ML20070H983
| ML20070H983 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 12/22/1982 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82-157, NUDOCS 8212270302 | |
| Download: ML20070H983 (33) | |
Text
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:157 DEC 22 1932 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Check:
ADDITIONAL INFORMATION FROM NOVEMBER 22-24, 1982, MECHANICAL ENGINEERING BRANCH (MEB)/ CLINCH RIVER BREEDER REACTOR PLANT (CRBRP) MEETING
Reference:
HQ:S:82:143, J. R.-Longenecker to P. S. Check,
Subject:
Meeting Summary, November 22-24, 1982, MEB/CRBRP Meeting, dated December 14, 1982 Enclosed are a.new Preliminary Safety Analysis Report (PSAR) Table 3.2-4 and an amended PSAR Section 4.2 that respond to MEB items 3 and 41 from the refererice letter. This information will be incorporated in Amendment 75 to the PSAR scheduled for January 1983 Questions regarding this submittal may. be directed to D. Robinson (FTS 626-6098) of the Oak Ridge Project Office staff.
Sincerely, n
HR ttfu John R. Longengker Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy 2 Enclosures cc: Service List Standard Distribution Licensing Distribution gel o
I/P
" 8212270302 825222 PDR ADOCK 05000537 PDR._
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TABLE 3.2-4 PRELIMINARY LIST OF APPLICABLE (X) DES FOR NON-SAFETY CLASS EQiANICAL SYSTEMS / COMPONENTS Applicable Codes Svstems/Comoonents and Standards location Reactor Refuelino Svstem Non-safety Reiated ASE V I I I/1, RCB, RSB Equipment AISC Supports ASE III/NF3 R2,RSS Nuclear Island Maintenance Svstemn Vessels ASME Vill /1 RCB,RSB Supports ASME III/NF3, RCB,RSB AISC PIplng ANSI B31.1 RCB,RSB Valves ANSI B31.1 RCB,RSB Steam Generator Svstem Water Dubp Subsystem ASME lil/3 SGB SWRPRS Piping ANSI B31.1 SGB (SGB walI to fIoor stack)
Normal Chilled Water Svstem Chillers Condensers, ANSI B31.1 SGB Evaperators Piping and Valves ANSI B31.1 Except RT i
Circulating Pumps Manufacturers SGB standards 3.2-12 Amend. 75 Na RS.S
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TABLE 3.2-4 (Cont. )
PRELIMINARY LIST OF APPLICABLE CODES FOR NON-SAFETY Q. ASS KOiANICAL SYSTEMS / COMPONENTS Applicable Codes Systems /Comoonents and Standards Location Radioactive Waste Svstem RSB,RWB Tanks ASE Sec. ViiI/
API 650/620
. Piping & Valves ANSI B31.1 Pumps Manufacturers standard Balance of Plant HVAC ASHRAE PSB, TGB, MSW, CWPH, SMACNA CB, sri, FPPH Main and Auxiitarv Steam ANSI B31.1 TGB, SGB Heat Rejection System ANSI B31.1 TGB, Water Cooling AWWA Tower Yard River Water System ANSI B31.1 River Water Pump House, AWWA Yard Plant Service Water System Normal Plant Water System ANSI B31.1 SGB/IB, CB, RSB/IB, TGB Hot Water Heating System ANSI B31.1 TGB, SGB, RSB, DGB,
- PSB, MSW Treated Water Svstem ANSI B31.1 TGB, RSB, DGB, MSW, PSB
& Manufacturers SGB, kB, CB, GH standard 3.2-13 Amend. 75 Jan. 1983
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TABLE 3.2-4 (Cont. )
PRELiMlNARY LIST OF APPLICABLE CODES FOR NON-SAFETY Q. ASS EOiANICAL SYSTEMS /COff0NENTS AppiIcabie Codes Systems /Ccroonents and Standards Location Non-Sodium Fire Protection Sprinkier and Spray System NFPA-13 & 15 RSB, MSW, PSB Gas Blanket System NFPA-12A G
Portable Fire Protection NFPA-10 SGB, G, DGB, TGB, RSB, System MSW, PSB Dry Chemical Fire NFPA-17 SGE, E DGB, TGB, RSB, Protection M5W. P3D Fire Detection Alarm System NFPA-72A, D & E SGB, G, DGB, TGB, RSB, MSW, PSB Feedwater Water and Condensate System Feedwater* Heater & Deaerator ASME Section V ill, TGB Division 1 Piping ANSI B31.1 TGB Startup drains piping ANSI B31.1 TGB, SGB and equipment-l l
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3.2-13a Amend. 75
.m
UPDATE TO PSAR CHAPTER 4 4-i 4-vi 4-xxi 4-xxx 4.2-17 4.2-127 4.2-151 4.2-152 4.2-153 4.2-175 4.2-176a 4.2-227j 4.2-459/thru-459fe 4.2-528 thru -528a.
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CHAPTER 4.0 - REACTOR TABLE OF CONTENTS Page No.
4.1
SUMMARY
DESCRIPTION 4.1 -1 4.1.1 Lower Internals 4.1 -3 4.1.2 Upper Internals 4.1 -4 4.1.3 Core Restraint
(.1.4 Fuel, Blanket and Removable Radial Shield Regions 4.1-4 4.1. 4.1 Fuel and Axial Blankets 4.1-4 4.1.4.2 Inner and Radial Blanket 4.1-6 4.1.4.3 Removable Radial Shield 4.1-7 4.1.4.4 Control 4.1-7
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4.1.5 Design and Performance Characteristics 4.1-9 4.1.6 Loading Conditions and Analysis Techniques 4.1 9 4.1.[
Computer Codes 4.1 10 4.2 MECHANICAL DESIGN 4.2.1 Fuel and Blanket Design 4.2-1 4.2.1.1 Design Bases 4.2-1 4.2.1.1.1 Functional Requirements 4.2-1 4.2.1.1.2 Operational and Design Requirements 4.2-2 4.2.1.1.2.1 Operating Conditions 4.2-2 4.~2.1.1. 2. 2 Design Requirements 4.2-5 4.2.1.1.2.2.1 Fuel and Blanket Assembly Structural 4.2-15 Component Design Criteria 51 4.2.t. t. 2. 2. 2 L.att ww suca Ass,ai, Sw+nt c.~,.. wi tes: p c a,:~
Amend. 51 Sept.1979 I.,
4-1
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TABLE OF CONTENTS Continued Page No.
4'.2.2.2.1.3 Bypass Flow Module 4.2-166 4.2.2.2.1.4.
Fixed Radial Shield 4.2-167 59l58l 4.2.2.2.1.5 Fuel Transfer and Storage Assembly 4.2-168 4.2.2.2.1.6 Horizontal Baffle 4.2-168 4.2.2.2.1.7 Upper Internal Structure 4.2-170
- 4. 2. 2. 2.1. 8 Core Restraint System 4.2-174 4.2.2.2.1.9 Removable Radial Shield 4.2-175 4.2.2.2.1.10 Core Former Structure 4.2-175 4.2.2.2.1.11 Maintainability 4.2-175 59 4.2.2.2.1.12 Surveillance 4.2-175 4.2.2.3 Design Criteria 4.2-175a
- 4. 2. 2. 3.1 Lower Internals Structure (LIS) 4.2-176
,j
- 4. 2. 2. 3.1.1 Core Support Structure (CSS) 4.2-176 4.2.2.3.1.2 Lower Inlet Module (LIM), Bypass Flow i
Module (BPFM), and Core Former Structure 4.2-176 (CFS) 4.2.2.3.1.3 Horizontal Baffle (HB), Fuel Transfer and Storage Assembly (FT & SA), and 4.2-176a.
- 4. 7. 2. 3. %.4 Fixed Radial Shield (FRS)
Remosate. (?.aA; l SW.ld 43fe-*bl y
- 4. 2. - We c.,
4.2.2.3.2 Upper Internals Structure (015) 4.2-177 4.2.2.3.2.1 Class 1 Appurtenances 4.2-177' 4.2.2.3.2.2 Internal Structure 4.2-177 4.2.2.3.2.3 Modifications to the High Temperature Design 4.2-178 59 Rules for Austenttic Stainless Steel 4.2.2.3.3 AddWonal Matedal No.oerties 51 0 2-181 4.2.2.3.3.1 Inconel 718 Fatique Properties 4.2-181 4.2.2.3.3.2 Environmental Effects on Material Properties 4.2-182
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4-vi hend. 59 Dec. 1980
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t1ST OF TABLES Continued
,P_ age No.
a 4.2-57 Irradiated Weld Material Tensile Properties 4.2-435
'.2-437 4
54l 4.2-58 Deleted 4.2-69 CRBRP Design Basis Transients to be Enveloped 4.2-439 in Fuel Rod Transient Tests 4.2-60 Deleted 4.2-61 Deleted 4.2-62 National Reference Fuel Steady State Program 4.2-441 Activities Summary 4.2-63 Deleted 58l 4.2-64 Design Considerations for Self-llelding and Seizing 4.2-445 of Rotating or Moving Parts 4.2-65 CRBRP Fuel Assembly Development Planning 4.2-448 4.2-66 Preliminary Cladding Fatigue Estimate 4.2-458
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4.2-67 Design para-eters for 6' Pin Tests 4.2-459 React [or uescripuon_4 - p NS E(LT 4
4.3-90 4.3-1 4,3-2 Fuel Isotopic Composition 4.3-96 4.3-3 Start-of-Cycle Excess Reactivity Requirement 4.3-97 4.3-4 Heavy Metal Mass (KG) Inventory in the CRBRP 4.3-98 4.3-5 CRBRP Fuel Management Scheme 4.3-104 4.3-6 Delayed Neutron Constants for CRBRP 4.3-105 4.3-7 Fuel and Inner Blanket Power Fraction Summary 4.3-106
- 4. 3'-8 Axial Blanket, Axial Extension Power Normali-4.3-107 zation Factors l
4.3-9 Radial Blanket Assembly Power Sumary 4.3-108
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51 4.3-10 Radial Blanket Peak Rod Power Sumary 4.3-109 f-4-xxi Amend.-58
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U 4.2 #
Typical Neutron Environment in the CRBRP Sh eld 4.2-459FJ T7 70 Assemblies 4.2-Shield Assembly Inelastic Strain Criteria 4.2-45 c.,
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. LIST OF FIGURES Continued Page No.
4.2-32 Maximum Fuel Rod Bundle-Duct Diametral Clearance 4.2-512 and Interference vs. Time - F/A 45. First Core 4.2-33 Maximum CRBR Blanket Assembly Bundle-Duct Inter-4.2-513 ference and Clearance vs. Time 4.2-33A Designation Scheme for Seismic and Core Restraint 4.2-514 Loads 4.2-33B F/A Shield Block Thermal / Mechanical Model Dimen-4.2-515 sional Extents and Finite Element Detail 4.2-34 LMFBP. Fuel Development Steady State Irradiation 4.2-516 Testing Program 4.2-35 Number of Fuel Rods Exceeding FFTF/CRBRP Goal 4.2-517 Burnup 4.2-35a Design Fallback Positions 4.2-518
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4.2-36 Reactor Elevation 4.2-519 4.2-37 304 Stainless Steel Core Support Structure 4.2-520 4.2-38 Core Module Liner 4.2-521 4.2-39 Co e Support Structure Secondary Plenum Concept 4.2-522 4.2-40 Elevation of Typical Lower Inlet Module 4.2-523 4.2-41 Inlet Module Core Map Showing the Orificing 4.2-524 Plan for the Compactor Assemblies 4.2. 41 A Bfpass F1ow Module Assembly 4.2-525 4.2-41B Bypass Flow !!odule Section View 4.2-526 4.2-42 The Fixed Radial Shield.ing 4.2-527 h2-43 Deletedh Rp_etm_E. wm (ws6er 4-W \\
4.2-44 Horizontal Baffle 4.2-529 4.2-45 Elevation of the UIS in the CRBRP 4.2-530 4.2-45 A InstrumentationPost(BenchAssembly) 4.2-531 51
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4-xxx Amend. 51 Sept. 1979
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INSERT 4-xu 4.2-43 Removable Radial Shield Assembly 4.2-528 4.2-43A Loading Histogram Used in Inclastic Analysis of 4.2g528a RR5 ACLP s
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The detailed formulation of the design criteria of Table 4.2-7 and the conditions for which they are applicable, are given in the CRBRP fuel assembly stress report, Reference 171.
d NSE.tt 4.2.- f1 (
4.2.1.1.2.3 Requirements for Design Features In addition to the preceding operational requirements, specific design features shall be incorporated into the fuel and blanket assembly designs to preclude the accident conditions discussed in Chapter 15.4 and any detrimental effects which could adversely affect the attainable design life.
1.
Sufficient constraint shall be applied to the fuel and blanket rods to minimize fretting and wear at the support points.
2.
The fuel and blanket assembly materials shall be compatible with ad-joining materials and environmental conditions during their design lifetime. Where potential for excessive galling or self-welding exists, mating components shall be hard coated.
3.
The relative location of the pellet column within the fuel and blanket rods shall be maintained during shipping to prevent damaging reactivity fluctuations during start-up by utilizing a properly designed axial spring support system.
4.
The assembly axial support system shall maintain fuel and blanket assembly
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axial positions under all steady state and transient operating con-ditions, while providing for differential thermal expansion of the internal structures and the irradiation induced expansion of the assamblies.
0 With the current CRBR baseline design, the limit is 2.5 inches at 70 f.
5.
Th)inletnozzlesfortheassemblies,inconjunctionwiththereactor internals, shall be designed with sufficient aperture redundancy to preclude total inlet blockage and to provide for adequate cooling even after total blockage of one inlet passage.
6.
To prevent loading of a fuel or blanket assembly into a position where it is undercooled, the folicwing situations must be prevented by a properly designed discrimination system.
Fuel assembly insertion into a position which it w:uld be under-a.
cooled, i.e., a position in which more coolant flow through the l
assembly is required for heat removal than can be admitted by the flow orifice in that assembly or receptacle.
i b.
Fuel assembly insertion into positions in the core that are pro-vided for the control rod assemblies, blanket assemblies and re-movable shield assemblies, except for those positions where fuel 51 and inner blanket assemblies are intentionally interchangeable.
o c-Amend. 51 4.2-17 Sept.1979
j INSERT 4.2-17 (Page 1 of 4)
New Section 4.2.1.1.2.2.2 Removable Radial Shield Assembly (RRS) Structural Component Design Criteria The design. criteria were selected so that the RRS structural components will satisfy their functional requirements described in Section 4.2.2.1.1.9 a
at the following operating and loading conditions over their design life within the constraints of the damage severity limits of Table 15.1.z-I.
(A) Operating and Loading Conditions 1.
Thermal-Hydraulic:
Steady state and transient conditions shall be considered.
The umbrella transients for the shield assembly incloat. nomal, upset, emergency and faulted categories. t b appropriate number of events will be detemined for the hottest shield assemblies based upon the designed replacement schedule, shee he d
% sv.edaucarn
.J. A b.u tear h = 3o ye., );te.
2.
Mechanical Loads b a 4.,.a4. 3 Joading' sources to be considered in the structural evaluation of Shield Assembly Duct structures; coce. re +ra. t, seh., h.a.a m.se cu a... n e p; e.+s m cur.a ena These loadings represent a generalization of all the currently post-ulated loading mechanisms applicable to CRBRP core assembly duct structurcs. While scre cf these loadings may be r.egligible or in-applicable to the shield assemblies, they nevertheless form the minimum basis for the conservative evaluation of the design adequacy of the shield assembly duct structures. The detemination of structural loads shall utilize the environmental conditions in~ ac-cordance with the requirements of Table is.s.2 -i l
l
4 INSERT 4.2-17 (Page 2 of 4) 3.
Nuclear Environment Table 4.2-t.? sumarizes the range of neutron environment of the shield assemblies.
(B) Structural Design Criteria The shield assembly is not currently covered directly by an existing structural design code. The shield assembly structural design criteria.. samnf s.d.t-r 4. K.a b 4.t i.t.2.2.i C., Li..a W. u + a u 6ftes. m. C.ii.
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. w s,
a,se,,w sp'.t:6 ev:=.A wo... on.a C., -N d.,rv.* hs Itti.
1.
General Criteria for All Shield Assembly Structures g..u ) w ruw I u.+ o*.
what
=,s und to N statJ.nss a ir s +<-%,
t. uw..
... a.cw ;,T.W 4.7.
ro.
The shield assembly structural components shall be designed so that defomation due to mechanical loading, themal expansion, and 1
neutron irradiation induced swelling and creep does not produce gross interference with adjacent ce=ponents such that the equip-rnent functional requirements of 4.2.2.1.1.9 cannot be satisfied.
The effect of loss of carbon, nitrogen, and alloying elements shall be considered when determining the strength of the material.
I 2.
Comoonent Structural Design Criteria Shield Asser.bly Inlet Hardware:
In the inlet hardware (nozzle, transition, orifice, rod support),
where maximum temperatures do not exceed 800*F, the base inetal is ductile and the time dependent themal creep is expected to be negligible. Time independent ductile failure modes covered by the
INSERT 4.2 17 (Page 3 of 4) applicable ASME Boiler and Pressure Vessel Code are used as the basis for structural design requirements.
Consequences of shield assembly inlet hardware loss of struc-tural integrity are sign!ficantly less severe than those for pressure boundaries and permanent components. Therefore, the emergency stress limits for shield essembly inlet hardware are higher than those given by the ASME B&PV codeII), but the i
maximum stress intensity shall not exceed the minimum ultimate tensile strength.
Shield Assembly Outlet Hardware :
The material is expected to be ductile and thermal creep may be significant. Both the time-independent and time-dependent
~ ductile failure modes covered by the applicable ASME B&PV code, code cases and RDT standards are used as the basis'for the
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structural design requirements.
Tke. time iri ependent stress limits for normal and uspet condition d
' specified by the.ASME B&PV Code are used directly for outlet nozzle design.
' Thermal fan a.
failure and creep related failures (creep rupture, s
excessive strain, creep fatigue) are prevented by imposing the appropsh design limits in Table 4.2-70.
(1) he ASME B&PV Code is applicable to pressure boundaries and pennanent components. The shield assemblies are not pressure boundaries and are removable.
INSERT 4.2-17 (Page 4 of 4) 3.
Shield Assembly Duct Where a minimum uniform elongation greater than 3% cannot be demonstrated. :r brittle fracture is
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as considered,a potential failure mode aad specific analytical methods to be used in fracture nachanics evaluations a<< dueinged.
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4.
Design Limits for Inelastic Analysis Because of the inherent simplifications in elastic and simplified inelastic analyses which are offset by additional conservatism in the corresponding limits, inelastic analysis method may be used in lieu of the elastic analysis methods.
~~he structural criteria acc given in Table 4.2-70.
5.
Special Weld Requirements Since all of the joining is accomplished by mechanically pinning the components and seal welding, there are no structural weldments in the shield assembly.
.I gg.pt.gm wrtA 141ECT 4 E~ C
,58 4.
2.1.1.9 Removable Radial Shielding (RRS)
The removable radial shield assemblies are core assemblies wh h mechani 11y interface with the radial blanket assemblies, the core f
.er structure the inlet and flow bypass modules, and the upper interna These assem lies have the following functional requirements:
a)
Atte. ate neutron fluence to levels consistent with year li fett. for peripheral components, based en resid 1 total elongati ductility limits.
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.)
58 b)
Idaximize solid volume frac 'o within the limits required to provide adequate cooling.
To o tain adequate shielding with 31655 80% minimum soli olume is required.
c)
Maintain the RRS st JCtural integrit by limiting RRS lifetime to assure 1.0 per nt minimum residual uctility based on total elongation.
In ddition maximum operati steady state plus transient str ns must be less than 0.056 rcent at a biaxial stress rat' (longitudinal to circumferentia of 1:1 and less than 0.3 ercent at a 1:0 ratio. The allowat'e strain varies l
approxir ely linearly with the biaxial stress. tic.
I d)
The emovable radial shield assemblies are to be in alled a
removed by normal fuel handling equipment.
e)
Transmit lateral core restraint loads without contributin significantly to the magnitude of these loads, f)
Provide circulation path for cooling to preserve structural integri ty.
51 4.2-127 Amend. 53 Nov. 1950
INSERT 4.2-127 l
- 4. 2. 2.1.1. 9 Removable Radial Shield (RRS)
The removable radial shield assemblies are core assemblies (See Figure 4.2-43) which mechanically interface with the radial blanket assemblies, the core former structure, the lower inlet modules, the bypass flow modules, and the upper internals. These assemblies have the following functional requirements:
I a) Provide radiation shielding to ensure the structural integrity of the reactor pemanent components beyond the radius of the RRS for 30 years.
b) Provide a compact structural unit that can be handled in and out of 'the reactor by the nomal fuel handling equipment.
c) Transmit lateral core restraint loads without contributing significantly to the magnitude of these loads. -
d) Provide flow paths fr-th: ::di - :::1 ~; for cooling of the RRS to preserve structural integrity.
e) Provide a means for locating surveillance specimens in the RRS and a 4 thod for their expedient recovery.
The bases for these requirements are described in 4.2.2.1.2.9.
O
. _..I Recuirement - Provide vertical load reaction to the core m.
barrel.
The core fomer structure is subjected to mechanical Bases -
loads caused by the incremental themal growth of the fuel assemblies. Due to stick-slip friction of the assemblies and alternating heating and cooling cycles, a net upward force may be developed on the core former ring. Additionally, the SMBDB loadings can cause upward forces on the upper core j
l fomer.
'(2.2.1.2.9 Removable Radial Shielding (RRS) a) Requirement - Attenuate neutron fluence to levels co istent ased on with a 30 year lifetime for peripheral component e following residual total elongation ducti y limits:
Shie ng Material Remote from Attac nts 5%
Shieldin S.aterial Attachments 10%
Core Barrel, ore Fomers, and essel 10%
Bases -
The component yond the radius of the RRS assemblies have thirt yea 'ifetimes. This includes fixed m
radial shielding,
..e core bar
, core fomers, and the reactor vessel the major items. For each of these items, there are 1.. ts established for max allowable fluence to assur a certain level of residual du lity of the stru ral material after thirty years. Ap imate fluence, al. E:0 MeV) limits to obtain the above duc
- ity limits Ref.1)areasfollows:
Structgral fixed radial shielding (with t n percent duct 1 at 800 F) has a fluence limit of 1.3 x 10 2 (n/cm2},
RFpt.4.E WITg i A/JE g7
- 4. b if f Amend. 51 Sept. 1979 4.2-151
INSERT 4.2-151 (Page 1 of 2) 4.2. 2.1. 2. 9 Removable Radial Shield (RRS) a) Requirement - Provide radiation shielding to ensure the structural integrity of the reactor permanent components beyond the radius of the RRS for 30 years.
Bases - The structural integrity of the reactor pemanent components is based on the requirement that at the end of the 30 year life the minimum total residual elongation of the materials must be not less than 10%.
l b) Requirement - Provide a compact strt.ctural unit that can be handled l
in and out of the reactor by the nomal fuel handling equipment.
Bases - Efficient and economical means of installing and removing l
i the RR5 by the fuel handling equipment. This is achieved by common-ality with the fuel and blanket assembly design, f
c) Requirerent - Transmit lateral core restraint loads without signif-icant contribution to the magnitude of these loads.
Bases,- The load buildup at "on power" conditions must be minimized to iner' ease the margin of safety against duct crushing. Also, it is desired to attain lower withdrawal forces at reactor refueling con-I ditions.
1 d). Requirement - Provide flow paths for the sodium for cooling of the RR5 to preserve the structural integrity.
Bases - Coolant flow must be provided inside the assembly to maintain the maximum material temperature below a reasonable limit (s1100*F),
and to limit the maximum temperature gradients due to nuclear heating so that t resulting thermal stresses will not exceed the allowable limit.
e) Requirement - Provide a means for locating surveillance specimens in the RRS and a method for their expedient recovery.
INSERT 4.2-151 (Page2of2)
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Bases - As described in 4.2.2.1.1.12, surveillance is required to obtain information on materials irradiation damage so that changes in material properties can be monitored and potential deteriorating conditions detected.
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9.69ALLp 61 INSEli A.2-ISI Nonstructural fixed radial shieldin (withfivepercentductility at 800*F) has a fluence limit of 2.
x1022(n/cm2).
he Core Barrel and Vessel (with ten percent ductility at 800*F 2
h ve a fluence limit of 1.3 x 10 z (n/cm2).
b) Req 'rement - Maximize solid volume fraction within the li ts requi ed to provide adequate cooling. To obtain adequate shield g with 316 SS eighty percent minimum solid volu is require Bases - Ba d on 316 SS, the minimum solid volume f ction is 807. to obta) the required residual ductility of t e midpoint of the core, wit the fixed radial shielding and th core barrel at five percent a ten percent ductility respecti ely.
c) Requirement - Mai 'ain the RRS structural in grity by limiting l
RRS lifetime to as re 1.0 percent minimum esidual ductility based on total elong tion.
In addition m imum operation steady-l state plus transient rains must be les than 0.056 percent at a biaxial stress ratio longitudinal t circumferential) of 1:1 l
and less than 0.33 perce t at a 1:0 r io. The allowable' strain varies approximately line -ly with e bi. axial stress ratio.
Bases - The strain limits gi n a ove are considered conservative
(
since a safety factor of thre
.s applied to rupture strain l
predicted from tensile testing o determine allowable strain.
This safety factor is consis nt ith safety factor levels employed in ASME Code crite a.
has been oburved from stressed tube experiments at rupt re strain w ries with stress state. For a 1:1 biaxia stress rat the predicted rupture strain is only one-sixt that of unia al tension tests. Thus this factor has been i cluded.
The minimum total f uence (E>0 MeV) assoc 1 ted with 1.0 percent residual total el gation is 0.875 x 1023 ( cm2) at 800'F and 1.2 x 1023(n/cm at 900*F (Ref. 29). The e of energy der.endent dama-functions will increase the uence limits I
beyond those ven above.
The minimu RRS lifetimes based on the one percent ductility limit are as follows:
Lifetime (Years) De nding RRS w Number on Position Within R 1
4.9 to 8.8 2
8.8 to 16.9 3
15.8 to 31.6 4
30.0 to 61.1
(
Amend. 51
.2-152 Sept.1979
h REe. ace D RV WJ6.cf 4."l-Ifl
- r., )
1 Thus some of the assemblies in row 3 and all of the assemblies in
' row 4 will not need to be replaced. A surveillance program is pected to extend the lifetimes beyond the minimum predicted li times.
The abo strain and ductility limits represent reasona e criteria ich reduce complexity and expense of RRS emblies and increas reliability as compared to less strin nt criteria.
d)
Requirement - Th removable radial shield ass. lies are to be installed and remos d by normal fuel hand 11 equipment.
Bases _- This requiremen affords effici t and economical means of installing and removin the remov e radial shielding by using the existing fuel han ing e ipment.
e)
Requirement - Transmit latera c e restraint loads without significant contribution to e ma itude of these loads.
Bases - It is desired minimize load 11 dup at "on power" conditions to provid a margin of safety inst duct crushing.
' Also, it is desire to attain lower withdraw forces at reactor l
refueling condi 'ons.
-)
f)
Requirement - Provide coolant paths for circulation -ooling to y
e structural integrity of the assemblies.
preserve Sufficient coolant must be provided for the follow' g two Base The maximum material tempergture must be kept with ons.
reasonable limits (approximately 1100 F) and the maximum allowable temperature gradients from irternal nuclear heating must be detemined by examining the allowable themal stresses.
51
' 4.2.2.1.2.10 Core Fomer Structure (CFS) a)
Requirement - Provide peripheral constraint for the reactor assemblies.
Bases - Positioning of the core fomer structure relative to the core maintains, as part of the core restraint system, proper core gecnetry during all modes of operation.
b)
Requirement - Provide lateral location and constraint for the lower end of the Upper Internals Stri.cture.
Bases - Maintain alignment of the Upper Internals Structure for Interface' constraints proper operation of the reactor control system.
are imposed by the In-Vessel Transfer Machine, the reactor vessel, 58 and the rotating plugs.
)
Amend. 58 4.2-153 r;oy, 1930
.8
The. RSS Asteaklics Art r e m o ** b te eer< assc%blies (3ta. F:3*'s 4*l*43) Aa.riaj s+,u4va bas: cutty simhr 4. -N. (ed A bt= k t4.
aim-q s4rme/, er a,
.oN4..stie, q een:M;-g.4 i-t:4 %.wls, ku %
.t A s+ +,4
.4 4.2.2.2.1.9 Removable Radial Shield (.f RS)
)
M&4ftia.1 $8 The e:di:1 shield :: c"Ite: :rc made up of stainless steel rods heMwithin thin walled i eni ducts. These assemblies are designed to be as flexible as possible in order not to contribute to the off-power restraint loads. A close-fitting support block is inserted inside the duct at the ACLP to provide axial restraint for the shield rods and to absorb seismic loads that are transmitted 4k<. 3, A RR5 13 fr** Mad by through the ACLP to the core former. ccm4r.1.f (lo.a 8
stack. of orMite. pla+as tota 4sd W idt 4kt intet,sottle A
4.2.2.2.1.10 Core Fomer Structure The core fomer structure is composed of three substructures,.
the upper core former ring, a spacing cylinder, and the lower core fomer ring. The core fomer rings are comprised of profile milled segments assembled into continuous rings, as illustrated in Figure 4.2-46.
The abov'e core load plane fomer ring, called tha lower core former ring, is mounted on a ledge machined in the inner diameter of the core barrel.
The spacing cylinder, called the support ring, provides holddown for the lower core fomer ring and support for the top load plane fomer ring called the upper core former ring. The upper core former ring has six lugs that fit slots in the top of the core barrel to transmit seismic'and other loads to the core barrel. A series of L-shaped keys are circumferen-tially slipped into the groove on the inside of the core barrel, between.
each of the six lugs, and trapped by means of a radially oriented dowell pin on either side of each slot. These L-shaped keys prevent vertical
}
displacement of the core fomer rings. The upper core former ring is centered in the core barrel cavity by means of the six radial lugs. The lower core former ring is centered in the core barrel cavity by means of radial shims.
4.2.2.2.1.11' Maintainability All the reactor internals except for th r ::ter assemblies, are designed for a 30 year life with no scheduled maintenance.
- However, provision has been made to permit removal of the lower inlet modules to assure full plant life and malfunction recovery capability.
Contributing factors which may require malfunction recovery capability include:
1.
Potentia damage to the reactor assembly receptacles, as a result of insertion and removal of reactor assemblies.
2.
Potential wear or partial plugging of strainers and orifices, as a result *of coolant induced changes.
58 4.2.2.2.1.12 Surveillance g,
Material surveillance coupons are contained within special
- ted " removable radial shield -petitiband a fuel
- -M ic 51 transfer and storage assembly.
In addition to these special assemblies, 1
Amend. 53 4.2-175 Nov. 1930
4.2.2.3.1.3 Horizontal Baf fl e (HB). Fuel Transfer a Storage Assambiv (FTLE f-
~
and Fixed Radial Shlald (FRS)
The 2, FT&SA, and FRS are Internal s structures and are not covered by mandatory Code rules, but the Owner's designee has required that the rules stated in 4.2.2.3.1 be applied to the design and analysis of these Niga T
- 4. 7 ~ N ** j I
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e 4.2-1760 Amend. 69 July 1932
~
INSERT 4.2-176a New Section
- 4. 2. 2. 3.1.4 Removable Radial. Shield (RRS)
The RRS is a replaceable core component with a structure basically the same as those of the fuel and blanket assemblies. Therefore, the struc-tural design criteria applicable for the RRS are also basically similar to those of the fuel and blanket assemblies, rather than those described in 4.2.2.3 for the permanent Reactor Vessel Internals components and structures. The RRS design criteria are described in 4.2.1.1.2.2.2.
o O
e
-ay
.. L Conclustens and Future )(ork Assembly motion reactivity of fects are conservatively predicted by current
' ')
analytical procedures. The results shown Indicate that reactivity related
' core restraint requirements of Section 4.2.2.1.2.8 ore satisfied.
Core component contact loads and distortions are predictable using current analytical methods. Additional work is planned to verify dilation Induced duct-to-duct contact predicted In NUBOW-30 with more detailed models.
Additional areas where further work is planned include:
- 1) Detailed analysis of core restraint performance beyond core 1.
- 2) The simulation of fuel management in the NUBOW-3D model.
- 3) Verification and improvement if necessary of the duct dilation Induced duct-to-duct contact model. f-REPL4cE WLTW l NJEd7 4.2-22 N 2.4.4 Removable Radfel Shieldino (RRS)
The remov e radial shielding Is in a preliminary phase of design;
- ius, stress analys taking into account the of fects of environmenta onditions has not yet been
.pleted. Analysis will be conducted on following considerations: the..
stresses and strains, refuelin d handling stresses, strain limits brittle materini, and e cts of Irradiation induced swelling and creep o + e core restrain stem.
n._
Pre!Intnery thermal analysis has show +
maximum RP.S tanperature to be less than 920 F when cooled with bypass atural circulation. Thus, it appears that bypass flow will pr de adequa ooling.
More detelled enalysis ow In progress in both th
- ermal and stress categ6tles.
Shielding an sis, to date, has shown a solid volume fraction o, t least 80%
316 stai - ss steel is required to meet ductility requirements on th lxed radi shielding and core barrel. This solid volume fraction is compati, h coolant channel volume requirements for adequate cooling.
4.2.2.5 Eelding and S 12fng gi Reacto.!nternal Parts The desten consideratiens for welding end seizing of rotating or moving parts for reactor internals are presented in Table 4.2-64.
~
4.2-227 Amend. 6S May 1962
INSERT 4.2-227 (Page 1 of 2) 4.2.2.4.4 Removable Radial Shield Assembly (RRS)
The RRS configuration is shown in Figure 4.2-43. The structural analyses included two critical regions of the RRS, the Above Core Load Pad (ACLP) of the duct and the outlet nozzle. The selection of these two regions was based on structural analyses of the basically similar fuel and blanket assemblies which indicated that the assemblies inlet regions comprising the inlet nozzle, orifice assembly and transition had very high design margins, while operating under more severe environmental conditions than the RRS. The *L
..w... a swe%< t.am toh4 e % Aa.?is he,,m sy t.aa as;w.
In the ACLP analyses, maximum values from all loading sources were assumed to occur simultaneously at the single assembly subject to the worst environ-mental conditions. The various loading conditions considered were seismic, core restraint, and steady state and transient themal loads. Three dif-ferent sets of thermal-hydraulic conditions were considered from which the worst combinations of maximum duct midwall and cross-duct gradient tempera-tures was obtained. Stresses, strains and damage were detemined for the l ACLP resulting from the above loading sources and conditions.
An ANSYS fipite element computer code was used for the RRS duct stress analyses. Based on the results of this analysis, inelastic and fracture toughness analyses of the ACLP were also performed.
The inelastic analyses were perfomed to detemine strain and damage of the ACLP.using the CHERN elastic-plastic-creep computer program. Two separate sets of analyses were performed, one representing BOL conditions, and one representing E0L conditions corresponding to a design life of 10 years.
Twenty ou,t of the 190 loading cycles of the unit histogram (see Figure 4.2-43 A) were run for each of the two conditions to overcome the transient portions of the elastic-plastic-creep domain, and the remaining 170 cycles th were conservatively duplicated by the 20 cycle. Both the calculated total maximum strain and the calculated maximum creep-fatigue damage when compared with the minimum allowable values yield positive design margins equal to 14.0 and 0.14, respectively.
INSERT 4.2-227 (Page2of2)
A fracture toughness analysis of the duct ACLP was also perfomed, since the minimum elongation in the load pad at end of life is < 31, and brittle fracture is a potential failure mode. For this analysis, an initial flaw size and shape, and two orientations were assumed, con-sistentwiththerequirementsdefinedinRDTE6-20T(nowNEE6-20T). The l
analyses consisted of the following steps:
1 a) The critical stress intensity factor KIC for BOL and EOL material con-ditions was derived from data on unirradiated sub-compact test speci-mens made from FFTF duct material.
b) The stress intensity factor for the assumed semi-elliptical flaw under tensile stress was determined.
c) A crack propagation analysis was perfomed to detemine the crack growth under the applied duty cycle.
d) The maximum stress intensity factor K was detemined for the final flaw size obtained from the crack propagation analysis.
e) Finally, the margin of safety was detemined by comparing the calculated maximum EOL stress intensity factor K with the allowable design value max which was obtained by applying a design margin of 1.5 on the calculated critical stress intensity factor K This evaluation yielded a positive IC.
margin of safety equal to 5.7.
1
ADD TABLE 4.2-Ff 70 e
TYPICAL NEUTRON ENVIRONMENT IN THE CRBR SHIELO ASSEMBLIES Total Neutron Flux, Fast Neutron Flux, RRSA Elevation E>
9 0 MeV E
>0 1 MeV Region (inch)
(n/cmZ-sec)
(n/cmZ-sec)
Maximum Minimum Maximum Minimum Outlet 11 10 10 9
Nozzle
-344.15 2.3x10 B.1x10 1.5x10 3.0x10 hIIN$ct 13 12 13 12
-405.15 6.1x10 8.1x10 1.5x10 1.1x10 Above Core 14 12 I3 12 Load Pad
-411.15 1.4x10 7.0x10 4.4x10 1.4x10 s k te.t a g
- 4 15 13 14 12 (fere. Nid r%) -437.15 1.2x10 2.4x10 5.1x10 5.7x10 I*lef M.ssk 14 12 13 12 hwsif tpw,
-469.15 1.3x10 5.5x10 3.4x10 1.1x10
!e
-512.15 1.3x10 1.5x10 1.2x10 9.9x10 II 11 10 9
e 0
O I
4.2-459{
l *,,.
ADD TABLE 4.2.wr}/
SHIELD ASSEMBLY INELASTIC STRAIN CRITERIA Normal and Normal. Upset and
_ Category Upset limits Emeroency Limits Membrane Plastic 8e Strain Limit I (c *) 1 0.25 I([or] < 0.35 u
y Total Inelastic 8c I(,f/TFr)10.5 6c' Strain Limit I ('f/TF
--) < 0. 7.
Creep-Fatigue Damage' D e t. n o *i. T a le 4.1 I 5
6 6
e O
G O
4.2-459j/c
Inf OUTLET NOZZLE h
l SURVEILLANCE SPECIMENS LOAD PAD % % q (AS REQ'D) i r
SHIELD RODS eru 1 A 4E I Tc u I 3 Y I I I y
%7J%.7
- SHIELD ROD REGl0N (57 IN.)
~ s_
DUCT ORIFICE ASSY.-
l S
OVER ALL LENGTH -14 FT.
INLET N0ZZLE y
i DISCRIMIN ATION POST i
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