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Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20148U0111997-06-17017 June 1997 Confirmatory Survey of Group E Effluent Discharge Pathway Areas Fsv Nuclear Station Platteville,Co ML20133D7661996-09-16016 September 1996 Confirmatory Survey Plan for Fsv Nuclear Station Decommissioning Project ML20129A4621996-09-11011 September 1996 Rev 0 to Fsv Decommissioning Project Final Survey Requirements for Liquid Effluent Pathway ML20100L3581996-02-22022 February 1996 Proposed Tech Specs,Submitting Corrected Version of Plant Decommissioning TS Updated to Reflect All Approved Amends ML20097C2601996-01-17017 January 1996 Confirmatory Survey Activities Plan for Fsv Nuclear Station PSC Platteville,Co ML20101F2091995-09-18018 September 1995 Issue 7 to DPP 5.4.2, Odcm ML20084B8801995-05-25025 May 1995 Rev 1 to Fsv Nuclear Station Decommissioning Project Final Survey Plan for Site Release ML20084B6881995-05-10010 May 1995 Issue 5 to Fire Protection Operability Requirements (Fpor) FPOR-7, Fire Extinguishers ML20082K2461995-04-14014 April 1995 Proposed Decommissioning Tech Specs Administrative Control 5.3.1,reflecting Organizational Changes That Impact Membership of Decommissioning Safety Review Committee ML20082T2151995-04-12012 April 1995 Issue 7 to Fire Protection Operability Requirements (Fpor) FPOR-12, Fire Detectors ML20082B9411995-03-17017 March 1995 Confirmatory Survey Plan for Repower Area,Fort St Vrain, Platteville,Co ML20082C0801995-03-16016 March 1995 Proposed Confirmatory Survey Plan for Repower Area,Fort St Vrain,Platteville,Co ML20082B9821995-03-15015 March 1995 Instrumentation Comparison Plan Between Orise & Fort St Vrain ML20086S2471995-02-0909 February 1995 Decommissioning Plan for Fort St Vrain Nuclear Generating Station ML20077C7171994-11-30030 November 1994 Issue 9 to FPOR-14, Fire Protection Operability Requirements ML20078C0641994-10-12012 October 1994 Revised Fire Protection Operability Requirements,Including Issue 21 to Depp Table of Contents,Issue 2 to FPOR-22 & Issue 3 to FPOR-23 ML20081J7951994-09-15015 September 1994 Issue 5 to DPP 5.4.2, Odcm ML20063M1551994-02-17017 February 1994 Rev 0 to Fsv Nuclear Station Decommissioning Project Final Survey Plan for Site Release ML20057A6151993-08-30030 August 1993 Issue 2 to FPOR-23, Fire Water Makeup Sys ML17291B3261993-05-18018 May 1993 Proposed TS Section 2.2 Re Activated Graphite Blocks,Section 2.4 Re Channel Calibr & SR 3.2.1 Re Verification of Reactor Bldg Pressure & SR 3.2.2 Re Verification of Pressure Drop Across Each HEPA Filter ML20118B2141992-09-25025 September 1992 Proposed Decommissioning Tech Specs Replacing Radiation Safety with Nuclear Safety, Revising Applicability Requirements for Specs Dealing W/Reactor Bldg Confinement Integrity & Clarifying Items Re Unreviewed Safety Questions ML20114D6871992-09-0101 September 1992 Tritium Leach Test on H-327 Graphite ML20096H1961992-05-19019 May 1992 Decommissioning TS Deleting Section 4.2.15 Re LCO 4.2.15 Covering Pcrv Cooling Water Sys Temps ML20095B0951992-04-14014 April 1992 Proposed Tech Specs Re Organization,Review & audit-administrative Controls ML20094L2831992-03-19019 March 1992 Proposed Tech Specs Re Controls & Limits Appropriate for Decommissioning ML20086C5301991-11-15015 November 1991 Proposed Tech Spec Limiting Condition for Operation 4.2.15 Re Pcrv Cooling Water Sys Temps ML20079M4261991-10-11011 October 1991 Revised Abnormal Operating Procedures,Reflecting Deletion of Issue 9 of EP Class ML20091D7631991-10-11011 October 1991 Proposed,Revised Limiting Condition for Operation 4.2.15 Re Prestressed Concrete Reactor Vessel Cooling Water Sys Temp ML20082L9311991-08-30030 August 1991 Proposed Tech Specs Re Decommissioning ML20082H8851991-08-16016 August 1991 Issue 2 to Abnormal Operating Procedure AOP-I-2, Chemical, Petroleum & Hazardous Waste Spill Response ML20091C4141991-08-0202 August 1991 Issue 58 to Abnormal Operating Procedure AOP-L, Loss of Instrument Air Header ML20024H3341991-05-10010 May 1991 Nonproprietary Rev 2 to FSV-P-SCP-100, Fort St Vrain Initial Radiological Site Characterization Program Program Description ML20072V5291991-04-12012 April 1991 Revised Defueling Emergency Response Plan,Including Section 1 Definitions,Section 2 Scope & Applicability,Section 3 Summary of Fsv Derp,Section 4 Emergency Classifications & Section 5 Emergency Organization ML20070V6871991-03-20020 March 1991 Issue 55 to Abnormal Operating Procedure AOP-R, Loss of Access to Control Room ML20072S0521991-03-15015 March 1991 Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 9 to CMIP-11, Classification of Emergency for McGuire Nuclear Station & Rev 11 to CMIP-12, Classification of Emergency for Oconee Nuclear Station ML20066J1171991-02-15015 February 1991 Issue 56 to Intro Section of Abnormal Operating Procedure (Aop),Issue 58 of AOP-A,Issue 58 to AOP-B,Issue 56 of AOP D-1 & Issue 2 of RERP-TRANSPORTATION ML20066A3781990-12-21021 December 1990 Proposed Decommissioning Tech Specs Re Reactor Bldg Integrity,Reactor Bldg Ventilation Exhaust Sys,Radiation Monitoring Instrumentation & Pcrv Shielding Water Tritium Concentration ML20059L5891990-09-14014 September 1990 Proposed Tech Specs Changing Design Features Section 6.1 to Permit Removal of CRD & Orifice Assemblies from Core Regions Defueled in Support of Plant Closure Activities ML20058N1071990-08-10010 August 1990 Issue 56 to AOP-I, Discussion of Fire ML20042F3151990-04-26026 April 1990 Proposed Tech Specs Re Defueling ML20006B7921990-01-25025 January 1990 Proposed Tech Specs Re Administrative Title Changes to Section 7.1 ML19332E8901989-12-0404 December 1989 Proposed Tech Specs Re Reactivity Control & Control Rod Pair Position Requirements During Shutdown ML19332F3561989-12-0404 December 1989 Proposed Tech Specs Re Limiting Condition for Operations 4.7.3, Fuel Storage Wells & 4.7.5, Instrumentation. ML19332C8151989-11-21021 November 1989 Proposed Tech Specs Revising Items 2.D.(1) & 2.D.(4) Re Max Power Level & Early Shutdown,Respectively ML20064B2111989-11-0909 November 1989 Fort St Vrain Cycle 4 RT-500L Test Rept ML19324B6961989-10-30030 October 1989 Proposed Tech Specs Re Reactor Core & Reactivity Control ML19327B1321989-10-13013 October 1989 Proposed Tech Specs,Reflecting Deleted Limiting Conditions of Operations 4.1.2 Through 4.1.6,deleted Surveillance Requirements 5.1.1,5.1.2,5.1.3 & 5.1.5 & Newly Added Reactivity Control Section ML19351A3271989-10-13013 October 1989 Proposed Tech Specs 6.1 Re Defueling Phase Document Design Features ML20248G4731989-10-0101 October 1989 Proposed Tech Specs Re End of Operations ML20248G4461989-09-30030 September 1989 Proposed Tech Specs Re Chlorine Detection & Alarm Sys & Control Room Emergency Ventilation Sys 1997-06-17
[Table view] Category:TEST/INSPECTION/OPERATING PROCEDURES
MONTHYEARML20148U0111997-06-17017 June 1997 Confirmatory Survey of Group E Effluent Discharge Pathway Areas Fsv Nuclear Station Platteville,Co ML20133D7661996-09-16016 September 1996 Confirmatory Survey Plan for Fsv Nuclear Station Decommissioning Project ML20129A4621996-09-11011 September 1996 Rev 0 to Fsv Decommissioning Project Final Survey Requirements for Liquid Effluent Pathway ML20097C2601996-01-17017 January 1996 Confirmatory Survey Activities Plan for Fsv Nuclear Station PSC Platteville,Co ML20101F2091995-09-18018 September 1995 Issue 7 to DPP 5.4.2, Odcm ML20084B8801995-05-25025 May 1995 Rev 1 to Fsv Nuclear Station Decommissioning Project Final Survey Plan for Site Release ML20084B6881995-05-10010 May 1995 Issue 5 to Fire Protection Operability Requirements (Fpor) FPOR-7, Fire Extinguishers ML20082T2151995-04-12012 April 1995 Issue 7 to Fire Protection Operability Requirements (Fpor) FPOR-12, Fire Detectors ML20082B9411995-03-17017 March 1995 Confirmatory Survey Plan for Repower Area,Fort St Vrain, Platteville,Co ML20082C0801995-03-16016 March 1995 Proposed Confirmatory Survey Plan for Repower Area,Fort St Vrain,Platteville,Co ML20082B9821995-03-15015 March 1995 Instrumentation Comparison Plan Between Orise & Fort St Vrain ML20086S2471995-02-0909 February 1995 Decommissioning Plan for Fort St Vrain Nuclear Generating Station ML20077C7171994-11-30030 November 1994 Issue 9 to FPOR-14, Fire Protection Operability Requirements ML20078C0641994-10-12012 October 1994 Revised Fire Protection Operability Requirements,Including Issue 21 to Depp Table of Contents,Issue 2 to FPOR-22 & Issue 3 to FPOR-23 ML20081J7951994-09-15015 September 1994 Issue 5 to DPP 5.4.2, Odcm ML20063M1551994-02-17017 February 1994 Rev 0 to Fsv Nuclear Station Decommissioning Project Final Survey Plan for Site Release ML20057A6151993-08-30030 August 1993 Issue 2 to FPOR-23, Fire Water Makeup Sys ML20079M4261991-10-11011 October 1991 Revised Abnormal Operating Procedures,Reflecting Deletion of Issue 9 of EP Class ML20082H8851991-08-16016 August 1991 Issue 2 to Abnormal Operating Procedure AOP-I-2, Chemical, Petroleum & Hazardous Waste Spill Response ML20091C4141991-08-0202 August 1991 Issue 58 to Abnormal Operating Procedure AOP-L, Loss of Instrument Air Header ML20024H3341991-05-10010 May 1991 Nonproprietary Rev 2 to FSV-P-SCP-100, Fort St Vrain Initial Radiological Site Characterization Program Program Description ML20072V5291991-04-12012 April 1991 Revised Defueling Emergency Response Plan,Including Section 1 Definitions,Section 2 Scope & Applicability,Section 3 Summary of Fsv Derp,Section 4 Emergency Classifications & Section 5 Emergency Organization ML20070V6871991-03-20020 March 1991 Issue 55 to Abnormal Operating Procedure AOP-R, Loss of Access to Control Room ML20072S0521991-03-15015 March 1991 Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 9 to CMIP-11, Classification of Emergency for McGuire Nuclear Station & Rev 11 to CMIP-12, Classification of Emergency for Oconee Nuclear Station ML20066J1171991-02-15015 February 1991 Issue 56 to Intro Section of Abnormal Operating Procedure (Aop),Issue 58 of AOP-A,Issue 58 to AOP-B,Issue 56 of AOP D-1 & Issue 2 of RERP-TRANSPORTATION ML20058N1071990-08-10010 August 1990 Issue 56 to AOP-I, Discussion of Fire ML20154M4361988-09-0909 September 1988 Rev a to EE-FP-0005, Evaluation of Cable Trays Outside of Congested Cable Area ML20154L8751988-05-26026 May 1988 Simulation Facility Program Plan ML20154L8911988-03-31031 March 1988 Rev 2 to Guidance for Development of Simulation Facility to Meet Requirements of 10CFR55.45 ML20148Q4681988-03-31031 March 1988 Rev 2 to Guidance for Development of Simulation Facility to Meet Requirements of 10CFR55.45 ML20150E3141988-03-18018 March 1988 Issue 1 to Procedure SR 5.4.1.3.8.abcd-R1, Steam Line Rupture Detection/Isolation Sys (Slrdis) Calibr & Testing for Panel I-93543 ML20150E2901987-12-31031 December 1987 Issue 5 to Procedure SR 5.3.4b1-A Loop I Shutdown Cooling Power Operated Valve Test ML20238E3541987-08-0707 August 1987 Rev a to EE-ISI-002, Condensate Line Erosion Insp Program ML20236N8681987-08-0505 August 1987 Exercise Manual Fort St Vrain Nuclear Generating Station Nrc/Fema Graded Exercise 870805 ML20235Y7861987-07-29029 July 1987 Issue 2 to MAP-4, Maint Dept Personnel Experience, Qualification & Training Requirements ML20237J6921987-06-0909 June 1987 Issue 6 to MQCIM-1, Maint QC Insp Program. Annotated Page to Rev 4 to Updated FSAR Encl ML20235Y7731987-05-25025 May 1987 Training Mgt Procedure Index ML20237J7151987-05-0101 May 1987 Issue 12 to Q-18, QA Monitoring & Audit Program ML20215H9731987-04-30030 April 1987 Revised Buckle Users Manual:Creep Collapse of Thin-Walled Circular Cylindrical Shells Subj to Radial Pressure & Thermal Gradients ML20215H6101987-04-10010 April 1987 Program Plan for Integrated Validation of NUREG-0737 Initiatives ML20215H6031987-04-10010 April 1987 SOAP-2,Issue 1 of Guidelines for Preparation of Emergency Procedures ML20215H5951987-04-10010 April 1987 Procedures Generation Package:Public Svc Co of Colorado Fort St Vrain Nuclear Generating Station ML20150E3041987-03-20020 March 1987 Issue 1 to Procedure SR 5.4.1.3.8.abcd-R2, Steam Line Rupture Detection/Isolation Sys (Slrdis) Calibr & Testing for Panel I-93544 ML20237J7231987-02-20020 February 1987 Issue 5 to QAAP-1, Planning & Scheduling Fort St Vrain QA Audits ML20212B3081987-02-15015 February 1987 Issue 7 to Procedure G-9, Controlled Work Procedures ML20210E3001987-02-0303 February 1987 Revised Administration Procedures Consisting of Issue 7 to G-3, Action Request-Preparation & Processing, Issue 21 to G-2, Fort St Vrain Procedure Sys & Issue 6 to G-6, Compliance w/10CFR21 Requirements ML20210A7571987-01-30030 January 1987 Fort St Vrain 1987 Power Ascension Plan ML20237J7421987-01-19019 January 1987 QA Audit QAA-502-87-01, Audit Plan - QA Program for Transport of Radioactive Matls ML20212B2851986-10-24024 October 1986 Issue 10 to Procedure Q-3, Design Control Sys ML20238E3461986-10-0202 October 1986 Rev C to EE-ISI-0001, Extraction Steam Lines & Turbine Vent & Drain Lines Erosion Insp Program 1997-06-17
[Table view] |
Text
. - ..
, Encl @sure (2) t to P-82430 PUBLIC SERVICE COMPANY OF COLURADO FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PROGRAM ADDITIONAL SURVEILLANCE REQUIREMENTS
'FOR THE STEAM GENERATOR Supplement to Report EE-22-0002 Rev.1 (Enclosure 4 to PSC Letter P-80064)
September 28, 1982 f'8210060286 820930 i PDR ADOCK 05000267
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. EE-22-0002 R:v. 1 Supplement Page 2 of 7 EXAMINATION OF STEAM GENERATOR TUBE BIMETALLIC WELDS
- 1. INTRODUCTION The Fort St. Vrain steam generator design includes a bimetallic weld in each crossover tube in the transition region between the EESHI bundle and the SH2 bundle. These bimetallic welds, which are subject to elevated temperatures, are not accessible for examination.
Inservice inspection and testing of the Fort St. Vrain steam generator was originally reviewed by PSC in letter P-80064. In their report Q-13:82:5, LANL/ASTA requested that data be supplied which describe material tests performed to date to assure the long term thermal perfonnance of the crossover tube bimetallic welds. These data and further references were provided in PSC letter P-82061. At a meeting between NRC, LANL, ASTA, and PSC held at Fort St. Vrain on July 29, 1982, NRC requested that PSC investigate the possibility of examining other bimetallic welds located in the steam header external to the PCRV steam generator penetration which would be representative of the tube bimetallic welds. This report includes the results of PSC's investigation.
- 2. CROSS 0VER TUBE BIMETALLIC WELDS The crossover tube bimetallic welds are between Incoloy and 2-1/4Cr-1Mo materials. Tube dimensions are approximately 1 in. 0D and 0.20 in, wall thickness. They operate with steam at SH1 outlet conditions in the inside, and primary helium on the outside. The primary helium is for all practical purposes stagnant in the weld region thus limiting heat trans fer, so that there is only a relatively samll temperature gradient from the outside surface to the inside surface.
- 3. EXTERNAL INCOLOY/2-1/4CR-1M0 BIMETALLIC WELUS The same material combination as the steam generator crossover tube bimetallic welds is found in various accessible steam generator bimetallic welds located below the PCRV penetration secondary closure. These welds are illustrated in the attached Figure 1, and are located at the connection of:
a) the main steam subheader thennal sleeves wi th the secondary closure (weld No. 5 in Figure 1),
b) the main steam ring header to the main steam piping (weld No. 2 in Figure 1) and, c) the main steam ring header collector to the header drain line (weld No. 3 in Figure 1).
~
, EE-22-0002 R:;v.1
~
/ Supplement Page 3 of 7 The fi rst weld listed above is not representative of the crossover tube bimetallic welds due to its very di fferent operating conditions as a part of the secondary closure boundary rather than steam piping.
The second and third welds are subject to main steam outlet conditions on the inside, and their outside surface is enclosed in themal insulation. However, the drain line weld is geometrically much more similar to the crossover tube weld than the large, thick wall, main steam pipe weld. Therefore, the main staam ring header collector drain bimetallic weld would be the first choice for monitoring examination. Its temperature is higher than the crossover tube bimetallic weld temperature. Due to its protection by thermal insulation, the heat transfer is al so small resulting in a low temperature gradient from inside to outside. A comparison of weld characteristics is included in Table 1.
The differences between the crossover tube bimetallic welds and the main steam ring header collector drain bimetallic welds appear to be small enough so that the latter welds can be considered representative of the former welds with respect to such phenomena as carbon migration at the fusion line and differential thermal expansion of dissimilar metals.
It should be noted, however, that due to the very different configurations of the steam generator tube bundle, and of the main steam ring header and drain line, operating stresses in the two welds may be quite different. Therefore, should indications be found by examination of the drain weld, the nature and significance of these findings for the crossover tube bimetallic welds will have to be thoroughly investigated before it can be concluded that a similar problem exists in those welds.
The potential effects of di f ferential thermal expansion on bimetallic weld structural integrity would appear to be more accentuated for a large diameter than for a small diameter bimetallic weld, i.e. at the main steam ring header collector to pipe junction than at the collector to drain line junction. Operating temperatures of these two welds are identical, so that potential temperature / time dependent phenomena (such as carbon migration) would be expected to affect both welds in the same fashion. Therefore, if examination of steam generator bimetallic welds is to be performed to allt.viate potential concerns about their long term behaviour, then the large collector to pipe weld should also be examined.
- 4. RECOMt1 ENDED EXAMINATION Examination of bimetallic welds is perfomed to monitor the behavior of these welds over time. Not all welds need to be
- EE-22-0002 Rev. 1 Supplement Page 4 of 7 examined, and it is recommended that two main steam ring header collectors be selected for repetitive volumetric examination of the collector to main steam pipe weld and of the collector to drain line weld at five calendar year intervals. It is also recommended that two additional collectors in each loop (8 welds) be volumetrically examined at the first examination, to establish a base line which could be used, should indications be found in the course of the examination progratn and additional examinations subsequently be required.
OD . ID Weld Max Weld Temperature (inch) (inch) Spec Code Allowable Operating -
Crossover Tube 1.0 0.6 WS-S-205 1015 F V50 P-950 F ,
Main steam ring 1.5 0.75 WS-S-173 1025 F 1008 F header collector /
drain
'l Main steam ring 8.7 5.9 WS-S-173 1025 P 1008 F l
header collector /
jj P i Pe 4
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i TABLE 1 - STEAM GENERATOR BIMETALLIC WELD CIIARACTERISTICS 1
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, EE-22-0002 Rev. '
1 Supplement Page 7 of 7 ATTACHMENT 1 Specification 5.3.11 - Steam Generator Bimetallic Welds, Surveillance The accessible portions of steam generator bimetallic welds shall be volumetrically examined for indications of subsurface defects as follows:
a) The main steam ring header collector to main steam piping weld for one steam generator module in each loop at five (5) calendar year intervals, b) The main steam ring header collector to collector drain piping weld for one steam generator module in each loop at five (5) calendar year intervals.
c) The same two steam generator modules initially selected shall be re-examined at each interval.
d) The bimetallic welds described in (a) and (b) shall also be inspected for two other steam generator modules in each loop during the initial examination.
Basis for Specification 5.3.11 The steam generator crossover tube bimetallic welds between Incoloy 800 and 2-1/4Cr-1Mo materials are not accessible for 4
examination. The bimetallic welds between the steam generator ring header collector, the main steam piping and the collector drain piping are accessible, involve the same materials and operate at conditions not significantly different fran the crossover tube l bimetallic welds. The collector drain piping weld is also geometrically similar to the crossover tube weld. Examination of selected bimetallic welds that are accessible will provide additional assurance Concerning the continued integri ty of steam generator bimetallic welds. Although no degradation is expected to occur, this specification allows for detection of defects which might result from conditions that can uniquely affect bimetallic welds made between these materials. Additional collector welds are inspected at the first examination to establish a baseline which could be used, should defects be found in later inspections and additional examinations subsequently be required.
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