstration that The capability of this system is containment injection and purging systems demonstrated during the preoperational operate within design will not be performed test phase. For details see the Abstract during power ascension testing. for the Containment Ventilation and Purge Functional Test Section 14.4).
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 13 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification Partial App. A.5.f.f Specific testing to verify ventilation systems The temperatures in upper and lower can maintain area design limits will not be containment will be monitored at least performed for containment systems during once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during operation and power ascension testing. verified to be within Technical Specification limits. This surveillance will be sufficient to verify the capability of the system to maintain temperatures within limits in the normal operating mode.
1.68 Rev. 2 Partial App. A 5.i.i. Plant dynamic response for limiting reactor The critical parameter of interest in the coolant pump trips will not be demonstrated analysis of the limiting loss of reactor at 100% F.P. coolant flow is DNBR verses time. Since DNBR is not a directly observable parameter the determination of DNBR behavior verses time following a loss of flow depends primarily on the determination of flow coast down vs. time and the behavior of local clad heat flux verses time following the loss of flow.
The behavior of local clad heat flux verses time cannot be determined directly and the analysis of this behavior is dependent on verification of reactor trip response time for a loss of flow event.
Both flow coast down and reactor trip response time for a loss of flow may be determined directly during the flow coast down test. No additional meaningful data could be obtained from performance of this test at power. Plant dynamic response from power following a four pump reactor coolant pump trip will be verified by the station blackout test.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 14 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification Partial App. A 5.k.k Dynamic response of the plant to the loss or The Feedwater Temperature Variation bypassing of the feedwater heater(s) from a Test will be performed at 90% power.
credible single failure or operator error Performance of this test at 90% power that would result in the most severe case of will provide a severe test of the ability to feedwater temperature reduction will be respond to this transient and control the performed from 90% F. P. Feedwater plant, and will assure that this can be reduction test will not be performed at 50% accomplished throughout the range of F. P. power operation.
1.68 Rev. 2 Partial App. A 5.1.1 Dynamic response of the plant to turbine Because of the reactor trip-on-turbine trip will be demonstrated from the maximum trip logic, reactor trips will automatically power level at which a reactor trip would be actuated upon loss of turbine during not be automatically initiated. the turbine trip test and the unit loss of electrical load test at full power. The resulting transients and plant dynamic response will not be significantly different for these two tests if both are initiated from full load. Performance of the turbine trip test from the highest power below the actuating point of the reactor trip-on-turbine trip logic will allow documentation of the plant dynamic response for the runback situation. The unit loss of electrical load test will be performed from full power to demonstrate the dynamic response to a turbine trip with reactor trip situation.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 15 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification App. A 5.m.m A main steam isolation will not be The severity of the transient to plant demonstrated at power. systems and components does not justify performance of the test. Proper operation of the main steam isolation valves is demonstrated during hot functional testing at full temperature and pressure. It should be noted that in the 15.0 analysis of the inadvertent closure of main steam isolation valves at power (Section 15.2.4), this transient is bracketed by the turbine trip analysis. A turbine trip will be performed at power during startup testing.
Partial App. A.5.o.o. Verify that piping and component, Acceptable expansion movements and movements, vibrations and expansions will vibrations will be demonstrated earlier in not be performed during power ascension the test program by the Reactor Coolant testing except as specified on FSAR Table System Thermal Expansion and Restraint 3-85. Test (Abstract, Section 14.4) and the Piping System Vibration Test (Abstract, Section 14.4). These tests will not be repeated during the power ascension testing program.
1.79 Rev. 1 Partial C.1.b (2) A cold recirculation test will not be Vendor testing and verification is performed as specified by C.1.b (2). Proper adequate to ensure vortex control, proper functioning of valves and interlocks system pressure drops, and that adequate required to ensure proper system alignment net positive suction head will exist for in the sump recirculation made following a post-LOCA sump recirculation.
LOCA will be verified.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 16 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification 1.140 Rev. 1 Partial C.3,5,6 The provisions of these sections of the Adoption of a graded implementation guide, insofar as they pertain to the philosophy allows more flexibility in preoperational test program, will be adopting the testing provisions of the implemented in accordance to a graded guide to specific system applications.
approach to system testing as defined in This will allow the intent of the guide to Regulatory Position C.1 of Regulatory be satisfied while maintaining consistency Guide 1.68 Rev. 2. The relative importance with the testing philosophy applied to of each normal ventilation system shall be other plant systems.
determined by evaluating the function of Alden Research Laboratory has each system in: (1) minimizing offsite demonstrated by a scale model testing of releases (2) lowering exceptional exposures the recirculation sump that there is in accordance with ALARA principles.
adequate vortex control, proper system pressure drops and that adequate net positive suction head will exist post-LOCA for the sump recirculation.
1.133 Rev. 0 Adopted C.3.a. (1) Only these sections of the guide pertain to C.3.a. (2). (a) to the start-up test program and are thus See Section 1.7 for adopted for use.
position on other sections of the guide.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 17 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification 1.80 Rev. 0 Partial C.9 Loss of instrument air simulation of a Location of essential portions of the gradual freezing and plugging of instrument instrument air system at Catawba air lines will not be performed. precludes the possibility of freezing of any small amounts of water which might accumulate. Verification of dryer performance serves to assure that large amounts of water will not accumulate in the system. If plugging should occur, system response to a gradual loss of instrument air is bounded by the loss of instrument air test performed under C.8.b.
C.10 The loss of instrument air test will not be Valves which are not in their normal repeated with valves in their normal operating position when the loss of air operating procedure. test of C.8.b is performed will be verified to fail to their correct position. Any valve which is in other than its normal operating alignment when the test of C.8.b is performed and which is already in its failed position will be individually verified to fail to the correct position from its normal operating position following the test of C.8.b.
C.11 The results of instrument air testing will not Instrument air testing is conducted as a be included in the Startup Report. preoperational rather than a startup test.
1.52 Rev. 2 See Section 1.7 1.41 Rev. 0 Adopted 1.30 Rev. 0 See Section 1.7 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-1 (Page 18 of 18)
Regulatory Affected Guide Compliance Section(s) Exception Taken Justification 1.22 Rev. 0 See Section 1.7 1.20 Rev. 2 See Section 1.7 1.9 Rev. 0 See Section 1.7 and Section 14.4 1.68.2 Rev. 1 Adopted 1.95 Adopted 1.128 See Section 8.1.5.2 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 14-2 (Page 1 of 1)
Table 14-2. Preoperational and Startup Test Schedule (HISTORICAL INFORMATION - NOT TO BE REVISED)
Unit 1 Unit 2 Begin Preoperational Testing Begin Hot Functional Testing 7/15/85 Initial Fuel Loading 7/18/84 1/1/86 Initial Criticality 1/7/85 3/17/86 Begin Low Power Testing 1/7/85 3/17/86 Begin Power Ascension Testing 1/21/85 4/8/86 Commercial Operation 6/29/85 8/19/86 (22 OCT 2001)