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Category:REFERENCE SAFETY ANALYSIS REPORT & AMENDMENTS (RSAR)
MONTHYEARML20042E9171990-04-30030 April 1990 Amend G to C-E SAR Design Certification ML20011D5171989-12-15015 December 1989 Amend F to C-E Std SAR - Design Certification (CESSAR-DC). ML20248C3801989-03-30030 March 1989 App 3.11A, Environ Qualification for Structures & Components, to CESSAR Sys 80+ Std Design ML20247G6661989-03-30030 March 1989 Chapter 1, Introduction & General Plant Description, to CESSAR Sys 80+ Std Design.W/One Oversize Encl ML20247G6781989-03-30030 March 1989 Chapter 2, Site Envelope Characteristics, to CESSAR Sys 80+ Std Design ML20247G7141989-03-30030 March 1989 Chapter 3, Design of Structures,Components,Equipment & Sys, to CESSAR Sys 80+ Std Design ML20247G8511989-03-30030 March 1989 Chapter 4, Reactor, to CESSAR Sys 80+ Std Design ML20247G8621989-03-30030 March 1989 App 4A Sys 80 Reactor Flow Model Test Program, to CESSAR Sys 80+ Std Design ML20247G8681989-03-30030 March 1989 App 4B, Hot Loop Flow Testing of Sys 80 Fuel & Control Element Assembly Components, to CESSAR 80+ Std Design ML20247G9131989-03-30030 March 1989 Chapter 5, RCS & Connected Sys, to CESSAR Sys 80+ Std Design.W/Two Oversize Encls ML20247G9281989-03-30030 March 1989 App 5A, Overpressure Protection for C-E Sys 80 Pwrs, to CESSAR Sys 80+ Std Design ML20247G9831989-03-30030 March 1989 App 5B, Structural Evaluation of Steam Line Break for Steam Generator Internals, to CESSAR Sys 80+ Std Design ML20247G9891989-03-30030 March 1989 App 5C, Structural Evaluation of Feedwater Line Break for Steam Generator Internals, to CESSAR Sys 80+ Std Design ML20247H0331989-03-30030 March 1989 Chapter 6, Esfs, to CESSAR Sys 80+ Std Design.W/One Oversize Encl ML20247H2911989-03-30030 March 1989 Chapter 7, Instrumentation & Controls, to CESSAR Sys 80+ Std Design ML20247H3081989-03-30030 March 1989 Chapter 8, Electric Power, to CESSAR Sys 80+ Std Design ML20247H3361989-03-30030 March 1989 Chapter 10, Steam & Power Conversion Sys, to CESSAR Sys 80+ Std Design.W/One Oversize Encl ML20248C3921989-03-30030 March 1989 App 3.11B, Identification & Location of Mechanical & Electrical Safety-Related Sys Components, to CESSAR Sys 80+ Std Design ML20248B0021989-03-30030 March 1989 App 3A, Discussion of Finite Difference Analysis for Analysis of Pipe Whip, to CESSAR Sys 80+ Std Design ML20247J0591989-03-30030 March 1989 Chapter 18, Human Factors Engineering, to CESSAR Sys 80+ Std Design ML20247H4781989-03-30030 March 1989 Chapter 17, QA Program, to CESSAR Sys 80+ Std Design ML20247H4701989-03-30030 March 1989 Chapter 16, Tech Specs, to CESSAR Sys 80+ Std Design ML20247H4531989-03-30030 March 1989 App 15C, Analysis Methods for Steam Line Breaks, to CESSAR Sys 80+ Std Design ML20247H4431989-03-30030 March 1989 App 15B, Methods for Analysis of Loss of Feedwater Inventory Events, to CESSAR Sys 80+ Sys Design ML20247H4271989-03-30030 March 1989 App 15A, Loss of Primary Coolant Flow Methodology Description, to CESSAR Sys 80+ Std Design ML20247H4171989-03-30030 March 1989 Chapter 15, Accident Analyses, to CESSAR Sys 80+ Std Design ML20247H4031989-03-30030 March 1989 Chapter 14, Initial Test Program, to CESSAR Sys 80+ Std Design ML20247H3971989-03-30030 March 1989 Chapter 13, Conduct of Operators, to CESSAR Sys 80+ Std Design ML20247H3941989-03-30030 March 1989 Chapter 12, Radiation Protection, to CESSAR Sys 80+ Std Desing ML20247H3601989-03-30030 March 1989 App 11A, Core Residence Times, to CESSAR Sys 80+ Std Design ML20247H3531989-03-30030 March 1989 Chapter 11, Radwaste Mgt, to CESSAR Sys 80+ Std Design ML20247H3161989-03-30030 March 1989 Chapter 9, Auxiliary Sys, to CESSAR Sys 80+ Std Design. W/Four Oversize Encls ML20150B8181988-06-30030 June 1988 Amend C to CESSAR-DC ML20151H8191988-04-11011 April 1988 CESSAR-DC Submittal Group B - Revs to Chapters 1,4,5 & 9 ML20236W0531987-09-11011 September 1987 Amend 12 to CESSAR-F ML20054C1001982-03-31031 March 1982 Amend 7 to CESSAR-F ML20039D4701981-11-20020 November 1981 Amend 6 to CESSAR-F ML19350E0871981-06-15015 June 1981 Amend 3 to Chapter 4 of C-E CESSAR-F, Reactor. ML20003C2771981-02-20020 February 1981 Amend 1 to CESSAR-F 1990-04-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20210R2131997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the System 80+ Design.Docket No. 52-002.(Asea Brown Boveri-Combustion Engineering) ML20148A8211997-01-31031 January 1997 Abb System 80+ Design Control Document - Volume 1 ML20078Q4311995-02-28028 February 1995 Revs to Sys 80+ Design Control Document ML20080E2051995-01-0505 January 1995 Rev 2 to Technical Support Document for Amend to 10CFR51 Considering Severe Accidents Under NEPA for Plants of Sys 80+ Design ML20076F3301994-10-0707 October 1994 Rev 1 of Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of Sys 80+ Design ML20069N4091994-06-20020 June 1994 System 80+ Certified Design Material ML20029E1991994-05-0606 May 1994 Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of Sys 80+ Design. ML20062L8631993-12-31031 December 1993 Certified Design Matl ML20058E9971993-11-15015 November 1993 Amend T to C-E Std SAR - Design Certification. W/31 Oversize Figures ML20062J3551993-10-29029 October 1993 C-E Ssar - Design Certification,Amend S. W/15 Oversize Encls ML20058N1001993-09-23023 September 1993 Rev 2 to Design Alternatives for Sys 80+ Nuclear Power Plant ML20046B3621993-07-15015 July 1993 Session 3.7;Review of SG Tube Rupture Procedure at Borssele Nuclear Power Station. ML20045D3421993-06-30030 June 1993 Rev 1 to Design Alternatives for Sys 80+ Nuclear Power Plant. ML20045C4901993-06-30030 June 1993 Evaluation of Sys 80+ Std Design to Interfacing Sys LOCA Challenges, Special Rept ML20045C6351993-06-15015 June 1993 Amend P to Sys 80+ Cessar - Design Certification. ML20044G7061993-05-19019 May 1993 Common Mode Failure Evaluation for Limiting Fault Events. ML20059F4751993-03-29029 March 1993 Application of Advanced Full Scope Simulators to Severe Accident Precursor Training, Presented at 1993 Simulation Multiconference on 930329-0401 in Washington,Dc ML20044C0061993-03-17017 March 1993 ABB-C-E Sys 80+ Structural Analysis of Nuclear Island & Nuclear Annex Structures. ML20044B9411993-02-23023 February 1993 Rev 0 to Human Factors Evaluation & Allocation of Sys 80+ Functions. ML20012G3351993-02-16016 February 1993 Sys 80+ Advanced LWR PRA-Based Seismic Margin Evaluation. ML20127M4171993-01-15015 January 1993 Control Complex Info Sys Bases for Nuplex 80+ ML20126F6771992-12-0808 December 1992 Review of CE 80+ FMEA & D&Did Analysis, Technical Ltr Rept ML20126J0521992-10-30030 October 1992 Amend K to C-E Std SAR - Design Certification ML20126E6191992-10-30030 October 1992 Amend K to CESSAR - Design Certification. W/50 Oversize Encls ML20012G3361992-09-30030 September 1992 Draft Basis for Seismic Provisions of UCRL-15910. ML20114F5611992-09-30030 September 1992 Evaluation of Defense-in-Depth & Diversity in ABB-C-E Nuplex 80+ Advanced Control Complex for Sys 80+ Std Design ML20104A8501992-09-0202 September 1992 Sys 80+ Design Certification PRA Flood Protection Assessment ML20104A8551992-09-0202 September 1992 Sys 80+ Design Certification Fire Hazard Risk Assessment LD-92-095, Draft Sys 80+ Design Certification Piping Analysis Specification1992-08-31031 August 1992 Draft Sys 80+ Design Certification Piping Analysis Specification ML20101S7281992-07-31031 July 1992 Criteria for Design of Main Control Room & Other Operating Stations for Sys 80+ ML20099D8031992-07-31031 July 1992 Final Sys 80+ Shutdown Risk Evaluation Rept ML20099D8301992-07-25025 July 1992 Draft Sys 80+ Design Certification Distribution Sys Design Guide ML20101G5261992-06-15015 June 1992 Draft Sys 80+ Shutdown Risk Evaluation Rept ML20097A1251992-05-31031 May 1992 Human Factors Engineering Stds,Guidelines & Bases for Nuplex 80+ ML20096F7271992-05-0808 May 1992 Nuplex 80+ Human Factors Design Process Summary ML20095L5041992-04-30030 April 1992 Draft Sys 80+ Shutdown Risk Evaluation Rept, Part 1 ML20095J2321992-04-30030 April 1992 Rev 0 to Design Alternatives for Sys 80+ Nuclear Power Plant ML20095A9601992-04-15015 April 1992 Selection of Control Motion for ABB-C-E Sys 80+ Std Design ML20090K4851992-03-12012 March 1992 QA Program Topical Rept LD-92-031, Marked-up Copy of QA Program Topical Rept1992-03-12012 March 1992 Marked-up Copy of QA Program Topical Rept ML20094E9431992-02-11011 February 1992 LOCA Aspects of C-E Advanced LWR - Sys 80+ ML20079C9641991-06-13013 June 1991 Description of Nuclear QA Program ML20043H2201990-05-18018 May 1990 Seminar on Use of Unusual Event Repts for Improving Nuclear Power Plant Safety. ML20042E9171990-04-30030 April 1990 Amend G to C-E SAR Design Certification ML20106D7341990-01-15015 January 1990 Nuplex 80+ Advanced Control Complex Design Bases ML20096F4761989-12-31031 December 1989 Nuplex 80+ Rev 2 to Verification Analysis Rept ML20011D5171989-12-15015 December 1989 Amend F to C-E Std SAR - Design Certification (CESSAR-DC). ML19324B6831989-10-31031 October 1989 QA Program:Description of Nuclear Power Businesses QA Program, Rev 5 NUREG-1044, Sser Supporting Vendor Responses to Confirmatory Item 1, Shutdown Cooling Sys1989-08-0404 August 1989 Sser Supporting Vendor Responses to Confirmatory Item 1, Shutdown Cooling Sys NUREG-0852, Sser Supporting Vendor Responses to Confirmatory Issue 2, Steam Generator Tube Rupture. Calculated Radiological Consequences of Postulated Steam Generator Tube Rupture Accident Meets 10CFR100.11 Dose Ref Values1989-08-0404 August 1989 Sser Supporting Vendor Responses to Confirmatory Issue 2, Steam Generator Tube Rupture. Calculated Radiological Consequences of Postulated Steam Generator Tube Rupture Accident Meets 10CFR100.11 Dose Ref Values 1997-05-31
[Table view] |
Text
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CESSAR naincy,,,
O APPENDIX 4B HOT LOOP FLOW TESTING OF SYSTEM 80 FUEL AND CEA COMPONENTS O
O 8904040335 890330 PDR ADOCK 05000470 PDC K
CESSAR EininCATION i
( . EFFECTIVE PAGS LISTING CHAPTER _4_
APPENDIX 4B Table of Contents Pace Amendment i
11 l
Text i
Pace
.4B-1 4B-2 4B-3 Ficures i l
4B-1 4B-2 4B-3 4B-4 1
Amendment B March 31,_1988
ls! !h!h k!I !Eb'$ ICATIDN l
1 TABLE OF CONTENTS l
CEAPTER 4 APPENDIX 4B Section Subiect Pace No.
l 1.0
SUMMARY
4B-1 1 2.0 TEST FACILITY DESCRIPTION 4B-1 1 3.0 TEST COMPONEHIS 4B-1
'3.1 STACKUP 4B-1 i 1
3.2 FUEL ASSEMBLIES 4B-1 3.2.1 DETAILS OF DESIGN 4B-1 3.2.2 SPACER GRID SPRING SETTINGS 4B-1 1
() 3.2.3 FUEL ARRAY IN TEST 4B-2
! 3.3 CONTROL ELEMENT ASSEMBLY 4B-2 3.4 SUPPORT STRUCTURES 4B-2 l 4.0 TEST RESULTS 4B-2 4.1 SCRAM TIME vs ACCEPTANCE CURVE 4B-2 4.2 PRESSURE DROPS IN FUEL 4B-2 4.3 FUEL ROD FRETTING 4B-3 4.4 GUIDE TUBE WEAR 4B-3 i
l l
i i
CESSAR EnHricari:n !
9 LIST OF FIGURES l l
Fiqura subiect )
4B-1 C-E TF-2 Hot Loop Piping Isometric j l
4D-2 System 80 Fuel Hot Loop Tests, Component Stackup j in TF-2 4B-3 Cross Section Through Fuel Shroud and Fuel Array TF-2 Tests 4B-4 Test Scram vs Acceptance Curve, 525*F Four Pump Flow Sett ing in TF-2 0
I 1
11
i i
CESSAR Mncarien I l
i O i1 1.0
SUMMARY
l Hot loop flow testing for System 80 was completed during 1980 and 1981. Test components included an array of five fuel assemblies, a- twelve rod control element assembly (CEA) and supporting 1 structures. Pressure drops and CEA scram times were measured for j a range of temperature and flow rate settings. Results were nearly as predicted based on analyses and the results of hot. loop ,
tests preceding those for System 80. A 1300 hour0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> wear test was j run, finding no fretting wear on fuel rods and no other wear on i fuel exterior surfaces. Some wear occurred inside rodded CEA )
guide tubes, but at acceptable rates.
2.0 TEST FACILITY DESCRIPTION Tests were performed in the 36" ID main section of the C-E TF-2 hot loop. The piping arrangement in this loop is shown in Figure 4B-1. The loop is rated at 15000 gpm, 650*F, 2500 psia. System j 80 wear tests were near loop limits, at 14000 gpm, 620*F, 2250 l psia. The loop includes systems which maintain steady ]
temperature and pressure and nominal reactor water chemistry. '
3.0 TEST CO)(_PONENTS O~ 3.1 STACKUP The test component stackup is shown in Figure 4B-2. The support .
structures were bolted to lugs and a support ring at the bottom )
of the test vessel . The fuel assemblies and CEA were enclosed within lower and upper support structures which were joined by bolts at the fuel alignment plate.
3.2 FUEL ASSEMBLIES 3.2.1 DETAILS OF DESIGN Fuel assemblies closely matched the production reactor fuel design, as described in Section 4.2.2. Differences include:
- a. Fuel rod loading -
46% of the fuel rods (544) were of prototype construction, with depleted UO pellets. 54%
(636) were dummy rods of solid stainless 8 teel. Prototype ,
rods were used in all positions of concern with respect to fretting wear.
- b. Lower end fitting leg braces -
Braces are deleted between legs of the lower end fitting. The test design, with braces, is considered less favorable with respect to O
d fretting wear.
4B-1
CESSARE h mu O
- c. Inside dimension of guide tubes and upper end fitting posts
- The I.D. in upper ends of guide tubes and in UEF posts for production fuel will be enlarged slightly as a precaution, allowing use of wear sleeves if needed. The tube enlargement should have little effect on guide tube wear tendency, and will speed CEA scrams very slightly.
3.2.2 SPACER GRID SPRING SETTINGS i
All spacer grid springs were set for the minimum restraint of i fuel rods expected during the fuel lifetime. ]
3.2.3 FUEL ARRAY IN TEST i
The array of five fuel assemblies is shown in Figure 4B-3, with !
relationships to the fuel shroud and CEA.
1 3.3 CONTROL ELEMENT ASSEMBLY System 80 reactors utilize both 4-rod and 12-rod control element assemblies. The twelve-rod CEA was chosen for hot loop tests because it has a lower weight per rod ratio (hence slower scrams) and is a more complex structure. The test CEA was functionally identical to that shown in Figure 4.2-4.
3.4 BUPPORT STRUCTURES The support structures were prototypical sections of a System 80 eactor and provided support and alignment of the fuel assembliea and CEA. The lower ends of the fuel assemblies engage alignment pins on an open grid beam array. The fuel shroud cross section is noted in Figure 4B-3. All possible corner shapes and fuel to l shroud clearances were included. Shroud tubes in the upper guide structure were held by the upper guide structure support plate (UGSSP) and fuel alignment plate (FAP), and engaged the four posts of each fuel upper end fitting. The region between the FAP and the UGSSP is an outlet plenum, where flow passes up around the shroud tubes and exits the outlet nozzle. Effects of a pressure gradient across the reactor outlet plenum were included in tests. The gradient causes flow circulation through small !
holes in the UGSSP upward near the reactor centerline and downward near the outlet nozzles. In TF-2 the flow circulation was driven through external piping, employing a CEA shroud and seal assembly above the UGSSP.
O' 4B-2
1 CESSAR naama J
.]
O l I I
4.0 TEST RESULTA 1 4.1 SCRAM TIME vs.' ACCEPTANCE CURVE Hot loop tests previewed the scram performance checks that will be required for. every reactor CEA, during pre-operation functional tests. Measured position vs. time for reactor scrams must fall below an acceptance curve applied in safety analyses.
The reactor scram tests are run at approximately 525'F with all pumps running.. The TF-2 scram for- this condition is plotted in Figure-4B-4, along'with the acceptance. curve.
4.2 PRESSURE DROPS IN FUEL
\
Fuel assembly pressure drops were measured over a range of flow I Reynolds Numbers. Results are in good agreement with I measurements for similar fuel in prior TF-2 tests. The test ,
results support the flow resistances used in design analyses for System 80 and based on the prior measurements. -Those analyses include fuel holddown spring design requirements, support structure ' hydraulic loadings and the prediction of system flow rate.
4.3 FUEL ROD FRETTING )
The test for fuel rod fretting ran 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, at 620'F, with !
flow 2.13,400 gpm through the fuel. The setting was based on ]
worst cace reactor conditions at full power, with 116% design l flow. Test fluid velocities exceeded those expected in any System 80 reactor. No fretting was found on any of the fuel rod surfaces.
4.4 GUIDE TUBE WEAR j In view of observed CEA motion dependence on upper guide structure flow, a saquence of different flow conditions was set during the wear test. With each change of flow conditions the ,
CEA was raised to a new position. Conditions included the l maximum pressure differences expected across the reactor UGSSP, '
both upward and downward, and an intermediate equalized condition. Following the tests all CEA guide tubes were '
inspected with an eddy current probe. Four tubes which gave the largest indications were removed and sectioned longitudinally (clamshelled) for precise inspection. Greatest wear occurred for the UGSSP upflow condition, at points of CEA rod tip contact in guide tubes. Guide tube wear at the highest rate observed in-tests will not contribute to violation of stress limits described in Section 4.2.3.
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SUPPORT BARREL AND p FUEL SHROUD ASSEMBLY
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, Figure SYSTEM 80 FUEL HOT LOOP TESTS, jyj / COMPONENT STACKUP IN TF-2 4B-2
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mra{]l6 STANDARD DESIGN ws T9 N-CESSAR !!!'rir",cor,o, l0 Volume 5 comansnon)emmas
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