|
---|
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217F9701999-10-14014 October 1999 Proposed Tech Specs,Incorporating ARC for Axial Primary Water Stress Corrosion Cracking at Dented Tube Support Plate Intersections ML20217E4301999-10-12012 October 1999 Proposed Tech Specs,Revising Requirements for Containment Penetrations During Refueling Operations ML20211M7341999-08-30030 August 1999 Marked-up & Revised TS Pages,Providing Alternative to Requirement of Actually Measuring Response Times ML20211K1721999-08-30030 August 1999 Proposed Tech Specs,Providing Clarification to Current TS Requirements for Containment Isolation Valves ML20209B7731999-06-30030 June 1999 Proposed Tech Specs Updating Requirmements for RCS Leakage Detection & RCS Operational Leakage Specifications to Be Consistent with NUREG-1431 ML20196F2211999-06-24024 June 1999 Proposed Tech Specs Pages for Amend to Licenses DPR-77 & DPR-79,allowing Use of Fully Qualified & Tested Spare Inverter in Place of Any of Eight Required Inverters ML20196G4701999-06-24024 June 1999 Proposed Tech Specs Pages Re Amends to Licenses DPR-77 & DPR-79,revising TS to Be Consistent with Rev to ISTS Presently Submitted to NEI TSTF for Submittal as Rev to NUREG-1431 ML20196G7961999-06-22022 June 1999 Proposed Tech Specs Bases,Clarifying Proper Application of TS Requirements for Power Distribution Systems & Functions That Inverters Provide to Maintain Operability & Providing Updated Info on Cold Leg Injection Accumulators ML20196G8071999-06-22022 June 1999 Revs to Technical Requirements Manual ML20195E9841999-06-0707 June 1999 Proposed Tech Specs,Increasing Max Allowed Specific Activity of Primary Coolant from 0.35 Microcuries/Gram Dose Equivalent I-131 to 1.0 Microcuries/Gram Dose Equivalent I-131 for Plant Cycle 10 (U2C10) Core ML20206E1611999-04-29029 April 1999 Proposed Tech Spec Change 99-04, Auxiliary Suction Pressure Low Surveillance Frequency Rev. Change Deletes Surveillance ML20206E1391999-04-29029 April 1999 Proposed Tech Spec Change 99-03, Main Control Room Emergency Ventilation Sys Versus Radiation Monitors. Changes Add LCOs 3.3.3.1 & 3.7.7 to Address Inoperability of Radiation Monitoring CREVS & NUREG-1431 Recommendations ML20204E8501999-03-21021 March 1999 Plant,Four Yr Simulator Test Rept for Period Ending 990321 ML20204H4081999-03-19019 March 1999 Proposed Tech Specs,Relocating TS 3.8.3.1,3.8.3.2,3.8.3.3 & Associated Bases Associated with Electrical Equipment Protective Devices to Technical Requirements Manual ML20207D6331999-02-26026 February 1999 Proposed Tech Specs Providing for Consistency When Exiting Action Statements Associated with EDG Sets ML20207D6011999-02-26026 February 1999 Proposed Tech Specs Relocating TS 3.7.6, Flood Protection Plan & Associated Bases from TS to Plant TRM ML20206S0131999-01-15015 January 1999 Proposed Tech Specs 3.3.3.3, Seismic Instrumentation & Associated Bases,Relocated to Plant Technical Requirements Manual ML20199K6001999-01-15015 January 1999 Proposed Tech Specs Adding New Action Statement to 3.1.3.2 That Would Eliminate Need to Enter TS 3.0.3 Whenever Two or More Individual RPIs Per Bank May Be Inoperable,While Maintaining Appropriate Overall Level of Protection ML20195H6111998-11-16016 November 1998 Proposed Tech Specs Revising EDG SRs by Adding Note That Allows SR to Be Performed in Modes 1,2,3 or 4 If Associated Components Are Already OOS for Testing or Maint & Removing SR Verifying Certain Lockout Features Prevent EDG Starting ML20154H7251998-10-0808 October 1998 Proposed Tech Specs Pages,Supplementing Proposed TS Change 96-08,rev 1 to Add CRMP to Administrative Controls Section & Bases of TS ML20238F1091998-08-27027 August 1998 Proposed Tech Specs Providing for Insertion of Limited Number of Lead Test Assemblies,Beginning W/Unit 2 Operating Cycle 10 Core ML20238F3001998-08-27027 August 1998 Proposed Tech Specs Replacing 72 H AOT of TS 3.8.1.1,Action b,w/7 Day AOT Requirement for Inoperability of One EDG or One Train of EDGs ML20209J1631998-08-0707 August 1998 Rev 41 to Sequoyah Nuclear Plant Odcm ML20236G5961998-06-29029 June 1998 Proposed Tech Specs Typed Pages for TS Change 95-19, Section 6 - Administrative Controls Deletions ML20249C6371998-06-26026 June 1998 Proposed Tech Specs Lowering Specific Activity of Primary Coolant from 1.0 Uci/G Dose Equivalent I-131 to 0.35 Uci/G Dose Equivalent I-131,as Provided in GL 95-05 ML20248F0051998-05-28028 May 1998 Proposed Tech Specs for Section 6, Administrative Controls Deletions ML20217N3511998-04-30030 April 1998 Proposed Tech Specs Pages,Modifying Surveillance Requirement 4.4.3.2.1.b to Change Mode Requirement to Allow PORV Stroke Testing in Modes 3,4 & 5 W/Steam Bubble in Pressurizer Rather than Only in Mode 4 ML20203J1681998-02-25025 February 1998 Proposed Tech Specs Pages,Revising EDG Surveillance Requirements to Delete Requirement for 18-month Insp IAW Procedures Prepared in Conjunction W/Vendor Recommendations & Modify SRs Associated W/Verifying Capability of DGs ML20202J7651998-02-13013 February 1998 Technical Requirements Manual ML20202J7141998-02-13013 February 1998 Proposed Tech Specs Adding New LCO That Addresses Requirements for Main Feedwater Isolation,Regulating & Bypass Valves ML20202J6961998-02-13013 February 1998 Proposed Tech Specs Incorporating MSIV Requirements to Be Consistent W/Std TS (NUREG-1431) ML20202J7601998-02-13013 February 1998 Proposed Tech Specs Section 3.7.9 Re Relocation of Snubber Requirements ML20198T4311998-01-21021 January 1998 Proposed Tech Specs Re New Position Title & Update of Description of Nuclear Organization ML20199F8231997-11-30030 November 1997 Cycle 9 Restart Physics Test Summary, for 971011-971130 ML20199K4571997-11-21021 November 1997 Proposed Tech Specs Adding one-time Allowance Through Operating Cycle 9 to Surveillance Requirement 4.4.3.2.1.b to Perform Stroke Testing of PORVs in Mode 5 Rather than Mode 4,as Currently Required ML20211A3191997-09-17017 September 1997 Proposed Tech Specs Re Pressure Differential Surveillance Requirements for Containment Spray Pumps ML20203B9731997-08-0505 August 1997 Rev 1 to RD-466, Test & Calculated Results Pressure Locking ML20217J5581997-07-31031 July 1997 Cycle Restart Physics Test Summary, for Jul 1997 ML20210J1671997-04-30030 April 1997 Snp Unit 1 Cycle 8 Refueling Outage Mar-Apr 1997,Results of SG Tube ISI as Required by TS Section 4.4.5.5.b & Results of Alternate Plugging Criteria Implementation as Required by Commitment from TS License Condition 2C(9)(d) ML20137T0871997-04-0909 April 1997 Proposed Tech Specs Re Elimination of Cycle 8 Limitation for SG Alternate Plugging Criteria ML20137M8581997-04-0101 April 1997 Proposed Tech Specs 2.1 Re Safety Limits & TS 3/4.2 Re Power Distribution Limits ML20137C8421997-03-19019 March 1997 Proposed Tech Specs Re Conversion from Westinghouse Electric Corp Fuel to Framatome Cogema Fuel ML20136J0381997-03-13013 March 1997 Proposed Tech Specs Section 5.6.1.2,revising Enrichment of Fuel for New Fuel Pit Storage Racks ML20134P8631997-02-14014 February 1997 Proposed Tech Specs Requesting Discretionary Enforcement for 48 Hours Which Is in Addition to 72 Hours Allowed Outage Time Provided by TS Action 3.8.1.1.b ML20134K9981997-02-0707 February 1997 Proposed Tech Specs Revising TS Change Request 96-01, Conversion from W Electric Corp Fuel to Framatome Cogema Fuel (MARK-BW-17), to Ensure That Core Analysis Computer Code Output Actions Are Consistent W/Hot Channel Factor SRs ML20138F2581997-01-17017 January 1997 Rev 39 to Sequoyah Nuclear Plant Odcm ML20134L9261996-11-0808 November 1996 Proposed Tech Specs Re Placing of Channel in Trip for Reactor Trip & Engineered Safety Feature Instrumentation Sys Solely to Perform Testing as Not Requiring Channel to Be Declared Inoperable ML20129D2661996-10-18018 October 1996 Proposed Tech Specs,Removing Existing Footnotes That Limit Application of Apc for Plant S/G Tubes to Cycle 8 Operation for Both Units ML20129G7301996-09-26026 September 1996 Proposed Tech Specs 3/4.3.3 Re Fire Detection instrumentation,3/4.7.11 Re Fire Suppression Systems & 3/4.7.12 Re Fire Protection Penetrations ML20134J9991996-09-23023 September 1996 Fuel Assembly Insp Program 1999-08-30
[Table view] Category:TEST/INSPECTION/OPERATING PROCEDURES
MONTHYEARML20196G8071999-06-22022 June 1999 Revs to Technical Requirements Manual ML20209J1631998-08-0707 August 1998 Rev 41 to Sequoyah Nuclear Plant Odcm ML20202J7651998-02-13013 February 1998 Technical Requirements Manual ML20138F2581997-01-17017 January 1997 Rev 39 to Sequoyah Nuclear Plant Odcm ML20134J9991996-09-23023 September 1996 Fuel Assembly Insp Program ML20138F2531996-02-23023 February 1996 Rev 38 to Sequoyah Nuclear Plant Odcm ML20108B4121995-11-0303 November 1995 Rev 37 to Sequoyah Nuclear Plant Odcm ML20094Q2301995-10-30030 October 1995 ASME ISI Valve Testing Program Basis Document, Rev 0 ML20094Q1931995-10-30030 October 1995 Rev 1 to ASME Sys Pressure Testing Program Basis Document ML20094Q1971995-10-30030 October 1995 Rev 1 to SG Tubing ISI & Augmented Insps, Rev 1 to 0-SI-SXI-068-114.2 ML20094Q2111995-10-30030 October 1995 Rev 1 to ASME ISI Pump Testing Program Basis Document ML20094Q1761995-10-13013 October 1995 ASME Section XI Isi/Nde & Augmented Nondestructive Exam Programs, SSP-6.10,rev 2 ML20094Q1841995-10-13013 October 1995 ASME Section XI Isi/Nde Program Units 1 & 2, Rev 0 to 0-SI-DXI-000-114.2 ML20149L6811994-09-30030 September 1994 Concerns Resolution Program - Sequoyah Nuclear Plant ML20063D6691994-01-24024 January 1994 Rev 4 to Sequoyah Nuclear Plant Restart Plan ML20065Q6251993-10-13013 October 1993 Rev 31 to Sequoyah Nuclear Plant Odcm ML20056F3181993-08-20020 August 1993 Rev 0 of Post-Restart Plan ML20056G1841993-08-10010 August 1993 Rev 2 to Sequoyah Nuclear Plant Restart Plan ML20044F3081993-05-20020 May 1993 Rev 0 to Sequoyah Nuclear Plant Restart Plan. ML18036B1961993-01-27027 January 1993 Rev 2 to Nuclear Power Training Procedure TRN-31, Fire Brigade Training. ML20127P5461992-12-16016 December 1992 Rev 20 to Surveillance Instruction SI-114.1, ASME Section XI ISI Program,Unit 1 ML20127P6361992-12-16016 December 1992 Rev 19 to Surveillance Instruction SI-114.2, ASME Section XI ISI Program,Unit 2 ML20114C8131992-04-17017 April 1992 Rev 27 to Odcm ML20101F2081992-02-0808 February 1992 Rev 16 to Surveillance Instruction SI-114.2, ASME Section XI ISI Program Unit 2 ML20101F2011992-02-0808 February 1992 Rev 17 to Surveillance Instruction SI-114.1, ASME Section XI ISI Program Unit 1 ML19332D2241989-09-22022 September 1989 Rev 23 to Odcm. ML20245H1421989-08-15015 August 1989 Rev 22 to Offsite Dose Calculation Manual Changes ML20246H4861989-05-16016 May 1989 Rev 1 to Technical Instruction TI-115, Instructions for Sewage Mgt ML20244E2521989-04-28028 April 1989 Rev 14 to Surveillance Instruction SI-114.2, ASME Section XI Inservice Insp Program ML20246E7771989-04-25025 April 1989 Rev 14 to Surveillance Instruction SI-114.1, ASME Section XI Inservice Insp Program ML20206D3421988-10-15015 October 1988 Rev 13 to Surveillance Instruction SI-114.2, Inservice Insp Program ML20154J3221988-09-0909 September 1988 Procedure EA-OR-003-S, Sequoyah Nuclear Plant - Unit 1 Design Baseline & Verification Program,Supplemental Engineering Assurance Oversight Review Rept ML20150F9291988-06-17017 June 1988 Diesel Generator Voltage Response Improvement Plan ML20154L1421988-05-0909 May 1988 Rev 3 to Revised Sequoyah Nuclear Performance Plan ML20153H3891988-03-30030 March 1988 Rev 19 to Sequoyah Nuclear Plant Offsite Dose Calculation Manual ML20196G3301988-02-24024 February 1988 Limited Test Program for Determining Axial Load Capacity of Cast One-Hole Conduit Clamps ML20147E9511988-01-21021 January 1988 Revised, Procedures Generation Package ML20153H3851988-01-0505 January 1988 Rev 18 to Sequoyah Nuclear Plant Offsite Dose Calculation Manual ML20147E5691987-12-17017 December 1987 Rev 4 to Special Maint Instruction SMI-0-317-61, Instrumentation Features Walkdown,Rework & Insp Instructions for CAR 87-014 ML20237C5451987-10-28028 October 1987 Rev 17 to Offsite Dose Calculation Manual ML20236E6931987-10-17017 October 1987 Rev 2 to Engineering Organization & Operating Procedures, TVA Employee Concerns Special Program ML20235X0551987-10-10010 October 1987 Rev 0 to Sys Operating Instruction SOI-74.2, Removal of RHR for Repair of 2-FCV-74-2 ML20235X0651987-10-10010 October 1987 Rev 0 to Special Maint Instruction SMI-2-74-1, Repair of 2-FCV-74-2 ML20237H2051987-08-28028 August 1987 Rev 0 to Civil Engineering Branch Instruction CEB-CI 21.89, Mod Priorities for Pipe Supports on Rigorously Analyzed Category I Piping - Sequoyah Unit 2 ML20207G5821987-08-24024 August 1987 Rev 14, Balance of Plant Temp Monitoring Sys ML20237L3651987-08-21021 August 1987 Unit 2,Regeneration of Support Design Calculations on Rigorously Analyzed Piping,Program Plan ML20236Q1771987-08-0707 August 1987 Rev 9 to Surveillance Instruction SI-114.1, ASME Section XI Inservice Insp Program. Rev Corrects Deficiencies Shown in Caqr CH5870006 & CH5870010,adds Punchlist & Incorporates Icf 87-708 ML20236N7601987-08-0606 August 1987 Revised Radiological Emergency Plan Implementing Procedures, Including Rev 13 to IP-8, Personnel Accountability & Evaluation & Rev 6 to IP-15, Emergency Exposure Guidelines ML20236M7181987-07-28028 July 1987 Rev 1 to WP-17-SQN, Vendor Weld Quality, Welding Project, TVA Employee Concerns Special Program ML20236M7351987-07-24024 July 1987 Rev 5 to 80503-SQN, Document Distribution Control, Element Rept,Tva Employee Concerns Special Program 1999-06-22
[Table view] |
Text
.:
O 4
't TENNESSEE '!ALLE( AUTHORITY
- b l
SEQUDYAH NUCLEAR PLANT EMERGENCY CONTINGENCY INSTRUCTION ECA-0.0 LOSS OF ALL AC PCHER Revision 1 PREPARED BY: G. Strickland RESPONSIBLE SECTICN: Ocerations REVISED BY: G. L. -erestra SUBMITTED SY:, <wt. ?IM.fbL_
[Jescondble section supervisor PORC REVIEW DATE: JW 2 41987 APPROVED SY- *-
- Plant Manager DATE APPROVED: J 24 037 Reason for revision (include all Instruction Change Form Nos.):
Revisea to add REP steo I due to ICF B6 1473 and for enhancement /c!aci'i-catien to step 23 (V and V comment) and steo 21 cart 6 'revisea to serie/
SFP levol \bove alarm) and ste M 8/20 refarencea 501-88.1 for sentainment isolation.
The last page of this instruction is numoer: 36 I
B708170471 870811 l PDR ADDCK 05000327 l p PDR I
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ ____ E
SEQUOYAH NUCLEAR PLANT PLANT INSTRUCTION REVISION LOG EMERGENCY CONTINGENCY INSTRUCTION ECA-0.0 REVISION Date Pages REASON FOR REVISION (INCLUDE COMMIT-LEVEL Approved Affected HENTS AND ALL ICF FORM NUMBERS)
New instruction written per commitment 0 08/21/85 ALL to NUREG 0737.
Revised to add REP step 1 due to ICF 86-1473 and for enhancement / clarification to step 3 (V and V comment) and step 21 part 6 (@ revised to verify SFP level JUN 2 4 1987 above alarm) and steps 18/20 referenced 1 ALL SOI-88.1 for containment isolation.
l l
0348L/ldw L______ _ _ _ _ . _ - - - - - - - - - - - - - - - - - - - - - - -
SQNP ECA-0.0 Unit 1 or J Page 1 of 16 Revision 1 l
l l
LOSS OF ALL AC POWER A. Pl!RPOSE l
l
- 1. Isolate RCP seal cooling
- 2. Cooldown RCS using Turbine-driven AFW Pump
- 3. Inject UHI and cold leg accumulators 1.
When AC power restored, then establish ERCW flow and go to recovery guideline B. SYMPTOMS Loss of all AC power C Tit ANSITION FROM OTlIER INSTRUCTIONS ~
E-0. React 6r Trip Or Safety injection ES1ERGl'! A lite
SQNP ECA -0. 0 Unit l'or 2 Page 2 of 16 Revision 1 l s l LOSS OF ALL AC POWER 1
STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED Note- The purpose of this ECA should be reviewed prior to using this instruction Note:
l Status Trees should be monitored for int'ormation only.
FRGs should not be implemented l
1 Initiate REP Per IP-1
- a. Notify Operations Duty Specialist within 5 min 2
Check If RCS is isolated
- a. Pzr PORVs - CLOSED t. IF_ RCS press
< 2035 pcig, TilEN close PORV
- b. Verify letdown isolated
- 1) FCV-62-69, 70, 72, 73, and 74 - CLOSED
- c. Verify excess letdown isolated l
- 1) FCV-62-54 or 55 -
CLOSED i
1 E.TIElt GP/A nac
SQNP ECA-0.0 Unit 1 or 2 Page 3 of 16 y Revision 1
.(
LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 3 Verify Turbine-Driven IF AFW flow ( 440 gpm, AFW Flow > 440 gpm TIIEN verify turbine runnmg and establish
- a. Dispatch operator with AFW flow radio to LCV handwheels
- a. Locally open turbine-driven LCV
- 1) S/G 2 and 3 LCV handwheel at Aux Bldg 714
- 2) S/G 1 and 4 LCV handwheel in West Valve Room 1
Trv To Restore Power To
'[ Shutdown Boards ~
- a. Refer to AO! .35, Loss Of Of fsite Power
- b. Energize shutdown boards from D/G
- 1) Start D/G 1) Emergency start D/G
- 2) Verify shutdown boards energized 2) IF shutdown boards can NOT be energized.
THEN trip D/G and energize shtttdown boards from offsite power e
IF at least one shutdown board energized, TilEN return to instruction tTeifect DIERGP/A -a-
. sac
a.
SQNP EC A-0. 0 Unit 1 or 2 Page 1 of 16 -
Revision 1
~ LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: When power is restored to a shutdohn board. recovery actions should continue with step 22 ,
1 CAUTION: If an SI signal exists or if an SI signal is aEtuated during this guideline, it should be reset to permit manual loading of equipment on' shutdown board l
CAUTION :
An ERCW pump shoilld be kept available to automatically load on the shutdow6 board to provide diesel generator cooling
\ ,.
5 . Place Following Eguipment , , !
JMULL TO LOCK
- a. CCPs _ ( 4
- b. S1 Pumps
- c. RHR Pumps
- d. Cntmt Spray Purrps > >
- e. >1otor-driven AFW Pumps
- f. CCS Pumps
- g. Pzr heaters s 4
i-EMERCP/A t-ne
!- i 1
l l* SQNP EC A-0. 0 Unit 1 or 2 Page 5 of 16 Revision 1
<u l LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l'
6 Isolate RCP Seal Cooling
- a. Close RCP number 1 seal leakoff valves
- 1) FCV-62-9
- 2) FCV O-22
- 3) FCV-62-35
- 4) FCV-62-48
- b. Isolate RCP seal return using reach 3 ods outside seal water filter cubicle
- Aux Bldg 690 l)62-642 seal water ~
filter inlet isolation
- 2)62-643 seal water filter bypass isolation
- c. Isolate RCP seal injection using reach rods outside seal water injection filter cubicles - Aux Bldg 690
- 1)62-546 seal water injection tilter bypass isolation '.
1
- 2)62-549 seal water injection filter A i
outlet isolation
- 3)62-550 seal water injection filter B outlet isolation
- Step Continued On Next Page -
I 1
l I
ENEllGP/A I
. tac i
l
t-SQNP ECA-0.0 Unit 1 or- 2 Page 6 of 16 f
4.
Revision 1 L
LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
'G lsolate RCP Seal Cooling (cont. )
- d. Isolate RCP thermal barrier - Aux Bldg 714
- 1) 70-677A thermal barrier booster pump A discharge isolation
- 2) 70-677B thermal i barrier booster pump B' discharge isolation 7 Isolate CST From Hotwell
- a. Isolate condenser vacuum drag LCV-2-9 using _
manual valves next to number 7 heater drain tank - Turb Bldg 685
- 1) 2-522 auto makeup LCV isolation
- 2) 2-525 auto makeup LCV bypass isolation 8 Check S/G Status
- a. CLOSE SISIV and bypass valves
- b. Verify 31FW reg and bypass valves - CLOSED
- e. Verify S/G blowdown valves - CLOSED D1ERGP/A -n-aac
I
- c. SQNP ECA-0.0 Unit 1 or 2 Page 7 of 16 Revision 1
. \:
LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONS3 NOT OBTAINED C AUT40N: A faulted or ruptured S/G that is isolated saould remam isolated. Steam supply to the Turbine-driven AFW Pump must be maintained from one S/G i
9 Check S/G Press IF S/G press low and decreasing in an
- a. Press in all S/Gs within uncontrolled manner, 100 psi of each other THEN isolate faulted S/G:
- b. All S/G press - STABLE Isolate AFW flow OR INCREASING Ensure Turbine-driven AFW Pump being supplied from intact S/G Ensure S/G PORV closed
!O Check Secondary Side IF secondary side radiation Radiation - NOTIMAL - Is high, TIIEN isolate ruptured S/G:
- a. Dispatch Rad Con to monitor secondary plant Isolate AFW flow Ensure Turbine-driven AFW Pump being supplied from intact S/G WHEN S/G press
< 1040 psig, THEN verify S/G PORV closed EMERGP/A -7 ilac
i SQNP ECA-0.0 Unit 1 or 2 Page 8 of 16 Revision 1 1
LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED ;
11 Check Intact S/G Levels
- a. Narrow range S/G a. Verify AFW flow and levels > 25% S/G 1evels returning to normal program
- b. Control S/G levels b. IF any S/G level between 25% and 50% continues to increase with no AFW tiow, THEN isolate ruptured S/G:
Isolate AFW flow Ensure Turbine-driven AFW Pump being supplied from
.. intact S/G WHEN S/G press
< 1010 psig, THEN verify S/C PORV closed I
1 i
EMERGP/A -H-JitC
SQNP EC A-0,0 Unit 1 or J Page 9 of 16 i' e - Revision 1
- i.
LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 12 Check DC Bus Loads
- a. Notify TSC to evaluate shedding all large non-essential DC loads
- b. Monitor. DC power supply
- 1) '125-V DC vital batteries
- 2) 250-V DC batteries
- c. Evaluate use of fifth vital battery 13 Check CST Level .
- a. CST _ level > 3 ft a. !F CST level < 3 f t.
THEN open ERCW valves to Turbine-driven AFW Pump suction and monitor pump suction press
- b. CST level > 17 ft
" b. IF CST level < 17 ft, TiiEN attempt to refill
~
CST while continuing with this instruction e Notify TSC to evaluate alternate means of supplying AFW suction i
j E31EltGl'/A aac
4 SQNP EC A-0.0 Unit I or 2 Page 10 of 16
/
Revision I ils LOSS OF ALL AC POWER l
STEP' i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAl'!ED l l
l CAUTION: RCS pressure should not be decreased to less than 180 psig to prevent nitrogen injection from cold leg accumulators 1
Note: RCS cooldown and depressurization should be as quickly as }
possible to minimize RCS inventory loss out RCP seals {
j i
Note: {
Pzr level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/Gs. Depressurization {
j should not be stopped to prevent these occurrences q
'ote:
S/Gs I and 4 PORVs have reach rods in 180-V SD Bd :icem 14 Cooldown RCS ..
- a. Maintain intact S/G level i. !!ecover level in at
> 10% (25% FOR ADVERSE least one intact 3/G CNTMTl in S/Gs used for prior to cooldown cooldown
- b. Maintain T-cold > 270 F
- c. Rapidly dump steam using S/G PORVs
- 1) Dispatch operator with radio to S. G PORVs 1
- 2) Normal 100"F/hr cooldown limit not required
- d. Control S/G l'ORVs to maintain RCS press
> 180 psig I
ENER GI" A aae l
l l
______________J
SQNP EC A-0.0 Unit 1 or 2 Page 11 of 16 Revision 1
.A LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED 1 15 During RCS Cooldown, Allow 1 Accumulators To inject .)
i l
- a. UHI i
- 1) Injection starts at
~ 1200 psig-
- 2) AFTER UHI injection, ensure UHI isolation valves closed
- 3) Locally gag isolation valves
- b. Cold leg accumulators
- 1) Injection starts at
~400 psig ~
l
- 2) ~ Maintain RCS press
> 180 psig to prevent nitrogen injection 16 Check Reactor Suberitical IF reactor returning critical,
- a. Intermediate range - TllEN evaluate tf reactor can be shutdown by injecting ZERO OR NEGATIVE accumulators or stopping RCS STARTUP RATE couldcwn
- b. Source range -
ZERO OR NEGATIVE STARTUP HATE ENEllGP ' A aue
SQNP ECA-0.0 Unit 1 or 2 Page 12 of 16
_.- Revision 1
(-
A LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 17 Actuate Phase A And Cntmt V
' ent Isolation
- a. Manually actuate phase A and entmt vent isolation
- b. IF SI actuated, TIIEN reset SI after 1 min l time delay )
18 Check Phase A And Cntmt Vent Isolation i
- a. Refer to SOI-88.1 A and !
I SOI-88.1B
- b. Notify TSC 'to evaluate any . . .
isolation valve which failed to clos _e ENERGl'/A aac
SQNP ECA-0.0 Unit 1 or 2 Page 13 of 16 Revision 1 t
A LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE .'iOT 03TAINED 19 Check Cntmt Hydrogen Control Equipment -
- a. Place hydrogen ignitors to off
- b. Place hydrogen recombiners in standby with power out switch to off ,
i 20 Check Cntmt Press < 2.31 nsig IF cntmt press > 2.31 psig ,
TIIEN :
l
- a. Check phase B isolation per 501-58.1C
' b. Notifv TSC *.o evaluate any isolation valve which faile,t to close
- c. Reset cntat spray DIEltGP,A
-13 n;\c
SQNP EC A-0.0 Unit I or 2 Page L4 of 16 Revision 1 LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE ItESPONSE NOT GBTAINED l
21 Check If AC Power Restored l
3 a. IF both shutdown boards a. IF at least one shutdown l 7011 de-energized, Toard energized.
, rIEN repeat steps 13 TIIEN continue with next to 21 while checking s tep status of following:
- 1) AC power restoration l
l
- 2) RCP seal isolation
- 3) DC power supply
- 4) SIT temp .> 145 F or evaluate diluting BIT i i
- 5) B AT temp > 145 F or ,
l evaluate diluting B AT
- 6) SFP' level above bottom rung on local level indicator on the west wall 22 Stabilize S/G Press DIERGP/A - 1 1-aac
_ _ _ _ _ _ _ _ . _ _ _ _ _ _ . -- - - - - - - - - - - - - - - - - - - - - - - - - - --------- --------~ ----- -- --~ -- -
SQNP EC A-0. 0 Unit .I or J Page 15 of 16
/ Revision 1 f
1 A-LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED C AUTION: The loads placed on the energized AC board should not exceed the capacity of the power supply 23 Verify Following Equipment Energized On Shutdown Board
- a. 480-V shutdown boards L. 125-V vital battery charger e 120-V AC vital inst power bds
- d. Standby lighting
- e. 250-V DC battery charger
- f. 48-V DC telephone charger ~
- g. 24-V DC microwave charger 21 Verify ERCW Operation
- a. Verify ERCW Pump -
RUNNING
- h. Verify ERCW to D/G EMEllGP/A .litC
SQNP ECA-0.0 Unit 1 or 2 Page 16 of 16 Revision 1
, LOSS OF ALL AC POWER STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAISED 25 Select Recovery Guideline
- a. IF the following a. IF either of the two E6nditions are met: Eriteria are NOT
" satisfied ,
1.) RCS subcooling THEN go to
- > 40 F RnCOVERY FROM LOSS OF ALL AC POWER
- 2) Pzr level ,> 20% WITH . SI . REQUIRED
[50% FOR ADVERSE CNTMT]
THEN go to ECA-0.1, Recovery From Loss Of All AC Power Without SI Required
~
-END-DIERGP/A aae
________________-___-____