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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
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ENCLO2URE 1 SEQ'JOYAH NUCLEAR PLANT - UNIT 2 REGENERATION OF SUPPORT DESIGN CALCULATIONS ON RIGOROUSLY ANALYZED PIPING PROGRAM PLAN TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 OBJECTIVES
3.0 BACKGROUND
4.0 PROGFAM LOGIC 5.0 . REVIEW AND REGENERATION OF CALCULATI e3 6.0 PRIORITI8S 7.0 PIPI: EUPPOET DESIGN CRITERIA P0 ACCEPTANCE CRITERIA l
0708280072 870821 Pl:P ADOCK 05000327 P PDR
SEQUOYAH NUCLEAR PLANT - UNIT 2 AND COMMON REGENERATION OF SUPPORT DESIGN CALCULATIONS ON RIGOROUS ANALYSIS PIPING
1.0 INTRODUCTION
This plan provides a description of the program for review and regeneration of pipe support calculations associated with rigorously analyzed category 1 piping systems (referred to as " supports").
2.0 OBJECTIVES The overall program for the review and regeneration of pipe support calculation associated with rigorously analyzed piping includes two major objectives:
- a. Determination of ovecall compliance of existing pipe support calculations to licensing commitments.
- b. Reg 6neration of all pipe support calculations that are not retrievable or are inadequate as determi-wd in activity A.
3.0 BACKGROUND
The TVA calculation validation program indicates that approximately 6,000 category 1 pipe support calculations will require evaluation / regeneration. The majority of the pipe support calculations generated during the plant construction phase were performed by two vendors: EDS/Impell designing pipe supports inside containment, and f Basic Engineering designing supports outside containment. The balance of the Sequoyah Nuclear Plant (SQN) supports were designed by TVA.
Subsequent to receipt of an operating license, design modifications affecting supports have been performed by TVA.
4.0 PROGRAM LOGIC The plan to establish the pipe support design adequacy involves the following logic:
- quantify and document total scope
- prioritize overall scope for regeneration
- collect pipe support design input
- verify the pipe support as-installed function
- evaluate existing support calculations
- regenerate noneetrievable calculations Figure 1 indicates the program logic.
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-g-5.0 REVIEW AND REGENERATION OF CALCULATIONS l 5.1 Definition of Scope and Collection of Input Information i 5.1.1 Identification of Pipe Support Calculations on Rigorously Analyzed Category 1 Piping Based on a definition of category 1 system boundarios and an index of all rigorously analyzed piping, stress calculations for category 1 systems will be determined. A set of stress analysis problem connectivity diagrams will be developed to list and show connectivity of rigorous stress analysis problems. The support locations will be identified from the mathematical models (stress problem isometric sketches), and all pipe supports on rigorously analyzed category 1 piping will be identified. Additionally, pertinent documentation for each support and analysis problem will be identified and retrieved through the use of the TVA database " Calculation Cross Reference Index System" (CCRIS) and " Records Information Management System" (RIMS).
5.1.2 Functional Verification A field review will be performed for functional verification of all pipe supports by a walkdown of the supports for critical attributes (e.g., directionality, type, location, and etc.). Additionally, the field verification will identify all other attachments on category 1 hangers. The field verification data will be utilized in review and regeneration activities.
5.1.3 Collection of pipe Support Design Input A review will be performed to establish and collect piping loads and movements and other applicable design input for use in the pipe support calculation review and regeneration effort.
5.2 Screening of Pipe Support Calculations At our meeting on June 19, 1987, TVA proposed a screening criteria for prioritizing support evaluations. The purpose of this screening was to segregate the supports that may require modification so that the corresponding calculations could be reviewed / regenerated and resulting modifications identified and installed on a priority basis.
Based on subsequent implementing experience, TVA has determined that the above-described screening was not resource effective. Efforts spent segregating calculations could be better utilized performing necessary support evaluations, thereby ensuring timely resolution of this critical issue.
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1 Accordingly, SQN pipe support calculations will not be evaluated with the recommended priority system based on screening attributes. )
Rather, all evaluations for inside containment supports and outside '
containment cupports will be completed and modifications identified before mode 4. Program documentation and report generation will be finalized by October 19, 1987.
The evaluation and regeneration of the pipe supports will utilize the pipe support design criteria described in section 8.0.
5.3 Pipe Support Calculation Evaluation / Regeneration 5.3.1 Review of Existing Calculations RIMS and business records of contractors involved in the uriginal design will be accessed for all pipe support calculations. All retrieved calculations will be assessed for technical and documentation adequacy and will be reviewed for applicable loading and specific configuration attributes, including data obtained through field verification activity.
5.3.2 Regeneration of Pipe Support Calculations That Are Not l Retrievable or Are Inadequate Support calculations that are inadequate or cannot be retrieved will be regenerated using calculation of record loads, applicable licensing commitments, and design criteria.
5.4 Resolution of Discrepancies Any modifications to pipe support designs identified as a result of this plan will be documented and evaluated in accordance with TVA Quality and Engineering Assurance Procedures. I 6.0 PRIORITIES The priorities for the various tasks and activities involved in this program are arranged in such a manner as to support SQN unit 2 restart.
All support calculations which require modifications inside unit 2 steel containment vessel will be identified and completed as a first step.
This will be followed by evaluation of all support calculations outside unit 2 containment.
7.0 PIPE SUPPORT DESICN CRITERIA The pipe support design criteria established for this review and regeneration effort are consistent with TVA's interpretation of the Final Safety Analysis Repore (FSAR), with the following addition.
Manufacturers' Standardization Society Standard Practice (SP)-S8 design criteria will be used for all standard component supports for which appropriate component load rating data are available. For those
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_4-components that SP-58 data are not'available, e.g., snubbers, struts, ar.d clamps procured to American Society of Mechanical Engineers (ASME). ,
subsection NF code requirements, the methodology and load rating data of NF will be utilized.
The governing code for piping evaluation, in accordance with section 3,9.2 of the FSAR, is ANSI B31.1-1967 Code. Accordingly, structural shapes (supplementary steel in accordance with ANSI B31.1) will comply with the rules 'specified in the American Institute of- Steel Construction
" Manual of Steel Construction," 8th edition (1980), for'" Normal" and
" Upset" loading conditions. Section 3.8 of the FSAR, which limits stresses in steel structures under Safe Shutdown Earthquake loading, will i be utilized for the faulted loading condition.
-Detailed definitions of the requirements used in the pipe support calculation regeneration / evaluation program are given in detail in TVA procedure SQN-DC-V-24.2.
8.0 ACCEPTANCE CRITERIA Supports that do not meet the criteria in section 7.0 will be evaluated in accordance with TVA's Nuclear Engineering Proceudre (NEP)-9.1 (Corrective Action) to determine the corrective action to be taken and
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whether this action is required before restart. The engineering design modification and supporting calculation will be completed'before the restart /postrostart determination and used as input to the NEP-9.1 process.
The restart criteria contained in Volume 2 of the Nuclear Performance Plan will be utilized as the basis of the pre- or postrestart determination.
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ENCLOSURE 2 T/A COMMITMENTS PROGRAM DESCRIPTION FOR UNIT 2 REGENERATION OF PIPE SUPPORI' CALCULATIONS ON RIGOROUS AllALYSIS PIPING
- 1. Unit 2 - Evaluate pipe support calculations and generate support modification packages for inside and outside containment systems before mode 4.
- 2. Unit 2 - Complete program documentation and report for pipe support regeneration effort by October 19, 1987.
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AUG 11 '87 13:04 (615) 632-4519
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- 1. : ' ' ~ FIGURE 1 Sequoyah Nuclear Plant . Unk 2 & Common, PIPE SUPPORT CALCULATION REGENERATION
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LOGIC DIAGRAM TOTAL SCCPE I
1f FUNCTIONAL VERIFICATION
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. i DATA COLLECTION 1
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SUPPORTS WITH SUPPORTS WITH EXISTING CALCULATIONS '
$8 SING CALCULATIONS '
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YES CALCULATIONS , f COMPLETE &
CLOSURE ACCEPTABLE 7 <
J N_O g REGENERATION OF CALCULATIONS
, f YES I
%- ACCEPTABLE 7 CLOSURE ,
1 NO q r I MODIFICATION ENGINEERED CAOR
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