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Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20210H2901999-07-30030 July 1999 Proposed Tech Specs Pages,Extending Operation of RBS from Current Licensed Power Level of 2894 Mwt to Uprated Power Level 3039 Mwt ML20205P8701999-04-15015 April 1999 Proposed Tech Specs,Implementing Bwrog/Ge Enhanced Option I-A (EI-A) Reactor Stability long-term Solution ML20205L7321999-04-0808 April 1999 Proposed Tech Specs,Removing Notes That Refer to Specific Cycles,Per LAR 1998-15 Re Change to TS 2.1.1.2, Reactor Core Safety Limits ML20206R9351999-01-12012 January 1999 Proposed Tech Specs Re Refueling Equipment Interlocks Required Actions ML20199C1881999-01-0808 January 1999 Revised Page 7 to Rev 13 of RBS Security Training & Qualification Plan ML20198E8091998-12-16016 December 1998 Proposed Tech Specs Pages Re Change to TS Safety Limit 2.1.1.2,changing Two Recirculation Loop MCPR Limit from 1.13 to 1.12 & Single Recirculation Loop MCPR Limit from 1.14 to 1.13 ML20196A1401998-11-20020 November 1998 Proposed Tech Specs Adding 3.10.9, Control Rod Pattern - Cycle 8, to Support Startup from Outage Planned for Early Dec 1998 ML20154H3691998-10-0808 October 1998 Proposed Tech Specs Implementing BWROG Enhanced Option I-A Reactor Stability Long Term Solution as Documented in NEDO-32339,Rev 1, Reactor Stability Long-Term Solution, Enhanced Option I-A ML20153H0311998-09-23023 September 1998 Proposed Tech Specs Changing Specific Gravity Acceptance Criteria for Div III Battery ML20153E8111998-09-22022 September 1998 Proposed Tech Specs Deleting License Conditions Associated with Transamerica Delaval,Inc EDGs ML20217J0951998-04-21021 April 1998 Proposed Tech Specs Bases Update Package for Revs Implemented Since Last Submittal Through Refuel Outage 7 ML20216F1941998-04-0909 April 1998 Proposed Tech Specs License Condition 2.C(13) Re Partial Feedwater Heating & Final Feedwater Temp Reduction RBG-44355, Cycle 8 Startup Physics Test Summary1998-01-19019 January 1998 Cycle 8 Startup Physics Test Summary ML20202F0361997-11-26026 November 1997 Rev 0 to IST Plan for Pumps & Valves Second Ten-Yr Interval ML20216E4861997-09-0505 September 1997 Proposed Tech Specs Re APRM Setpoints,Reactivity Anomalies & RPS Instrumentation ML20210K2131997-08-15015 August 1997 Proposed Administrative Changes to TS 2.1.1.2 W/Note to Indicate Operating Cycle for Which SLMCPR Value Has Been Approved & TS 5.6.5 W/Note Providing Link to Specific GE Documents Used in Approving SLMCPR Changes for Given Cycle ML20217J0481997-08-0505 August 1997 Proposed Tech Specs 2.1.1.2, Reactor Core (Safety Limits), Increasing Two Recirculation Loop MCPR Limit to 1.13 & Single Recirculation Loop MCPR Limit to 1.14 ML20151L6041997-07-31031 July 1997 Proposed Tech Specs,Requesting Approval of Change to TS 3.6.1.8, Penetration Valve Leakage Control Sys, for Upcoming Outage ML20196J7741997-07-30030 July 1997 Proposed Tech Specs Clarifying Surveillance Requirements 3.6.4.2.1 & 3.6.4.2.2 to Match Corresponding Changes to NUREG-1434 ML20134A7021997-01-20020 January 1997 Proposed Tech Specs to License NPF-47,enhancing Operation of Plant to Allow Operation at 100% Power W/Lower Core Flows & Use of Flow Control Spectral Shift Strategies to Increase Cycle Energy ML20133M0451997-01-10010 January 1997 Proposed Tech Specs 3.4.11 Re RCS Pressure & Temp Limits ML20140F0151996-12-26026 December 1996 Rev 6 to RSP-0008, ODCM Procedure ML20134M2271996-11-15015 November 1996 Proposed Tech Specs,Changing TS SL 2.1.1.2, Reactor Core (Sl), Which Increases Two Recirculation Loop MCPR Limit from 1.07 to 1.10 & Single Recirculation Loop MCPR from 1.08 to 1.12 ML20134M0151996-11-15015 November 1996 Proposed Tech Specs 3.8.1 Re AC Sources Operating ML20134M2061996-11-0606 November 1996 Proposed Tech Specs,Permitting Increase in Allowable Leak Rate for MSIVs & Deleting Penetration Valve Leakage Control Sys & Ms Positive Leakage Control Sys Requirements ML20129J4231996-10-24024 October 1996 Proposed Tech Specs 3.8.1 Re A.C. Sources - Operating ML20117K1071996-08-29029 August 1996 Proposed Tech Specs Re Reactor Vessel Matl Surveillance Program Capsule Withdrawal Schedule ML20117G7401996-08-29029 August 1996 Proposed Tech Specs 3.1 Re Reactivity Control Systems ML20116P1341996-08-0808 August 1996 Technical Requirements Manual ML20116J3151996-08-0707 August 1996 Proposed Tech Specs,Submitting TS Bases Pages for Revs Performed Since 951001 Approval of TS Bases ML20116L6321996-08-0101 August 1996 Proposed Tech Specs Re Primary Containment/Drywell Hydrogen Mixing Sys & Drywell Isolation Valves ML20113F2271996-06-12012 June 1996 Rev 0 to Calculation PI-0103-00203.002-101, Ampacity Derating Evaluation ML20117L2581996-06-0404 June 1996 Proposed Tech Specs Re Replacement Page for Ltr RBG-42764, Proposed Tech Specs 3.9.1 Re Refueling Equipment Interlocks1996-05-31031 May 1996 Proposed Tech Specs 3.9.1 Re Refueling Equipment Interlocks ML20112E7831996-05-30030 May 1996 Proposed Tech Spec 3.8.3 Re Incorporation of 5-start Air Pressure Design Criterion for Div III DG ML20117H3271996-05-20020 May 1996 Proposed Tech Specs Re Change in Co Name from Gulf States Utils Co to Entergy Gulf States,Inc B130180, Surveillance Specimen Program Evaluation for River Bend Station1996-04-30030 April 1996 Surveillance Specimen Program Evaluation for River Bend Station ML20107E4681996-04-18018 April 1996 Proposed Tech Specs Re Fuel Loading W/Control Rods Withdrawn or Removed from Defueled Core Cells ML20094N7741995-11-22022 November 1995 Proposed Tech Specs,Clarifying Info Contained in License Amend Request (LAR) 95-21 ML20094M8001995-11-20020 November 1995 Proposed Tech Specs Re Drywell Leak Rate Testing Requirements ML20094N0091995-11-20020 November 1995 Proposed Tech Specs Eliminating Selected Response Time Testing Requirements ML20094B0001995-10-26026 October 1995 Proposed Tech Specs,Consisting of Change Request 95-10, Deleting TS 5.5.12, Biofouling Prevention & Detection, & Various Editorial Corrections ML20093L7361995-10-24024 October 1995 Proposed Tech Specs 3.6.1.1 Through 3.6.1.3 Re Containment Sys ML20108B5941995-10-0101 October 1995 Rev 5A to River Bend Station Offsite Dose Calculation Manual Procedure ML20087J5451995-08-17017 August 1995 Proposed Tech Specs Re Engineered Safety Sys Which Responds to Fuel Handling Accident Conditions ML20084L5001995-05-30030 May 1995 Proposed Tech Specs,Allowing TS 3/4.6.2.2, Drywell Bypass Leakage, Tests to Be Performed at Intervals as Long as Five Yrs Based on Demonstrated Performance of Drywell Structure ML20084L4541995-05-30030 May 1995 Proposed Tech Specs Re Request for Exemption from App J, Primary Reactor Containment Leakage Testing for Water- Cooled Power Reactors ML20078M1171995-02-0808 February 1995 Proposed Tech Specs Page Re Secondary Containment Automatic Isolation Dampers ML20077S4401995-01-18018 January 1995 Revised Pages of Proposed Tech Specs Re Conversion to TS Based on NUREG-1434, Improved TS Rev 0 ML20078H0881994-11-10010 November 1994 Proposed Tech Specs Reflecting Clarifications to LAR 91-11 1999-07-30
[Table view] Category:TEST REPORT
MONTHYEARRBG-44355, Cycle 8 Startup Physics Test Summary1998-01-19019 January 1998 Cycle 8 Startup Physics Test Summary ML20116M8051992-08-31031 August 1992 Reactor Containment Building Integrated Leak Rate Test for River Bend Station ML20043D4611990-05-30030 May 1990 Suppl 16 to Startup Test Rept, for Period Through 900530. W/900601 Ltr ML20042E9891990-04-27027 April 1990 Suppl 15 to Startup Test Rept. W/900430 Ltr ML20006C5581990-01-27027 January 1990 Unit 1 Startup Test Rept (Suppl 14). W/900131 Ltr ML19354D4471989-10-31031 October 1989 Startup Test Rept (Suppl 13), for Period Through 891027. W/891031 Ltr RBG-31323, Unit 1 Startup Test Rept (Suppl 12), Covering Period Through 8907271989-08-0404 August 1989 Unit 1 Startup Test Rept (Suppl 12), Covering Period Through 890727 ML20246M6951989-05-31031 May 1989 Reactor Containment Bldg Integrated Leakage Rate Test Types A,B & C Periodic Test RBG-30602, Unit 1 Startup Test Rept (Suppl 11)1989-04-27027 April 1989 Unit 1 Startup Test Rept (Suppl 11) RBG-29153, Suppl 9 to River Bend Station Unit 1 Startup Test Rept1988-10-13013 October 1988 Suppl 9 to River Bend Station Unit 1 Startup Test Rept RBG-28346, Startup Test Rept (Suppl 8) for Unit 11988-07-18018 July 1988 Startup Test Rept (Suppl 8) for Unit 1 ML20043A9081988-05-25025 May 1988 Determination of Fast Neutron Flux Density & Fluence:River Bend Station. RBG-27778, Suppl 7 to River Bend Station Unit 1 Startup Test Rept1988-04-15015 April 1988 Suppl 7 to River Bend Station Unit 1 Startup Test Rept RBG-27434, Suppl 6 to River Bend Station-Unit 1 Startup Test Rept1988-01-27027 January 1988 Suppl 6 to River Bend Station-Unit 1 Startup Test Rept RBG-26875, Gulf States Utils Co,River Bend Station-Unit 1 Startup Test Rept (Suppl 5)1987-10-27027 October 1987 Gulf States Utils Co,River Bend Station-Unit 1 Startup Test Rept (Suppl 5) RBG-26366, Suppl 4 to Gulf States Utils Co River Bend Station Unit 1 Startup Test Rept1987-07-28028 July 1987 Suppl 4 to Gulf States Utils Co River Bend Station Unit 1 Startup Test Rept ML20210B2831987-04-23023 April 1987 Suppl 3 to Startup Test Rept ML20212C8861987-01-30030 January 1987 Suppl 2 to Startup Test Rept ML20214P2001986-11-18018 November 1986 Suppl 1 to Startup Test Rept RBG-24132, Startup Test Rept,River Bend Unit 11986-07-31031 July 1986 Startup Test Rept,River Bend Unit 1 ML20132B2801985-08-31031 August 1985 Evaluation of CLASIX-3 Conservatisms & Quarter-Scale Tests ML20134H2151985-06-30030 June 1985 Ani/Maerp Std Method of Fire Tests of Cable & Pipe Penetration Fire Stops:Internal & External Seal of Aluminum Conduits,B&B/Promatec CTP-1092,A Issue & CTP-1093,A Issue, Final Rept ML20134G8601985-04-30030 April 1985 Reactor Containment Bldg Integrated Leakage Rate Test, Types A,B & C,Preoperational Test ML20094N7891984-01-13013 January 1984 GE Plant Piping Analysis Ii,Design Memo 150-96 ML19305E7131980-03-20020 March 1980 Rept of Potential Alkali Reactivity ASTM C227. ML19305E6841980-03-11011 March 1980 Rept on Potential Alkali Reactivity ASTM C227 & C-289. ML19322E9601980-02-20020 February 1980 Rept of Potential Alkali:Reactivity,Astm C227. ML19322E9631980-02-20020 February 1980 Rept of Potential Alkali:Reactivity,Astm C227. ML19322E9521980-02-12012 February 1980 Rept of Potential Alkali:Reactivity,Astm C227 & C289. ML19322E9531980-02-12012 February 1980 Rept of Potential Alkali:Reactivity,Astm C227 & C289. ML19322E9471980-01-23023 January 1980 Rept of Potential Alkali:Reactivity,Astm C227. ML19322E9461980-01-23023 January 1980 Rept of Potential Alkali:Reactivity,Astm C227. ML19322E9341980-01-15015 January 1980 Rept of Potential Alkali:Reactivity,Astm C227 & C-289. ML19322E9381980-01-15015 January 1980 Rept of Potential Alkai:Reactivity,Astm C227 & C-289. ML19294B2821979-12-20020 December 1979 Rept of Potential Alkali Reactivity ASTM C-227 & C-289 for Gulf States Utils Co - S&W Agent, for Fine Aggregate ML19294B2871979-12-20020 December 1979 Rept of Potential Alkali Reactivity ASTM C-227 & C-289 for Gulf States Utils Co - S&W,Agent for Number 57 Coarse Aggregate 1998-01-19
[Table view] |
Text
Ent:rgy Cper:tiJns,Inc.
River Bend Station e
5485 U. S H.ghway 61 PO. Box 220 l 2'25 Fax 504 635 5008 Rick J. King 4 c ar SWety & n.,sulatory Af' airs Jar.uary 19,1998 U.S. Nuclear Regulatory Commission Document Control Desk, OPI-17 Washington,DC 20555
Subject:
River Bend Station - Unit 1 Docket No. 50-458 Cycle 8 Startup Report File No.: G9.5, G9.25.1.4 RBI 1-98-0026 l* RBG-44355 Ladies and Gentlemen:
In accordance with River Bend Station (RBS) Techmcal Requirements Manual TR 5.6.8, enclosed is a Startup Report that piovides a summary of the startup physics testing conducted on the Cycle 8 core reload, if you have any questions, please contact Mr. T. W. Gates at (504) 381-4866.
Sincerely, h
RJK/rWG/bni ,
enclosure t
.}(p g 9601290174 900119 PDR ADOCK 05000458 p PDR iil Il l
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cc: NRC Resident Inspector P. O. Box 1050
. St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington,TX 76011 LA Department of Environmental Quality
' Radiation Protection Division P. O. Box 82135 Baton Rouge, LA 70884-2135 Attn: Administrator Mr. David L. Wigginton NRR Project Manager U.S. Nuclear Regulatory Commission M/S OWFN 13-H-13 Washington, DC 20555
e ENCLOSURE River Bend Station Unit 1 Cycle 8 %
St:stup Physics Test Summary 1
Cnclost.te Cycle 8 Startup Repon -
- Page_1of5-OVERVIEW-River Bend Station (RBS) resumed commercial operation in Cycle 8 on October 19,1997 q following a Refueling / Maintenance Outage, The Cycle 8 reload consisted of replacing .
- 212 GE8B General Electric fuel assemblies with 212 GEiI fuel assemblies The
- following startup tests were performed during Refueling Outage RF07 and while -
attaining full power after RF07 and are summarized in this repon:
.1) Core Loading Verification
- 2) Control Rod Functional Testing
- 3) Shutdown Margin Determination- '
t l' 4) TIP Asymmetry In rddition to the above startup physics tests, the startup test program included:- Core Monitoring System Verification and ' Recirculation System Calibration as well as other L serveillance testing required by RBS Technical Specifications. The additional test results - ;
- are available at the site on request.
\
h
-~
Enclosure
_ Cycle 8 Startup r. port
_p _ Page 2 of 5 CORE LOAll.NG VERIFICATION Purpose -
Ensure each reactor fuel assembly is: R
- - in its correct core location
-* oriented properly -
Le1 seated properly in its support piece Criteria -
The reactor core is visually checked ?.o verify conformance to the vendor supplied core
'g
- loading pattein. Fuel assembly serial numbers, orientations, and core locations are recorded. A height check is performed to verify all assemblies are properly seated. ,
Results 1
' The as-loaded core was verified for proper fuel assembly serial numbem, locations, orientation and seating in accordance with the RBS Cycle 8 core load:ag pattern. There '
were no location or orientation deviations from the Cycle 8 core loading pattern. .
The core verification procedure was successfully conipleted on October 1,' 1997. ..
9
. _ _. l
J' Enclosure
.,_ Cycle 8 Startup Report
,g ._ Page 3 of 5 CONTROL ROD FUNCTIONAL TESTING
Purpose:
l Verify operability of each control rod by:
. *L performing normal withdrawals and insertions '
. _ ensuring it is latched'to its control rod drive J. assuring that it moves at design speeds without excessive friction t
Criteria L Functional testing of each control rod is performed to ensure proper operability. This testing includes withdrawal and insertion timing, coupling . verification, friction testing .
_ where required, and scram time testing.
G
.Results Friction testing was required for 21 control rods based on the channel management l analysis for cycle 8. This testing was performed per REP-0037,-LPRM Operatality,"
prior to startup.
Each individual control rod was timed during a normal withdrawal and insertion -
sequence. Control rods with stroke times outside the tolerance of normal stroke time +/-
20% were readjusted to within the normal strokt time +/- 10%. This was in accordance with GE recommendations.
( Each control rod was scram time tested during reactor startup in accordance with RBS
' Technical Specification Surveillance Requirement 3.1.4.1. All of the control rod scram times were within the allowable limits.
'A control rod coupling check was performed in accordance with RBS Technical
. Specification Surveillance Requirement 3.1.3.5 each time a control rod was fully 1 withdrawn.
- The fhnge 0-rings of a Control Rod Drive Mechanism were replaced during the outage.
. This one control rod was SCRAM'd at atmospheric pressure prior to startup and scram time tested during reactor start-up in accordance with Technical Specification Surveillanee Requirements 3.1.4.3 and 3.1.4.4 with satisfactory results.
~
Enclosure L
Cycle 8_Startup Report
- Page 4 of 5 l
SHUTDOWN MARGIN DETERMINATION Purpose
- To ensure that:
- - - the reactor can be made suberitical from all operating conditions ei the reactivity transients associated with postulated accident conditions are controllable wMin acceptable limits e the reactor will be maintained sufficiently suberitical to preclude inadvertent critically in the shutdown condition
- Critenia The suberitical demonstration verifies the reactor remains suberitical with the analytically determined strongest worth control rod fully withdrawn.
l The in-sequence roc Nwal shutdown margin calculation ' . gins by withdrawing control rods in their stanumd sequence until criticality is achi <ed. The shutdown margin
- of the core is determined from calculations based on the critical rod pattem, the reactor period, and the moderator temperature.
Reactivity Anomaly verification is performed in accordance with Technical Specification Surveillt.nce Requirement 3.1.2.1 after reaching equilibrium xenon concentrations at greater than 80% reactor power.-
Results The suberitical demonstration was performed on October 2,1997. !
1 The in-sequence critical shutdown margin surveillance procedure was completed on
. October 19,' 1997. The shutdown margin (SDM) at the beginning-of-cycle (BOC) was calculated to be 1.34 % delta k/k. The Cycle 8 "R" value is equal to 0.0% delta k/k.
- Therefore, the Cycle 8 minimum shutdown margin at the most reactive point in the fuel cycle is 1.34 % delta k/k which is well within the RBS Technical Specification 3.1.1 requirement of 0.38% delta k/k.
It was verified on October 26,1997 that no reactivity anamoly was present by performance ofTechnical Specification Surveillance Requirement 3.1.2.1.
l^
i
.? iEnclosure -
i
- 7 Cycle 8 Startup Report -
Page5of5.
TIP ASYMMETRY CHECK Purpose -
To determine the reproducibility of tlie Traversing Incore Probe (TIP) system readings .
Criteria =
An asymmetry determination is performed as part of a detailed statistical uncertainty -
- . evaluat.'on of the TIP System. A complete set of TIP data is obtained at steady state conditions while greater than 50% rated core thermal power. A total TIP statistical' evaluation is then performed by assessment of the random noise and geometric
. uncertainties. The results are evaluated to assure proper operation of the TIP System.
Resu'.ts The TIP reproducibility and symmetry uncertainty calculations were performed on l
- October 29,1997 at 100% core thermal power The total TIP uncertainty was evaluated to
.l be equal to 3.2336 % which is less than the limit value of 6%.
-