ML20244B633
| ML20244B633 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/31/1981 |
| From: | Turnock R PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20039B221 | List: |
| References | |
| PGE-1013, PGE-1013-R03, PGE-1013-R3, TAC-03606, TAC-3606, NUDOCS 8112220436 | |
| Download: ML20244B633 (200) | |
Text
{{#Wiki_filter:* o ..
, PGE-1013
. . . 1
/J ( [ ,~ _.i>f 'e -
7
?%u .l/ ,-l l , . a ,. G, , -
N -
) ,
'A -
/'
' N,' * % -
SPENT FUEL POOL STORAGE RACK
~ 'I./ -
j
=-
L DESIGN REPORT FOR THE TROJAN NUCLEAR PLANT s 's { < N
\ . f.r
~ -
_g ,
, . . , - +- i
, og ' /
\ ^
[b f BI
, 1 4 1.(x, l
,'., u r s
. - M. ge \ R. K. Turnock
,. , \ Coordinator, PGE e t i
,N : Y-g
< G ..
December 1976 ,
-1 O '
j l* /) '. f Portland General Electric Company 621 S. W. Alder Street:
' ,f - '. Portland, Oregon 97205 y [
^l -
),
N
- v. p
-/
V
' /s '
p, j y 4 y 7
~
cuannan Ihrtland
) ,,
* ~
v, r' % d( @ s\\ " General
- N ' Electric P6E Gimpany ..
v, 9
(}
~~#
-FGE-1013 Revision 3 December 1981~
1 SPENT FUEL POOL STORAGE RACK- 1 DESIGN REPORT FOR THE TROJAN NUCLEAR PLANT 1 i I 1 i l i December 1976 Portland General Electric Company 121 S. W. Lulmon Street Portland, Oregon 97204 O
)
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. :)
l i p- SPENT FUEL POOL STORAGE RACK t / DESIGN REPORT FOR THE V TROJAN NUCLEAR PLANT 4 CONTENTS Section Title Page 1.0 Introduction . . .. ..... . . . . . ..... 1-1 2.0 Summary. . . ... .. . . . . . . . . . ..... 2-1 3.0 Spent Fuel Storage . . . . . . . . . . . . . . . . 3-1 3.1 Description and Design Bases . . . . . . ..... 3-1 3.1.1 i Structural and Thermal Expansion Analyses. .... 3-3 c i, 3.1.2 Criticality Analysis . . . . . . . . . . . . . . . 3-10 3.1.2.1 Analysis Overview for Delayed Modification . ... 3-16 i 3.1.3 Seismic Analysis . . .. . . . . . . . . ..... 3-17 l 3.1.4 SFP Liner Plate and Fuel Clad Integrity. . . . . . 3-19 ; 3.2 Spent Fuel Pool Cooling and Demineralized System . 3-21 3.2.1 Design Bases . g !
. . ' . . ... . . . . . . ..... 3-21
- 3.2.2 Design Description . . .... . . . . . ..... 3-22 3.2.3 System Description and Operating Modes . ..... 3-24 3.2.4 Instrumentation. . . . . . . . . . . . . . . . . . 3-28 g
l 4.0 Safety Evaluation. ... . . . . . . . . . .... 4-1 p 4.1 Spent Fuel Pool and Spent Fuel Fool Rack Safety ' (. Evaluation. . . .. . ...... . . . . .... 4-1 4.2 Spent Fuel Pool Cooling and Demineralized System , Safety Evaluation . . ... . . . . . . . .... 4-7 l$ , t , 5.0 Radiological Evaluation. . . . . . . . . ' . .... 5-1 5.1 Source Termc . . . . .. . . . . . . . . . .... 5-1 5.1.1 Activities in Spent Fuel . . . . . . . . . .... 5-1 5.1.2 Activities in Reactor Coolant and Refueling Water 5-2 l 5.1.3 Tritium. .. .. . .. .............. 5-6 5.1.4 Activities in Spent Fuel Pool Cooling and Demineralized System. . . . . . . . . . . .... g j 5-7 v ; 5.1.5 Activities in Ventilation Air. . . . . . . . . . . 5-8 5.1.6 I j Environmental Releases . . . . . . . . . . .... 5-9 5.2 Radiation Doses. . . . . . . . . . . . . . . . . . ) 5-12 j 5.2.1 Doses to Plant Personnel . . . . . . . . . .... 5-12 5.2.1.1 Direct Radiation Dose From Spent Fuel Assemb'ies . 5-12 ] 5.2.1.2 Direct Radiation Dose From Activity in Water . .. 5-13
.)
5.2.1.3 Dose From Airborne Isotopes. . . . . . . . .... $ j 5-13 5.2.1.4 Miscellaneous Sources of Exposure. . . . . . . . . 5-15 j 5.2.1.5 Plant Man-Rem Doses. . ..... . . . . . .... 5-16 4 j 5.2.1.6 SFP Modification Exposure. . . . . . . . . .... 5-17 5.2.2 Site Boundary Doses. ... . . . . . . . . .... , j 5.3 5-18 $ Disposition of Existing Racks. . . . . . . .... 5-18 3 1 1 6.0 Need for New Spent Fuel Storage Rackts. }
. . .... 6-1 O 7.0 Tests and Inspection . ..... . . . . . .... 7-1 !
I l l l i 1 Revision 3 { (December 1981) 1
SPENT FUEL' POOL STORAGE RACK DESIGN REPORT FOR THE TROJAN NUCLEAR PLANT CONTENTS Section __ Title .g 8.0 References . . . . . . . . . . . . . . . , . .., g.1 r-Appendix A Responses to Requests for Additional Information . 'A-l' - Appendix B Summary of Models. . . . . . . . . . . . . . .' . . 'B-1 < h. Appendix C Supplemental Seismic Analysis of 7 x 8 Rack Module. ..................... C-1 j^
'l
- l O . 1 11 Revision 3 (December'1981)' j
.'J
SPENT FUEL POOL STORAGE RACK DESIGN REPORT FOR THE
/9 TROJAN NUCLEAR PLANT 'w)
TABLES Table _ Title _ 3-1 Beginning of Life Critical Configurations for PWR's 3-2 Physical Parameters of Re"erence Calculations 3-3 Results from the Evaluation of Two Of f-Center Cluster Loading Configurations 3-4 Locating Frame Stresses m C 3-5 Embedment Loade 3-6 SFP Chemistry Specifications O 3-7 SFP Sampling Schedule t v 4-1 SFP Heat Load Summary 5-1 Cumulative Fission Product Inventories in Fuel at 4 Days after fg 8th Fuel Cycle 5-2 Peak Concentrations of Fission and Corrosion Products in Refueling Water at 4 Days af ter 8th Fuel Cycle (3-1/3-Core Storage) 5-2a Peak Concentrations of Fission and Corrosion Products in Refueling Water at 4 days after 8th Fuel Cycle (1-1/3-Core Storage) 5-3 Peak Concentrations of Fission and Corrosion Products in Refueling Water at 18 Days After 8th Fuel Cycle 5-4 Parameters Used in Tritium Analysis 5-5 Tritium Concentration in Plant Water Systems (pCi/g) 5-6 Accumulated Activities in Fuel Pool Purification System After m 1 Year Accumulation IO 5-7 Activities in Ventilation Air From Refueling Cavity and Spent Fuel Pool During 8th Refueling (pC1/cc) 5-8 Maximum Releases to Atmosphere From Refueling and Fuel Storage j Operations (Ci/yr) l \ 0c 111 Revision 2
. (August 1977)
L-_-_____-_-_-_-__--______ _ _ _ _ _ - .
p' SPENT FUEL ~PODL STORAGE RACK - 3 DESIGN REPORT FOR THE TROJAN NUCLEAR PLANT TABLES. Table Title' ' 5-8a Expected Change in Release Rates of Radioactive Gaseous g-
. Effluents from Spent Fuel Area (C1/yr)' +
Dose Rates at Pool Surface and Edges.From Isotopes-Contained in
~
5-9 ' Refueling. Water During 8th Refueling'.(area /hr) 5-10 Inplant Dose Rates.From Airborne Isotopes During 8th Refdeling (mrem /hr) , g ;- v3 5-11 Refueling Manpower Requirements'(3-1/3-Core Storage)- 5-11a Refueling Manpower' Requirements _.(1-1/3-Core Storage) 5-12 Maximum Site' Boundary Doses From Refueling and Fuel Storage Operations (arem/hr) 5-13 Expected Change in Site Boundary Doses From Spent Fuel- Area - g (arem/hr) v 5-14 Factor for Inplant Radiation Exposure Estimates for Spent Fuel Storage Rack Work 5-15 Radiation Exposure Accumulated During SFP Modification (mrem) B-1 Ratio of Allowable Stress to Feel Assembly Component Maximum-Stresses for the Uniform Loading Case at 400 Lb B-2 Ratio of Allowable Stress to Fuel' Assembly Component Maximum Stresses for Triangular Loading Case at '850 and 1090 Lb' C-1 Combined Stresses 'for 7 x 8 Module - SRSS Method (First 8 Modes)
.A-U iv Revision 3
.(December 1981) . r-c- _- .. .
q
- 3. a SPENT FUEL POOL STORAGE RACK DESIGN REPORT FOR TH7, 1 TROJAN NUCLEAR FLANT FIGURES j u
Figure Title 3-1 Spent Fuel Poo1' Arrangement 3-2 Embedment Module Locating Frame Arrangement 3-2a Locating Frame Details', Sheet 1 of 9 l J 3-2b Locating Frame Details, Sheet 2 of 9 p. 3-2c Locating Frame Details,-Sheet'3 of 9 e 1 3-2d Locating Frame Details, Sheet 4 of 9 3-2 e Locating Frame Details, Sheet 5 of 9 , , e : 3-2f- Locating Frame Details, Sheet 6 of.9 3-2g Locating Frame Details, Sheet 7 of 9 3-2h Locating Frame Details, Sheet 8 of 9 3-21 Locating Frame Details, Sheet 9 of 9 , 3-2j Cruciform Detail 3-3 Spent Fuel Storage Rack Installation , 3-3a Spent Fuel Storage Rack Installation Details:for Underwater ;; Installation o
~
3-4 Typical 6 x 8 Module Assembly-3- 5 Reference Design 3-6 Fuel Assembly Cross Section 17 x 17 3-7 Criticality Analysis Results 3-8 Quarter Section of a Sixteen Of f-Center Fuel Assembly Cluster 3-9 Reactivity Void.Effect, 200*F-3-10 Cluster Geometry of Four Off-Centered Fue1' Assemblies , 1 _s 4-1 -- Geometry of Fuel Assembly Drop Accident
' s J/ 6-1 Rack Installation Schedule j((
v Revision 2 L -(August 1977) i 1 k_ { - - I___ __m__ _ . _ . _ _ _'5-____s._m___
1
'~4 SPENT FUEL POOL STORAGE RACK is DESIGN REPORT FOR THE TROJAN NUCLEAR PLANT-1
. FIGURES l Figure Title B-1 General and Typical Module I b
1 B-2 6 x 8 Module B-3 6 x 7 Module ., B-4: 7 x 7 Module n B-5 7 x 8 Module D-B-6 Computer Model,'2 Module Floor Truss B-7 ' Computer Model, 1 Module Floor Truss "t B-8 Rack Module Model Detail B-9 Fuel Assembly Deflected Shape 400 Lb'Reac' tion Load . B-10 Fuel Assembly Deflected Shapes for Triangular Load Distribution l a i ( vi Revision 3> (December 1981) ______u_________ _ _ _ _ _ _ _ _ _ --:- _ _ . . _ _---- _a
l l l (]
1.0 INTRODUCTION
V- 1 i This report is submitted in support of Portland General Electric I Company's (PGE) Amendment 34 of the Trojan Nuclear Plant Operating h. License, NPF-1, to more fully utilize the storage capacity of the spent fuel storage facility. i { The modifications to the spent fuel storage facility allowed increased 1 flexibility in scheduling offsite shipment of spent fuel. The modifi- q cations were x complished without significant changes to the facility by replacing existing spent fuel' racks with racks of a high-density ~ o design. l . 1 Section 2.0 is a summary. Section 3.0 is the description, design i bases and supporting analyses for the modification. . Section 4.0 and Section S.0 are respectively the safety and radiological evaluations. j \ m , Section 6.0 explains tae conditions that necessitated the modification. W, A Section 7.0 includes the tests and inspections involved in the fabri-cation and installation of the racks. Section 8.0 lista references. 3 l Appendix A includes a summary of NRC requests for information and m. ) I u 1 l references PGE documentation of compliance on each of the issues. l l Appendix B is a summary of the models used to analyze the racks. Appendix C is supplemental detailed information demonstrating compli- ], l ance with Regulatory Guide 1.92 and NRC Standard Review Plan, < Section 3.8.4. l Revision 3, being af ter the fact, incorporates into this text PGE commitments made during the hearings that eventually resulted in the approval of Amendment 34. Throughout this text, " existing" denotes b l status prior to the expansion modification, whereas "new" implies l po st-modification . l l The design of the spent fuel racks was performed by the prime contractor p Programmed and Remote Systems Corporation (par), St. Paul, Minnesota. V The static and seismic structural analyses were performed by l 1-1 Revision 3
-(December 1981)'
.{
,., d Environmental Services- Inc. (ESI), tiinneapolis,1111nnesota. The 'criti-
~
\ cality and heat . transfer ' analyses < were performed' by Nuclear Associates i
. International Corporation (NAI), Rockville; Maryl nd. ';Bechtel Power Corporation, San Francisco, California, verified .the compatibility of the Spent. Fuel . Pool (SFP) Cooling and. Demineralized System with the new.
SFP design temperature limit. ~ The ; radiological evaluation was' performed . by PGE. t I 1 1 1 .g j i j l
- 1 b
.{ 1 ' Revision 3' l (December 1981)- ,!
-l
]
l l l
'l
/w 2.0
SUMMARY
C q 1 The modification provides safe storage for up to 651 spent fuel assem-blies in the SFP and included replacing the existing spent fuel racks- .j with those supplied by par. The modifications did not alter the 1 structure of the SFP or the supporting cooling systems. The new spent fuel rack design consists of arrays.of stainless'> steel to i form cavities having a nominal center-to-center spacing of 13.3 in, and . f 1 a wall thickness of 3/16 in. The preparatory work took place-in two phases. Phase 1 was ocheduled to be completed near the end of .0ctober i 1977. In detail, Phase 1 included. the following work: 1
- 1) To facilitate rack removal and' reinstallation', existing'
~
spent fuel rack leveling shims.which were tack-welded originally to the SFP liner plate were ground. loose and -j tack-welded to the existing spent fuel racks. . This action g permitted the existing racks to be easily removed and I h reinstalled.
- 2) To facilitate thin work required removal of the existing
, spent fuel racks a few at a time. Sufficient rack capacity m was being retained in the SFP during this phase to. permit i discharge of a full core.
- 3) The installation of the duet heater in the SFP ventila-tion exhaust system described in Section 3.2.2 and the reorientation of the SFP Cooling and Purification System diffuser header to improve' circulation of the cooling water in the SFP.
I Phase two, scheduled to commence in F wember 1977 involved the follow-l l ing work: .
< .,a
- 1) A sufficient number of the existing spent fuel storage
( racks were removed from the SFP.to provide adequate room
.q
'l i l
2-1 Revision-3 (December 1981)
(N for work described below. As many of the existing
'v) spent fuel storage racks as possible were retained in the SFP. Sufficient existing spent fuel storage racks could be replaced before the Trojan TechnicalLSpecifi-cat ions permit fuel to be moved into SFP (100 hr). l l
i
- 2) A sleeve was field-welded around each embedment stud. 1 i
Since this sleeve did not extend above the SFP liner .l plate, it did not interfere with the reinstallation of the existing spent fuel storage racks. The welds were inspected. ;
- 3) The alignment sleeves and embedment modules locating frames were located over the embedment studs. The l
embedment module locating frames were welded to the align-ment sleeves. Note at any point these components could i i be removed leaving the embedment stud ready to receive j
,q the existing apent fuel storage racks. l C
- 4) The module support cups were welded to the embedment module locating frames af ter they were located using a template and the new rack modules. The welds and tolerances were inspected.
- 5) The new rack modules were positioned in the SFP,
- 6) The new rack modules were leveled, and the top module-to-module was fit and welded.
i 1 1
- 7) The new rack modules and embedment module locating frames were removed and the existing spent fuel storage racks reinstalled.
l l The design description is comprised of structural, criticality, thermal h and seismic analyses that conform to applicable NRC regulations, NRC () guides, and industry standards. The design is supported by a safety l 2-2 Revision 3 (December 1981)
- j. >
]
1
'i ,
i Ll
'is l L and l radiological . evaluation and an assessment of the f need and schedule '
. for installation of new ' spent fuel' storage racks at Trojan' 1 - . .
-s .
i The 1974 edition ' of. ASME Section III and- all , applicable ' addenda up to the date of . the purchase . order between PGE' and par (January- 29,1976), . . were used for the design. b
\
f'
/ The design ensures that an: ef fective multiplication fac tor _-(k,ff) .of.
J J less than 0.95 will be maintained; / that adequate ' cooling during- postu- ' j
- lated ' normal and 'special operating conditions will be provided; and ' ,
2 that ' the structure will. withstand safe shutdown' earthquake (SSE) . ;;; v. loadings. ' The modifications- for the Trojan Nuclear Plant provide 1 safe storage for upito 651 fuel assemblies an'd are compatible with the ' plant.
)
design as' provided in the Trojnn Final Safety' Analysis' Report .(FSAR). 1) and Operating License'(NPF-1). I
' H l
a I 1 l
.]
, .I
.k 2-3 Revision 3-(December 1981);
y 4
.3 l
h,
-l
-)
3.0 SPENT FUEL STORAGE,.
~.The design basis for the spent fuel storage facility has.been. changed
~
from the storage of four regions to the.' storage of 10 'regionsfdescribed . lb: . in this report. The operating temperature design: limit of the SFP . water:is increased from.125'F to the~ design, limit described ir. this-report of 140*F.' 3 .1 DESCRIPTION AND DESIGN BASES b The SFP is a reinforced concrete structure with seam-welded stainless steel plate liners enclosing a pool volume of approximately 51,900 cu f t.y The reinforced structure of the SFP.is designed.in accordance with ;
]
codes described in FSAR Section'3.8.1. The SFP structure and spent :l i fuel racks are designed in.accordance with Seismic Category I require- -i ments. Gaseous radioactivity in the area of the SFP is designed to be maintained below the 10 CFR 20 limits. O The SFP provides a safe and reliable means of storage and facilitates handling of. irradiated fuel. No equipment or materials other than g_ spent fuel components and associated tools will be stored in the SFP. , The safety function of the SFP is to retain water and passively support spent fuel assemblies in a position amenable to natural circulation. cooling. The SFP is designed to accommodate 10 regions (plus one' spare cavity) of spent fuel in a suberitical array such that a keffj,0.95is maintained. ]
-l Accommodation in the SFP for 10 regions or 651 spent fuel assemblies l3 allows the concurrent storage of one full core (193 assemblies),- the' ,
spent fuel assemblies from seven normal refuelings ,(65 assemblies each), and one spare cavity. The spent fuel assemblies are stored.in cavities in parallel rows and'have'a center-to-center distance of 13.3 in. in both horizontal directions. Burnable poison rods removed from the ! reactor are stored in spent fuel assemblies. O
%)
3-1 Revision.3
- (December-1981) $
_ __ _ _ _ __ ________ _ ______ __ _ _ _ _ _ ________ _ ____ a__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . J
-1 i
The general arrangement of the storage space is illustrated in Figure 3-1. The embedment frame arrangement is shown in Figures 3-2 g i through 3-2i; details of Figures 3-1 and 3-2 are.shown in Figure 3-3. A typical module arrangement is shown in Figure 3-4. Figure 3-3a was submitted in anticipation of underwater installation. The decision to l use the details in Figure 3-3a was made'in September 1977, when the G. licensing and refueling schedules, indicated.that the rack ~insta11ation might have to be performed underwater. The water level in the SFP is maintained to a normal value of 91 ft 8 in.
.This provides at least 23 ft of water above the top of a spent fuel~
assembly in the storage racks and at least 10 f t above the actual ~ fuel rods during fuel transfer operations. This water barrier serves as a radiation shield, enabling the gamma dose rate at the pool surface to be maintained at or below 2.5 mr/hr. A drainage system is below the liner plates of the SFP, cask loading pit Ib - and fuel transfer canal. Maaual diaphragm valves are left opened to- Ib l ,Q monitor drainage into a manifold connected to the' Dirty Radioactive Waste 1^ 0 j\ System. The drainage system will be routinely inspected for leakage g during shift tours.
- Adjacent to the SFP are two smaller pools - the fuel transfer canal and l the cask loading pit. The fuel transfer canal is connected by the fuel transfer tube to the Containment refueling cavity. /> leaktight door is provided between the SFP and the fuel transfer canal. The cask loading pit is connected to the SFP by another leaktight door on the opposite side of the SFP from the fuel transfer canal. ~
The gaseo6s radioactivity from the atmosphere above the SFP is control- ' led by a, ventilation system. The.SFP Vent Monitoring System (PRM-3)' I continuously monitors noble gases in the vents from the SFP areas and ' alarms .if the gaseous airborne radioactivity level reaches a preset O limit. This system is described in Trojan Nuclear Plant FSAR O Section 11.4.2.2.2. U 3-2 Revision 3 (December 1981)
l l b C\ After passing through HEPA and charcoal filters, the exhaust from 'the i SFr is combined with the Fuel and Auxiliary Building Exhaust System. This combined flow is continuously monitored by the Auxiliary Building 3 Vent Exhaust Monitoring System (PRM-2) for noble gases, iodine and par-ticulates. This system is described in Trojan FSAR Section 11.4.2.2.2. ;
)
.)
The SFP Area Radiation Monitoring System is provided for personnel pro-
.1 tection and general surveillance of'the SFP area. Equipment in the ]
control room provides continuous monitoring,. recording and alarms. ] Audible and visual indicators are provided locally. Potentially radio- ] active crud at the bottom of the SFP can be removed by an underwater portable vacuum cleaner. C The importance of the SFP as a potential sabotage target has been recog-nized by PGE in developing the Trojan security plan. PCs has identified j the SFP as a vital area. To elaborate on security maasures refer to g v I PGE's most recent security plan. 1 I l 3.1.1 STRUCTURAL AND THERMAL EXPANSION ANALYSES I i The cavities in the SFP are on 13.3-in. nominal center-to-center spacing l l l and are welded into the following modules, arranged as illustrated in l Figure 3-1. I Module Module Total i Size Quantity Cavities (lbs) _(lbs) 1 6x8 7 336 > 20,400 142,800 6x7 5 210 17,850 89,250 3 7x7 1 49 20,825 20,825 7x8 1 56 23,800 23,800 l 14 651 82,875 276,675 The new racks occupy the same exterior space envelope in the plan' view as the existing racks.
'3-3 Revision 3' (December 1981)
R l I l
)
l t [D 1 'y The racks and their interfacing structures to the existing SFP floor embedments are constructed almost entirely of Type 304 stainless steel. The module threaded feet and top tie bolts are ARMCO Type 17-4PH stain- b less steel heat-treated to condition H-1100. These are precipitation-hardened stainless steel because of the galling and higher strength requirements. Material properties for ARMCO Type 17-4PH were taken from the ARMCO Steel Corporation, Advanced Materials Division Publicatica, and have been verified to be the same as established in the ASME Code Section II; ie, the ARMCO Type 17-4PH minimum yield values used are the ! same as ASME SA 564 for Type 630 steel minimum yield values. With S regard to the NRC comments concerning stress corrosion characteristics of Type 17-4PH for heat treatments below H-1100 series, the vendor's proposed use of heat treatment H-900 on the top tie-bolts was not per-mitted by PGE, and all Type 17-4PH material remained at H-1100. This is l compatible with with the pool liner plate and embedmonts of Type 304 g v stainless steel. p The analysis of structural loads imposed by dynamic, static, seismic i
\
and thermal forces is made in accordance with NRC Standard Review Plan, g Section 3.8.4, and is discussed in Section 3.1.3 of this report. The combined stress ratios are less than the allowable interaction factor, j
^ J therefore the corresponding safety margin is at least the difference d ]
between the allowable and the yield stress. l Allowable stress limits for combined loading conditions are in accor-dance with Section III, Appendix I, and Appendix XVII of the ASME Boiler & Pressure Vessel Code. Allowable stress limits for linear com-ponent support welds are defined in Table NF-3292.1-1 of Section III, Subsection NF of the Code. All materials in these calculations are based on Type 304 stainless steel manufactured to ASTM specifications of A-240 (sheets and plates), A-276 (angles and flats) and 4-312 (pipe).
- 1) Minimum Yield Strength "F" (see Table 1-2.2, " Yield Strength of Austenitic Steel", Appendix I, ASME P & V r]
/ Code, Section III, and see ASTM-A240 specification): V 3-4 Revision 3 (December 1981)
i l l 1 Fy = ~ 30,000 psi at 0.2 percent permanent set . at 100*F. Fy = 25,000 psi at 0.2 percent permanent set at 200*F.
- 2) Minimum, Ultimate Strength "Fu~ )
}
Fu = 75,000 psi at 40 percent minimum elongation. J
- 3) . Modulus of elasticity."E" (Type 304 stainless steel)
[see Table 1-6.0, " Module of Elasticity of Material", Appendix I, ASME F & V Code, Section IIIJ: E = 28.3 (106) psi at 70*F.~ E = 27.7 (106 ) psi at 200*F. ., l i
)
The irdividual cavities of a module are welded as an open-ended box ;
/% 1 section, 8.96 in. sq by 14 ft 2-1/2 in. long with a 3/16-in.'. wall. thick-V ness. Each cavity has a welded cruciform (see Figure 3-2j).in the bot-tom end to support the fuel bundles. The cavity length is such that the ,
top of the fuel assembly plus a rod cluster control. is 3-3/8 in. below . the cavity funnel top. The cavities.are first welded into columns of. six or seven (called 1 x 6 or 1 x 7 arrays) as shown in. Figure 3-4. The top funnel structure, the lower horizontal channels and the "X" bracing provide the structural integrity for interconnecting the cavities..'The arrays are then welded into seven or eight' rows. to inake up the modules. The side channels and "X" bracing are the interconnecting members; adjacent funnels are welded to form ' shear ties between the arrays. The end cavities of the end row array have. welded supports in-the' bottom of s the cavities which form the four-corner supports. for the module. Each support has a threaded pad which is used to level the module by means . of a long-handled tool reaching down the centerline of the corner cavi-ties. The corner supports raise the racks 11-7/8 in. ab'ove the SFF1 floor, thereby providing an unimpeded passage of. this height. for water . ( circulation under the racks. 3-5. Revision 3. (December.1981)
1 l l p). a The modules are supported in the SFP by the existing anchor bolt embed-l i ments and by tie plates between the tops of the modules. The loads, f1 which are mainly due to seismic forces, are transmitted from the sup- j ports of the modules to the embedment anchor bolts by stainless steel locating frames as shown in Figures 3-2 through 3-21. The frames are designed to evenly distribute the loads from the supports to the anchor ! bolts and eliminate, insofar as practical, increased shear due to moment couples on the bolt groups which would be inherent without the frames. . The stresses in the locating f rames are shown in Table 3-4. The modules j are designed to be installed in the dry pool. A sleeve is field-welded { 1 around each stud to increase its shear capacity to accommodate the cal- i a culated loads; the existing embedments are adequate for the calculated j I loads. The locating frames are placed over the studs and alignment sleeves field-welded to the frame plates after precisely aligning the sleeves relative to the studs. The cups for the module supports are field-welded to the frame as shown in Figure 3-3. Special lif ting and l welding fixtures are provided to assure safe handling and precise align-i [a t ment. The ties between the tops of the modules shown in Figure 3-1 (detail B in Figure 3-3) stabilize the modules against seismic forces s I and maintain the top spacing for criticality considerations. l l The differences in thermal expansion between the pool floor and the modules are accommodated in the interface (stud to plate sleeve) between the locating frames and the embedments so that there is no significant shear on the embedment for the case of thermal expansion over the 75'F to 212*F operating temperature range. The total clearances provide for b the dif ferential thermal growth between the racks and the floor. There- 3 fore, no significant induced horizontal thermal loads are on either the rack or the embedment studs. Rack temperature gradients horizontally due to a full rack next to an empty rack are not significant. Also, as described above, any resulting thermal growth in this direction is unconfined and no stresses result. b The largest vertical temperature gradient through the water inside a [ cavity tube, as calculated f rom the thermal analysis, is 23*F. If it is assumeu that the cavity tube has this sarne temperature distribution 3-6 Revision 3 (December 1981)
1 I (j (ignoring external water surrounding the tube which reduces this gra-Edient), and if one tube . is heated to a mean temperature produced by the ; 1 vertical temperature gradient and all other surrounding tubes. are not . l l (confining the heated tube), the resulting thermal' axial stress in this , D3 ' tube is approximately 3100 psi, which gives an axial interaction ratio ? I 1 of 0.19. When this maximum thermal stress interaction ratio .is com- 1 bined with those due to dead, live, seise.ic, and fuel bundle impact, :
.the total interaction ratio on the most highly. stressed _ cavity; tube l is 0.88, which is less than 1.00 and therefore acceptable.
t. The fuel pool 1/4-in. thick stainless' steel liner was initially designed' and functions as a leaktight barrier, and is not a fuel pool ~ structural ~ element. As described, the liner is capable of resisting the maximum ,. accident temperature conditions of 212*F without compromising leak' tight-ness. Although no code load combinations with respective stress: limits apply directly in this case, the fuel pool P.irer has been reevaluated for the governing load combination of deed, if.ve (hydraulic), seismic -s. () (SSE) and new thermal loads (140*F). Under this load combination, mem-brane stresses remain below the code value for structural elements of
.t 1.6 S per Standard Review Plan Section 3.8.4.
Similarly, combined stresses for the reinforced . concrete spent fuel pool walls remain within. allowable limitu for load combinations which -{ include the increased nominal design temperature of 140*F. The cups which horizontally restrain the bottom mo3ule support have a 0.015-in. radial clearance. Each sleeve has an outside and inside diam-eter respectively of 1.75 in, and 1.062 in. Each stud has an outside diameter of 1.0 in. The locating f rames which hold the~ cups have a ; 0.062-in. radial clearance with the sleeved embedment- studs. These '} clearances provide thermal expansion space for the module frames. Alternate details for underwater installation,' using the same clearances for shear transfer f rom the module cups to the existing embeds, are shown in Figure 3-3a. a 3-7 -Revision 3
'(December 1981)
_ _ _ - - - _ - - _ _ _ - - - = =
n k The tops of adjacent modules are clamped together with shear oolts which have a 0.008-in. radial clearance. The shear bolts are 3/4-in. bolts. Under thermal expansion from 70*F to 212*F, the top of the end module moves further than the bottom because the bottom is restrained by the g embedment; this results in a 0.012-in. deflection of the top of the module. This deflection is approximately 1/10 of the deflection experi-enced under seismic loading and does not significantly increase module loading. Finite element models used to determine module member stress for static 3 dead loading and seismic response conditions are described in Section 3.1.3. Model assumptions are that the floor trusses are ideal planar frames composed of flexural beam - column elements, that module feet reactions are applied at the cup location and that the embedments m are pinned to the floor. The model was analyzed using the SAGS ( ) com-puter code. Member stresses and embedment bolt stresses were compared against allowable limits as set forth in Section 3.1.3 of this report.
/^\
V The rack configuration and tolerances are designed to the requirements imposed by the seismic, criticality, thermal, hydraulic and interfacing considerations. For example, a bar is provided in the design on the out-side modules to preclude a dropped assembly from locating closer than 6 in. to the nearest stored fuel assembly. This dropped assembly condi-tion is discussed in Section 4.0 of this repert. The design also accommodates an impact resulting from a 2000-lb rigid object in the size and shape of a fuel assembly dropped from the maximum crane hook height above the rack funnel (10-in. drop). A prototype module was constructed consisting of four cavities connected to form a module which is representative of a cluster of cavities of the actual module design. An impact plate 8 in. sq was anchored to the bottom of a test weight to simulate the bottom fitting of a fuel bundle. 'Ihe test weight was suspended over the prototype module and a guiding structure was constructed around the module so that the impact plate would strike n the module at the funnel intersection of the four cavities. Three load (v) cells were mounted on the module base and connected to an oscillograph 3-8 Revision 3 (Deceinber 1981)-
i l l p] f, to record the force-time response during impact. Vertical drop tests were then conducted (in dry conditions) to determine the effects of the l j 20,000-in.-lb impact energy on the prototype module, and to provide data for further analyses. Observations and measurements after full-scale testing showed only very localized deformation at the funnel top j l (on the order of ,16 in.), which would not affect fuel bundle inser-tion or removal. The recorded impact time history was a triangular impulse lasting 0.006 sec. The equivalent static load was calculated, J
)
applied to the center of the largest full module in the computer model, l and the resulting member stresses were calculated. For this condition all structural members except the impact interface were below 90 percent of yield. Under the condition where the dropped fuel bundle does not strike the. l top of the rack and travels unimpeded down through the cavity, the fuel bundle will strike the bottom fuel support cruciform. The cruciform is- i designed to shear-out under this impact energy such that the fuel bundle 'I l hd will come'to rest on the fuel pool liner plate after shearing out the ] cruciform. Resulting stresses on the remainder of the rack module mem-bers will be less than yield.
)
)
The effect of a fuel bundle dropping vertically and then pivoting and hitting the rack was examined. The total net impact energy for this con-dition (approximately 13,500 in.-lb) is less than the energy of the straight vertical drop and resulting member stresses are less than yield. The hypothetical dropped fuel bundle accidents described above have been prototype-tested and analyzed to demonstrate design adequacy under worst- l case conditions. The criterion that .the fuel supporting structure remains elastic under accident impact conditions, such that the center-to-center spacing of the stored fuel is maintained, is therefore satisfied. In the unlikely event of an actual fuel rack impact incident, all affected structural elements would, of course, be immediately investigated and corrective action taken, as necessary, to ensure structural integrity and safe fuel assembly storage conditions. v 3-9 Revision 3 (December 1981) J
l 1 1
]
'l M
'The effects of upward loading due to the maximum force that'the crane would exert on a fuel assembly stuck inside a~ can was considered and i the stresses.due to this condition are well below' yield stresses, and. g.I are within code-allowable limits. -The racks are designed for stresses 1
due to loads in the downward direction which are greatly in excess of i 1 l any load that coul/. be applied by the crane in-the upward l direction. i
.I \
3.1.2 CRITICALITY ANAIXSIS d 1 l The cria.icality analysis of the spent fuel storage rack design was per-' forw.d by means of a' series of diffusion theory calculations utilizing -l tb CHEETAH-P/PDQ-7 model. CHEETAH-P 'is a modified version of the ] ' original LEOPARD code (3) and uses a modified ENDF/B-II cross-section, library. NAI has' extensively tested the-CHEETAH-P/PDQ-7 model through' benchmarking calculati:ns of measured criticals as well as through core physics calculations of operating power reactors. Based on NAI'. bench-I marking experience, the maximum model uncertainty is 0.008 Ak. This O represents the '.argest difference between analytical,results and mea-V sured critica).s in actual reactor cores. Table 3-1 lists'some of the 4 excellent ar,reements achieved for PWR beginning-of-life critical'- configurations.
']
To ensure that the criticality analysis followed a very conservative' ) m 1 approach and conformed to the general guidelines' for safety ; analysis, dH the calculations were performed with the following assumptions:
- 1) Enrichment: 3.5 wt% U-235
- 2) Fuel: fresh and nondepidted L
l 3)- Burnable poison rods: replaced by water holes
- 4) Control rods: ' replaced by water holes O
Q-3-10 . Revision.3-
-(Decemberc1981)
_ __ _ _:_ __ -Q
a n O x) 5) Minor structural members replaced by water l
- 6) Fresh pool water 9 l
- 7) Hard thermal spectrum in the water gap -l 1
l l
- 8) SFP water temperature of 68'F -
- 9) Macroscopic ti.ormal absorption cross sections (calculated):
~
a) Water at 68*F - 0.2636 cm
~1 b) Stainless' steel at 68'F - 0.0222 cm .
H The physical parametera assumed -for the referenea case (or base case) analysis are listed in Table 3-2. The basic cavity dimension is I 13.3 sq in. The rack cavity, which is made of Type 304 stainless steel p of 3/16-in. thickness with a design tolerance of +0.05 in, and -0.00 in., . has an outer dimension of 9.335 sq in. Since the fuel assembly itself I is 8.426 sq in., there exists a free space of 0.267 in. on each of the four sides between the fuel assembly face and the inside wall of the cavity. The reference cavity geometry is shown in Figure 3-5; Figure 3-6 gives a detailed sketch showing the dimensions inside a fuel assembly. A two-dimensional X-Y four group PDQ-7 model, including an axial buckling to account for the axial leakage, was used for the criticality analysis. A zero-current boundary condition was employed on all four outer bound-aries of a storage cavity, as shown in Figure 3-5, to produce an infinite array effect.- The reference case calculation result is k gg e = 0.9136. by the PDQ-7 model. In Figure 3-7 the reference case result .is compared with the results of an independent calculation using the multigroup, multidi:.nensional Monte Carlo neutron-transport KENO-Il(') computer ' code with the 16-group Knight-modified Hansen-Roach cross-section library. I ! The comparison confirms the validity of the diffusion theory calculation. l The sensitivity of the PDQ-7 model results to the basic cavity spacing l 1 3-11 Revision 3
.(December 1981) l -. _ __ ._ _ _ .
D was evaluated by performing calculations wit'n the spacing varied from [O 12.5 in, to 14.0 in. Figure 3-7 shows the calculated behavior of keft as a function of the cavity center-to-center distance. Because of the free space existing between a properly centered fuel assembly and the cavity wall, it is possible for an assembly to be loaded off-center in a cavity. Two extreme cases of off-center loading configurations were evaluated to determine ke gf. The first case was a 16-assembly cluster with assemblies loaded in their cavities off- i center and preferentially leaning toward the center of the cluster as shown in Figure 3-8. The zero-current boundary condition applied to the cluster outer boundaries produces an effect of an infinite array of these 16-assembly clusters in both directions of the X-Y plane. For this off-center loading of the assemblies, all other dimensions in the configuration were maintained as in the reference case. l l The second configuration, as illustrated in Figure 3-9, is four assem-Q blies loaded off-center with all assemblies leaning toward the center l of the cluster. The zero-current boundary condition also produces an infinite array of four-assembly clusters in the X-Y plane. In addition to the effect of clustering, this second configuration includes the worst condition design geometrical and mechanical tolerances. The center-to-center spacing was reduced by 0.06 in, from 13.30 in to 'l 13.24 in. The Type 304 stainless steel canisters with the reference wall thickness were also enlarged by 0.06 in., making each side of the-square 9.395 in. instead of 9.335 sq in. as before. This reduces the intercanister spacing by 0.120 in, from 3.965 in, to 3.85 in. The total effect of the clustering configuration and of the geometrical and mechanical tolerances is shown in Table 3-3. The values of k egg for the 16-assembly and the four-assembly reduced-tolerance off-center cluster configurations are calculated respectively to be 0.9209 and 0.9241. From a comparison of the results from the four-assembly off-center cluster calculation (keff = 0.9241) and the reference case {v (kegg = 0.9136), the effect due to possible off-center loading is 0.01 ok. From an an intercavity spacing-sensitivity calculation to 3-12 Revision 3 (December 1981)
[ account for any possible bowing and tilting of the racks, the. reactivity effect is 0.002 Ak. Thus, the' total possible reactivity effect due to j dimensional and positional tolerances is 0.012 Ak. 1 Additional measures were taken to assure that a' dropped fuel assembly. in any possible configuration will remain l6 in, away.from the rack sides. Figure 3-4 shows a deflector on the side of the rack module approxi-mately 13 ft 0 in. above the SFP floor. A second identical deflector
]
was included in the design at approximately 1 ft 0 in above the floor. This second deflector will keep.the base'of an assembly at least 6 in.. from the bottom of the rack sides. In addition, a criticality analysis was performed showing that even if the assembly was assumed to be tipped.
~
sideways and somehow pushed in contact with the rack module between the deflectors, the keff would be'less than 0.95. This calculation was performed using the Monte Carlo neutron-transport KENO-IV computer code with the 16 group Knight-modified Hansen-Roach cross-section library. l The spent fuel racks were assumed to be in the actual array of 21 by.. . 31 assemblies surrounded on all sides by water and approximately.5.ft O , , w J of concrete. A reduced pitch of 13.24 in, and assembly clustering was " assumed. The result of this analysis was: ; 1 1 keff = 0.9180 0.0088 (95 percent confidence interval) This analysis, including a 0.002 Ak for temperature effects, yields a maximum keff of 0.929. Subsequent analyses show that the kegg should be 0.949 1 0.0094 (95 percent confidence interval). The following additional assumptions were included in these analyses:
- 1) The steel in the assembly grids was included in the 1
calculation model.
]
.i i
i O u l
-l 3-13 ' Revision 3 j (December 1981).
r
;r !. t I
- 2) The AMPX library and the discrete fuel pin representation l
- which had been used in other licensing applications (eg,- g
~ Prairie Island) were used in place of the Hansen-Roach l library.
The ef fect of temperature on reactivity has been found to be less than . 0.2 percent ok. The effect of boiling was evaluated by simulating the l presence of equal void content inside as well as between cavities. The I analysis results, which are plotted in Figure:3-10, show a continuous 1 decrease in reactivity as the amount of voids increase equally'inside l 1 and outside the cavity. This void-reactivity behavior will hold true ~ j as long as the void' content inside the cavity is equal to or greater than the void content between' or outside the cavities. As calculated in the thermal / hydraulic analysis, this is.always the. case for.this ^l rack design.
'l Calculations using two types of neutron spectra were compared to assure conservative PDQ-7 results from the base set of cross-section values O' generated. The conservative PDQ-7 results were used throughout the -
j l criticality analysis. l 1 The results of the criticality analysis of the proposed spent ' fuel storage racks for Trojan are summarized below: , kegg, reference case 0.914 ] Dimensional and posi-tional tolerance, ak 0.012 Temperature effect, Ak 0.002 Model uncertainty, Ak 0.008 Total 0.936 A calculated margin of 0.014 ok under the design.11mit keff 0.95 is maintained even when all the possible positive reactivity effects I listed above are added to the reference case. The SFP normally con-tains 2000 ppm of borated water; this concentration is estimated to
. decrease the total kegg from the criticality analysis (kegg = 0.936) l 3-14 Revision 3
.(December 1981) l
,s , , ,
o a.' ("N-y by 20. percent. Therefore, the normal condition is estimated to; result: t V- in a k egg of 0.75 with borated water.in the.SFP.- The'Istge1 difference in keff between'the borated and-unborated= water conditions.gives'a significant margin of safety below both the design limit (k e'gf =.0.95) and that calculated from the criticality analysis.- l
.The. sensitivity of k gg e . to stainless steel thickness," fuel enrichment and burnup was con'sidered in the design. Values for these' parameters 1
are as follows:
- 1) Stainless steel thickness of 0.1875 in. for.the Type.'304-stainless steel' cavity wall. .Under the reference condi--
tions as stated in this report, the. reactivity worth of q Type 304 stainless steel in the neighborhood..of the. nom-inal 0.1875-in.-thick container wall'is approximate 1y' O.03 percent ak/ mil .-(thickness) of stainless' stee1~. ., s
- 2) Maximum projected enrichment of 3.5 wt% U-235. No.
%. enrichment sensitivity study was specifically.per- ;
formed for the Trojan SFP racks, however, based on-information from other calculations, it is estimated that under the Trojan SFP' conditions, approximately b; l 0.6 percent ak in reactivity can be associated with:an 1 l increase or decrease of 0.1 wt% enrichment around the bnee enrichment of 3.5 wt% U-235.
- 3) Fresh fuel as a conservatism. The spent fuel'actually expected to be loaded in the spent fuel racks' willlbe approximately.30 percent. lower.in reactivity!than.the Ak/k assumed in the' analysis because of lower initial enrichment, depletion.of U-235~and' buildup of fission-products. The reactivity effect due'to buildup of plutonium in the fuel is more than compensated for by depletion of U-235 and-buildup of fission-produets.
The rod bowing that has been observed in' spent-fuel from-
'O 6
'PWRs.has been s'o slight that it would.not significantly 3-15 Revision 3-
'(December 1981)
_ m___._a ____.___ _ ma ma_____.:m.m
I affect the reactivity of the fuel. - Similarly, any () damage that might be expected to occur to the fuel g during normal operation would also not significantly increase the reactivity of the fuel. The assumption of fresh (unborated) water in the pool ensures that there need be no limit .on the minimum boron concentration in the SFP in the ]' I Trojan Technical Specifications. 3.1.2.1 Analysis Overview for Delayed Modification If the reracking was delayed until af ter refueling, it was not expected to be started until several months af ter the refueling outage. However, calculations showed that if the rerack was not started until at least 1 2 months af ter reactor shutdown for refueling, the offsite doses from the postulated accident will be less than 10 CFR 100 guidelines. The 2-month decay time was calculated using the assumptions of Regulatory Guide 1.25 as described in the Trojan FSAR Section 15.5.9.2.1 except for the following:
]
l
- 1) It was assumed that all the rods in all 65 assemblies were damaged. j
- 2) A radial peaking f actor was not included because of b; the assumption that all the assemblies were damaged. i
- 3) The spent fuel was assumed to have decayed for 30 days.
Using the above dose calculational parameters of Trojan FSAR Section 15.5 and the above assumptions, the calculated 0- to 2-hr site boundary doses were: Beta and Gamma Whole Body Dose - 2.5 rem (10 CFR 100 guidelines - 25 rem). m Thyroid Dose - 5.6 rem (10 CFR 100 guidelines - 300 rem). 3-16 Revision 3 (December 1981)
Even if no credit is assumed for the SFP Ventilation Exhaust System charcoal filter, the offsite doses are well below 10 CFR 100 guidelines. A criticality analysis using the LEOPARD code and the following assump-tions was performed to demonstrate that a minimum boron concentration in the SFP water of 2000 ppm would ensure a keff of less than 0.95:
- 1) 2.1 wt% inicial enrichment of U-235. The exposure of all assemblies was set at 8000 mwd /Mtu, which corresponds to the minimum nodal exposure expected at the end of Cycle 1.
- 2) The assemblies were assumed to be stacked in an infinite array at a minimum pitch (assembly width). ,
S
- 3) No control rods or burnable poison assemblies were E assumed to be present.
- 4) The water temperature was assumed to be 120*F.
- 5) A boron concentration of 2000 ppm was assumed.
Using these assumptions, a keff of 0.87 was calculated. Therefore, should reracking be delayed until af ter refueling, PGE proposed that the minimum boron concentration be established during the reracking evolution at 2000 ppm and that the reracking not begin until at least 2 months af ter the reactor shutdown for refueling. 3.1.3 SEISMIC ANALYSIS Seismic analysis and design is in accordance with the criteria speci- e w-fled in NRC Standard Review Plan, Section 3.8.4. Horizontal and vertical
- seismic response spectra are described in Trojan FSAR Section 3.7.1.1.
Specifically, for structural design of the racks, the following absolute 3 sum load combinations and factored allowable stresses were used: 3-17 Revision 3 (December 1981)
, ' D + L + E '+ 1 j S i
-l s
D + L.+ E' + I +:Ta f.1.5S'. D + L + . FD < 1.5S - l l where . 1 { D= stresses.resulting from dead load'of' rack structure' " 4 L= stresses resulting from live load of fuel assemblies . t< l E= stresses resulting from the OBE [h :{ E' = stresses resulting from a SSE i
~
Ta = stresses resulting from thermal' loads at. accident condition I= stresses resulting from fuel bundle " rattling"' impact in cavity 1 i FD = stresses due to. equivalent static loads resulting from ' fuel. drop of.the module S = allowable normal operating stress limits as' delineated in ASME Section III, Appendix XVII. Each of the four configurations of storage rack modules were 'modeled.andl y analyzed for static and seismic loads ~as described-above.using'the ' ' ~ SAP IV( } finite element computer program. It was assumed for the I mathematical model~that the fuel assemblies add' mass but no stiffness ,, : 1 to the general structure. The surrounding externals and entrapped . water ' E3 j was conservatively calculated to be equal to approximately 30 percent , of the combined. dry mass of the fuel and module. l^ l' D 'b 3-18 Revision.3' L(December:1981)? s
i
) ,m Seismic modal and spatial responses were combined in accordance with
(')
)
NRC Regulatory Guide 1.92, Revision 1. The first'eight modal frequencies \ were determined, which included the first and second modes in each of the three spatial coordinate directions. The analysis and results are further described in Appendices B and C. g. a 3.1.4 SFP LINER PLATE AND FUEL CLAD INTEGRITY i The SFP is a' reinforced concrete structure with a team-welded stainless i steel liner plate. The liner plate is 0.25-in.-thick Type 304 stainless 1 steel. The liner plate specification requires all welds to be smooth and allows grinding to produce smooth and/or clean finished weld surfaces... :)
'd Also, Section VIII of the ASME Code permits grinding of these welds.
Therefore, the grinding flush of either plug or seam welds is consistent j with the vendor drawings, the liner plate specifications, and'the ASME Code. I a Written instructions do not impose special means to control grinding heat input to affected zones because PGE's metallurgical engineer's ] evaluatiun concluded that the degree of microstructural sensitization of the base metal where weld grinding is performed is not increased by heat -l generated during the grinding operation. In comparison with initial g, welding, heat input by grinding is not considered to be significant. 'I Similarly, residual stresses that ~could be induced due to grinding, j within the bounds of normal good workmanship, are not considered to be of sufficient significance to warrant special concern. As on all Q-listed work, supervised experienced workmen were utilized to accomplish the work, and quality control was performed by qualified inspectors. The welds were specified to be ground flush. Visual and liquid penetrant inspections were performed af ter the grinding, and all-areas of the liner plate where grinding was performed were inspected by the liquid penetrant method. Any unacceptable conditions (overgrinding, gouge, etc) was rejected and repairs performed and requalified before acceptance by Quality Control inspectors. TO
\.j 3-19 Revision 3 (December 1981)
.)
i
- c. A drainage system is provided beneath the liner plate to collect leak' age
-(. / if it'should occur. The drainage system is designed so' that the loca- -
tion and magnitude of a leak can be identified and repaired. j Small amounts of leakage can be identified visually by. checking for evi- i, 4 dence of water at the drainage system collection manifolds. Routine m w shift and tour inspections of the drainage system assures early detection of leaks. l The SFP water chemistry will be sampled and maintained in accordance I with Tables 3-6 and 3-7. The SFP water chemistry limits and sampling frequency are specified by Westinghouse and are incorporated into the .j plant operating procedures. Therefore, from a materials viewpoint, the environment to which the SFP liner or spent fuel is exposed is relatively benign and not conducive to corrosion. For example, Berry (4) W s ~ I corrosion rates of approximately 0.02 mils /yr in 500*F water for zircaloy and 0.17 mils /yr in 600*F water with 1600 ppm H3B04 for Type 304 stainless steel. Berry further stJates that zircaloy is not affected by (u.) H3 B04 in the range of 1500 ppm. 1 hut, given a fuel clad thickness of 22.5 mils, a liner plate thickness of 250 mils and the corrosion rates G above, over 1100 and 1400 yr, respectively, would be required to pene-trate the fuel clad and liner plate. The temperature increase of 15*F (125'F to.140*F) as a result of the rack modification is not of long duration (Section 4.2) and is well within the normal operating temperature limits for stainless steel and zircah y-Based on the above, it is concluded that neither the SFP liner nor spent fuel cle.d integrity should be af fected significantly by the rack modification. O Q! 3-20 Revision 3 (December 1981) w _!_ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ -_ _ _ - _ _ - _ _ .
)
)
l i 7- x 3.2 SPENT FUEL POOL COOLING AND DEMINERALIZED SYSTEM h (,,! 3.2.1 DESIGN BASES The SFP Cooling and Demineralized System ic 6esigned to remove the decay heat from spent fuel in the SFP and to continuously purify the system k3 i water inventory. The SFP Cooling and Demineralized System is designed to-perform the functions described below.
- 1) Maintain the SFP borated water below 140*F by removing the decay heat from seven regions of 33,000-mwd / tonne-U stored af ter each of seven annual refueling operations.
The heat load under these circumstances will not exceed 1.89 x 107 Btu /hr.
- 2) Maintain the SFP borated water below 140*F by using the Residual Heat Removal (RRR) System in the event that seven regions with a burnup of 33,000-mwd / tonne-U are 7-~
( ,) stored for 1, 2, 3, 4, 5, 6 and 7 yr, respectively, and a complete core is unloaded 150 hr after shutdown. The heat load under these circumstances will not exceed 4.12 x 107 Btu /hr. l l
- 3) Maintain the clarity and purity of the borated water and control potential fisson product releases and other con-taminants. Even after appropriate scaling of present FSAR values, the resulting pool activity does not lead to appreciably larger releases than the.previously estimated liquid or gaseous releases from the Plant. The SFP puri-fication subsystem can be used during the time an RHR train is being used for SFP cooling. A core flow path also exists using the normal letdown to the Chemical and Volume Control System (CVCS) demineralizers to provide additional purification.
f% () 3-21 Revision 3 (December 1981)
1 l l i g 4) Maintain the clarity and purity of borated water in the () refueling water storage tank. #
- 5) Supply makeup (normal and emergency) water to the SFP.
The SFP, refueling canal, the emergency makeup supply line, and the l interconnecting lines to the RHR System are designed to Seismic Cate-gory I requirements. The SFP Cooling and Demineralized System has been designed to Seismic Category II requirements. I 3.2.2 DESIGN DESCRIPTION ) l l No equipment modifications are required for the SFP Cooling and Deminer-l alizer System; however, the SFP Cooling and Demineralized System opera-tion under normal and special conditions is changed. During normal l operat. ion, with one to seven regions stored in the SFP, both SFP cooling 1 pumps will be operated initially to maintain the SFP water temperature 1 below 140*F. The RHR System and SFP Cooling and Demineralized System n ( components are not adversely affected by increasing the SFP temperature j design limit from 125'F to 140*F. The refueling cavity and Reactor Coolant System (RCS) water temperatures are limited to 140*F during j i refueling operations in accordance with the Trojan Technical Specifica- l tions. The proposed changes will make the SFP temperature design limit l I compatible with that of the refueling cavity and RCS. When normal cooling of a full pool is considered, the individual assem-blies within the pool are cooled by natural circulation. For worst-case a i design conditions, the cooling water enters the pool at 112*F and is O i mixed with 140*F water at the sides and below the racks. This provides- j i an inlet temperature to the individual channels of 130*F. For this con- 1 dition, analyses of the SFP Cooling and Demineralized System show that the average SFP water temperature will be maintained below 140*F for the design bases listed in Section 3.2.1. e l
, For a maximum power fuel assembly, conservatively assumed to be unloaded f s V) e at the Technical Specification limit of 100 hr following reactor shutdown, <
1
.]
3-22
'l Revision 3 l
-(December 1981) !
I g-( the maximum clad temperature in the fuel assembly and the assembly b d coolant outlet temperature are calculated using the HYDROPOOL computer ] code to be respectively 183*F and 154*F. IlYDROPOOL is an NAI-derived computer code from Control Data Corporation computer code HYDROP, CDC-84004500. This analysis assumed that 100 percent of the decay heat was generated.within the fuel assembly. Under these conservative condi-tions, no voiding was calculated to occur in a cavity containing a j l maximum power fuel assembly. 1 i Natural circulation of pool water between the fuel racks removes the . 'f decay heat that is deposited between the racks. Slots at the top of.the ! racks permit natural' circulation between the racks. The decay heat generated between the racks was calculated to be 6.5 percent of the total j decay heat; 3.5 percent of the total was calculated to be generated in the stainless steel channel. To determine the natural circulation con-ditions between the racks, a total of 10 percent of the decay heat 4 generated was assumed to be deposited in the space between the racks. Based on this conservative assumption, the maximum coolant temperature f between the racks was calculated to be 146'F for the regions adjacent to ! Qi maximum power assemblies. Under these conditions no voiding will occur 1 in this space. I l The vapor produced from the SFP could cause condensation in the SFP exhaust system that could adversely affect the charcoal filtration. To b maintain the efficiency of the charcoal filter, a four-stage 90-kW heater is located in the duct ahead of the charcoal filter. The relative humidity of the air entering the charcoal filter is continuously n.oni- - N tored. The heater operates as necessary to maintain the relative humid-ity below 70 percent. The design conditions for this heater are as follows: l l 1) Temperature of air entering SFP ventilation exhaust - 145'F. l , 2) Relative humidity of air entering SFP ventilation y exhaust - 100 percent. l 3-23 Revision.3 1 l (December 1981) .
i l p 3) Temperature of air entering charcoal filter - 155*F. . O _ 6 1
)
- 4) Relative humidity of air entering charcoal filter - i
<70 percent.
3.2.3 SYSTEM DESCRIPTION AND OPERATING MODES f The SFP Cooling and Demineralized System consists of the following components: 1 l i
- 1) Two half-capacity cooling pumps (P-207A and B). l l
1
- 2) Two half-capacity heat exchangers (E-205A and B).
- 3) One purification pump (P-208). ]
- 4) One purification filter (F-201).
O
\ / 5) One demineralized (T-224). ;
- 6) One demineralized after filter (F-211). I i
i
- 7) One skimmer pump (P-209). ;
i
- 8) One skimmer filter (F-203). !
l i
- 9) Valves and piping.
i
- 10) Instrumentation.
FSAR Tables 9.1-1 through 9.1-9 give physical data for the heat exchangers, ! I purification pump and skimmer pump, purification filter, demineralized !
.I after-filter and skimmer filter, the demineralized, and the SFP. FSAR Figure 9.1-4 shows the piping and equipment for the SFP Cooling and g m Demineralized System. I
/ h -l V l 3-24 Revision 3 (December 1981)
.j
-_- _ ._ A
The SFP Cooling and Demineralized System is a. closed-loop system consist-ing of three subsystems - cooling, purification, and skimmer. 3 The cooling subsystem utilizes two half-capacity cooling pumps and two l half-capacity heat exchangers.. The. cooling pumps draw suction from the SFP and discharge back to the SFP through the. heat exchangers. j The suction piping from the SFP to'the cooling pumps is connected to the-RHR System through a normally closed 10-in.-diameter line. One 8-in.- j diameter return line from the RilR System ties into the 10-in.-diameter header returning the discharge from the heat exchangers to the SFP. ) l 1 The purification subsystem utilizes the purification pump to divert j 250 gpm of the total flow through the purification' filter and/or the' demineralized. Both the filter and the demineralized have suf ficient capacity to recirculate the entire SFP volume (390,000 gal) approximately l l once daily. One 4-in.-diameter line from the refueling water storage 1
)
tank (RWST) is connected to the suction sides of both the purification j G O 1 and cooling pumps. This feature also enables the purification system to. recirculate and purify the RWST water.
- I The SFP purification subsystem will be operated continuously during f
refueling operations and intermittently thereaf ter as required to main-tain SFP water clarity and purity. When it is not being used to purify .l
?
SFP, the SFP purification subsystem can be used to purify kWST. -l l i 1 The operation of the SFP cooling pumps is indicated locally. Plant j operating procedures require that these pumps be inspected once a shift for proper operation. This surveillance includes inspection for abnormalities such as excessive vibration, and improper suction and j discharge pressures. Plant operating procedures also require that both . SFP cooling pumps and externally-actuated valves be tested quarterly to DI verify adequate system performance.- This testing includes running the SFP cooling pumps on normal recirculation flow, exercising the valves with the SFP purification pump idle, and recording system performance 1
/"% 1 V) characteristics such as pump suction and ~ discharge pressure. -s 3-25 Revision 3 (December 1981) -j
.]
p Modification of the SFP racks increases only the storage capacity of the SFP and not the refueling frequency or-the: amount of spent fuel moved during the refueling. Thus, the amount of corrosion products ' 1 which are introduced into the SFP water primarily during. refueling 3 operations will remain at about the same level regardless of the SFP storage capacity. Increasing the SFP capacity may, however, increase ; the amount of fission products introduced into the SFP water. However, j
. i the evaluation in Section 5.1.4 shows that the purification system is capable of accommodating the potential increase in fission products even if a leaking fuel assembly is transferred to the SFP. Therefore, the .
i SFP purification system is adequate for the modified rack design. ; i 1he need for filter cartridge-and demineralized resin replacement is C j based on the differential pressure across the filter cartridge and the decontamination factor of the demineralized resin. Replacement of I i the filter cartridge and demineralized resin will result in approxi- i mately 2 ft 3 and 50 ft3 of solid waste, respectively. It is anticipated: that filter cart. ridge and demineralized resin replacement frequency will j O \g not increase significantly above the expected annual replacement rate as i a result of the rack modification. i l The purification subsystem can be used with the RHR System to maintain
'l q
SFP water purity. The purification subsystem, including the demineral- 1
)
i izer resin, can accommodate a maximum water temperature of 140*F without - degraded performance. The skimmer subsystem utilizes the skimmer pump to draw suction from the surf ace of the SFP, fuel transf er canal, and refueling. cavity. The flow is then discharged through the filter back to the SFP. Borated makeup water to the SFP is provided by the holdup tank recirculation pump. Demineralized makeup water is provided by the demineralized water trans-fer pumps. Emergency makeup water supply is provided by the service water pumps. Overflows f rom the SFP and the fuel transfer canal are directed to the () clean waste receiver tank. Drain and overflow connections have also been provided for the refueling cavity. 3-26 Revision 3 (December 1981)
1 i
)
-J
._ Provisions have been made' to pump.a portion of the'SFP flow to the'CVCS.
bdidup tanks to maintain the' required boric acid concentration. 4. p Ventilation is, provided by a once-through-ventilation system described in FSAR Section 9.4.2. The discharge is. monitored for radiation by.the <[ ( Process and. Effluent Radiation lionitoring System, discussed in FSAR- .] 5, 4 Section 11.4. During normal operation,1/3 'of ti?e reactor' core fuel elements are . stored in the SFP. Initially,'both cooling pumps are operated-to main-' l 1
' tain the SFP temperature at or below 140'F.
~
When the decay heat emitted; -! by the spent fuel decreases, one cooling pumplis stopped, and the remain-- ing' pump is sufficient to provide cooling water. After one. cooling pump' . is stopped, one of the two heat exchangers may be taken out of service for maintenance purposes. ' A portion of the SFP flow is pumped by, the purification pump through.the r
]
y purification filter and/or the demineralized for removal of impurities', .1
) fission products, and other contaminants present in the SFP.-
The cooled and partially purified water is discharged.~1nto the SFP.- r 1 through a diffusion header located near the bottom. Boric acid concen- , I tration in the SFP is maintained at minimum 2000 ppm boron. Provisions have been made for supplying borated water from the CVCS holdup tanks and demineralized water from the demineralized' water storage tank, in con-junction with pumping the required. volume of water to one CVCS holdup; tank. i The skimmer pump operates intermittently to. prevent dust and debris from accumulating on the surface by pumping a small amount of SFP volume from- I near the surf ace through -the filter and returning it to the SFP. Provi-sions are made to filter and purify the content of-the refueling water storage tank during normal operations by isolating the purification pump ~ from the main cooling loop and opening ~two automatic valves in'th, line connecting the suction of the purification pump and the refuelin; Jater-storage' tank. 3-27 . Revision 3
,(December 1981)$
y
' During cold sh'utdown and refueling operation, the refueling . cavity, 1
. refueling canal, fuel transfer tube, and the SFP 9re kept full ~of' borated-
- water from the refueling water storage tank. utilizing the residual heat 1 removal pumps.. 1 The cooling subsystem. lineup procedure during the' refueling' operation'is: J similar. to that durb.3 normal operation. ..The refueling cavitycwater can i be cooled-by circulating it through the SFP Cooling'and Demineralized System heat exchangers. The ref ueling . cavity water can also be . filtered" 1 and purified by circulating it through- the ' purification- filter and j 1 demineralized.
]
Air circulation velocity in the SFP area is kept sufficiently low so as not to generate ripples on the surface that would inhibit clear ~ vision from the Fuel Building crane and transfer platform operator into the SFP. ]l Af ter completion of the refueling operation, water from the refueling cavity and refueling canal is pumped into ~ the refueling water. storage tank. Special operating conditions are described in Section 4.2 of this A report. -j
)
3.2.4 INSTRUMENTATION 1 1 A level switch is provided in the SFP to transmit high and low water level { annunciation signals to the control room. This continuous monitoring -will , ensure rapid detection and appropriate corrective' action for significant leakage. SFP water temperature-sensing element, transmitter, indicator, and control room high water temperature annunciator have' also been provided. Thermowells are provided in the piping upstream and downstream of the heat exchangers to permit measurement of temperatures when required. Di f f e ren-tial pressure indicating switches are provided across the- filters to moni-tor their effectiveness. Pressure indicators are provided upstream and downstream of both cooling pumps for monitoring satisfactory pump operation. O U-a 3-28 Revision 3 .. (December 1981)'
7
- e !: c i
[ }- .i. : _ i r i.. i
? Automatic air-operated Seismic . Category 'I control valves (located. on' thef
.p)
.g . suction line from.the'RWST) are designed to;.faillin.a closed position and~'.
.y
~
also to close on receipt;of'a safetyninjection' signal,.thereby isolating 5-l
'the Seismic Category II SFP Cooling-and Demineralized. System from' the!
' Seismic' Category I; RWST. !
n
- The SFP ccoling pump characteristics and LRWST isolation valves are : tested: ]
every 90 days. . The ' position of the Containment iso 1Ation' val'ves is ; ~ w che'cked monthly either by visual verification. or by checking .the controli room locked valve list. , , d
-s
'l H
1 1 I i - . :i
?
-t ' .1
{ .I t i I
.1 i
r !
,Q. ,
?!
. (f t
3-29 . Revis' ion'3 s . , (December'1981)-
- 7
.i
9 TABLE 3-1 BEGINNING OF LIFE CRITICAL CONFIGURATIONS FOR PWR'S Calculation Measurement Utility Condition (ppm) (ppu) FP&L HZP 1114 1173 SMUD HZP 1577 1552 NSP HZP 1505 1519 DUKE HZP 1458 1476 WPS HZP 1563 1576 PGE HZP 1324 1316 0 1 i O ,,_,,,, , . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ - - = _ - - - - - - '
,a 1
j
-a p:
A' TABLE 3 _2_
.1
.4 PHYSICAL PARAMETERS OF .
REFERENCE CALCULATIONS U-235 enrichment '3.5 w/o ' UO density 2 95% theoretical 'i Fuel active length 144 in. Fuel rod clad Zircaloy "4 Grid material Inconel Sleeve material Stainless steel Cavity center-to-center' distance 13.300 in. Canister material -304 stainless steel I Canister wall thickness -3/16 in. *
, [
Pool water Clean and nonborated Control material or poison None O
& Q 7, (!(l Fall a LODd) l q
I s l l.) -- q
.___--____w
a l
,j j ,-~ .
O i.' l TABLE.3 )
]
RESULTS FROM THE EVALUATION OF TWO OFF-CENTER CLUSTER LOADING C0 FIGURATIONS i Cavity Center- , to-Center ) Distance g l Case Description (in.) eff i i 16 off-center fuel assembly cluster, no 13.300 0.9209 , spatial tolerances included, all dimen- l l sions nominal except fuel assemblies off- 'l centered ' l l . . ! 4 off-center fuel assembly cluster, 13.240 0.9241 j including spatial tolerances q
.i k
i i >
'w) 1 I
e N
- - -- -----..-.A
7 5 4 6 9 8 0 0 1 0 0
'1 Us t e 4 q -
E 3 3 0 3 1 nz y 8 5 2 7 0 3 0 obb 2 0 2 0 0 0 0 iiF t c a r a 3 5 9 5 e F 1 2 5 5 2 0 0 8 t / 5 5 2 5 0 0 1 0 n a I f
}/'
g n 0 0 0 0 0 0 0 0 f i z z 0 0 0 0 0 0 0 0 e d - b 8 8 8 8 8 8 8 8 l nzF 6 6 6 6 6 6 6 6 1 1 1 1 1 b s e 1 1 1 asB we or l t l Sl 0 0 0 0 0 0 0 0 A ax _ 0 0 3 0 0 3 0 0 1 1 0 0 1 1 i - a 1 1 5 9 5 9 5 9 5 9 A xxl 1 1 1 - 1
@j g
[s n 8 7 1 1 0 0 2 7 si z a ed - b 9 0 2 6 7 6 3 mr nzi 6 9 7 1 5 ut e 4 3 1 mSB i x n ag ~ Mil 0 s ax 8 6 7 3 2 6 3 ei- a 1 7 4 8 4 5 6 8 D xxf 7 8 7 0 3 6 7 S A 7 4 3 5 1 E S S E r R / 5 8 5 8 5 0 5 C T s 1 3 0 3 0 3 1 3 1 4 S e K 1 1 1 1
- i 3 E t M r E A e ns p ou L R B F oil z 6 6 4 6 4 6
.Y. A r t u - 5 3 5 3 9 1 9 1 T C P cd z 5 3 5 3 3 4 3 4 N eo
+
I nSM
/ T o A i C t h c a 3 1 3 1 9 5 9 5 t e e 7 7 1 2 1 2 S r A 1 1 1 1 1 1 6
nB) . r o 7 eis - 7 7 t eB 2 5 2 5 4 3 0 9
- b. a r m
e c ud
/ 2 6
/ 2 6
1 1 1 M ogn 2 2 I i a F ( e e e e e e l l l lg l g g g l g g n 3 n n n n n a a a a a a n g n g n. n :g n g n u. o n i n i i n i n i l i i - i - - i - i - - t d 8 d 8 8 d 4 d 4 4 p a / a / / a / a / / i o 3 o3 3 o1 o 1 1 r c s s s L c x r l c x r
- n o
x r s s l c x r L c x r
- n x -
o r _ e u i - a i - a i - a u i - a i - a i - a D r m 2 b m 2 b t 2 b r m 2 b m 2 b t 2 b T s / s / a / T s / a / n i 1 . i 1 n 1 i s / .1 l. 1 . n 1 . o e - n e - n. r e - n r bm 2 tm e - n i - n i - i o S 2 i S 2 i o S 2 i S 2 i b 2 i _ t o - - - o - - m - c l n x 2 n -x 2 o x 2 1 n x 2 n x 2 o x 2 - e F o o C F o o C _ S i - x i - x - x i - x i - x - x _ e t 2 t 2 S 2 e t 2 t 2 S 2 - l c / - c / - S / - l c / - c / - S / - _ u e 1 2 e 1 2 R 1 2 u e 1 8 e 1 8 R 1 8 r - / r - / S - / d r - / r - / S - / . WD 2 i 2 X 1 i 2 D Y 1 Y X 2 1 M 1 o i 2 5 D X i 2 5 D Y Y X 2 5 _O E$a oo u ~ 9Sc*, G%-
.. I
s TABLE 3-5 EMBEDMENT LOADS [a]
.(
P / Stud P / Stud Result Load Case Node *(1b)' 7 (1b). (1b)
~
i ( 2-Module Floor Truss 2 14088 1680 141[88 X-Direction Seismic Loading 11 16476 1680 16560 14 8658 2262 8948 23 16338. 2262 16494 2-Module Floor Truss 2 0 14760 14760 Y-Direction Seismic Loading 11 0 14760 14760 14 0 12960 12960 ! 23 0 12960 12960 I i 1-Module Floor Truss 1 14820 2317 15000 X-Direction Seismic Loading 2 12564 1818 12696 17 12092 '9668 12472 18 15228 9165 17773 ) i 1-Module Floor Truss 1 0 18000 18000 i Y-Direction Seismic Loading 2 0 6557 6557 ! 17 0 17844 17844 18 0 12312 12312 X-Direction Y-Direction Overall f Load Case Node Result Result Result 2-Module Floor Truss 2 14188 14760 20473 b ' X-Y SRSS Combination 11 16560 14760 22183 14 8948 12960 15749 23 16494 12960 20976 1-Module Floor Truss 1 15000 18000 23430 X-Y SRSS Combination 2 12696 6557 L14289 l 17 12472 17844 21701 18 17773 12312 21620 [a] The existing embedments are anchored to the SFP concrete base slab with welded headed studs. Each embedment has a code basic allow-able shear capacity of 62,400 lb. Revision 2 (August 1977) ( V) 1. L___________._._.1____ _ _ _ _ _ _
')
a
.I<
l TABLE 3 1 f]
.[] . SFP CHEMISTRY SPECIFICATIONS
_. =j Analysis Value Remarks 'i Gross Gamma NA Determine purification. H requirements, evaluate leakage -) from spent fuel; analytical ! sensitivity 5 x 10-7 pCi/ml. i Gross Beta NA Monitor activity level. Tritium NA' Evaluate in-Plant buildup'of l tritium; analytical ^sensi-' i ' tivity.10-5 uC1/ml .' l pH 4.0 - 4.7 Determined by' concentration of boric acid present. Boron >2000 ppm B . Prevent positive moderator coefficient.- Conductivity 1-40 umho/cm Consistent with pH. , Chloride <0.15 ppm Prevent' chloride stress g , i corrosion. 3 J i Fluoride f0.15 ppm Prevent fluoride stress
-corrosion. i Calcium fl.0 ppa Prevent deposition.
Magnesium fl.0 ppm Prevent deposition. Suspended Solids f1.0. ppm Reduce' deposition.' $ Sodium fl.0 ppm Control soluble impurities.. DEMIPERALIZER INFLUENT: ; Gross Gamma NA Evaluate performance of demineralized. Analytical sensitivity 10-5 uC1/ml. i DEMINERALIZED EFFLUENT: , i l Gross Gamma NA Evaluate performance of l demineralized, AnalyticalJ t sensitivity 10-5 uci/ml. ; L Co-60 DF >25 Determine resin replacement time.
/,g^T o
i Revision 3 (December .'1981)- 7
7_-_. , i l . 'l { 5 1 y, , :. Y' t
, y ,, 4, 0 .;t; y , }o
- q, 1 , .
>1
'!. 3
.; j
< TABLE' 5- 7. -
ft.. w- . .
,j .%
)
SFP SAMPLING: SCHEDULE- .g g
;l Analysis- . Frequency , ]
A. Gross Gamma Monthly u .
~
$ t .
.. Gross Beta. ~ Quarterly ,
-i-d1
.l Tritium . Monthly l
.pH Weekly'
)
.i 22 Boron Weekl'[a]'
y :j c] Conductivity Weekly. 1 q l Chloride Weekly , Fluoride ' Monthly Calcium. 'AR[b[. , y Magnesium AR[b) , G 'l
, Suspended Solids- Monthly
\
\ s s !
Sodium Weekly qJ
\
DEMINERALIZED INFLUENT: ~;
, "l
- Gross Gamma . Monthly 'i i
DEMINERALIZED EFFLUENT: Gross Gamma Monthly' I Co-60 DF Monthly ; [a) In Mode.6, t maximum time between , samples -is 72 hr. , [b] As requested; these analyses would normally lbe requested . as an explor-atory. measure:following,other out - of-sp'eci fica t ion condit ions ,' eg', . L high sodium.or,high-suspended-solids. ,
- i l
- .(
~
. Revision?.3,. 9 .
.(December,1981)- 1
- t. 1
- '( ' '
.w- my . x n :. ,
- h. .
, .},
. l .' l, '
l1 ' --
% ,. r s M'l.
i ., I o i 3 i s t 1 t *I
- m '
. i
- 5 , 1 a _. ,e -
t, s . J g*% ..1 g .2 : nd" t e. 4 bg c J.,e 5* t-4
?g - %
- 1 33 :q
- g. ,.
. .g 1
EMHUFHFi l + + + m ul m 1 M W41 '- B W 7 . . , .. , Fa ss . d.E , iT.,..
. ,j . , [
r ..',j.... 3, , . M$p p E 3 amg GE : 7 . . ,; . . [
~
. . .:. 4 . Gd.g gE g..
L*E B I:aunteurad@!t RMM M T-i r.m w p gr~ a rry;nm).q.n. l' ! ig g j f-METEIEIEM - + E +'+g- *. 7 : n i 9fMEllBEUR) * +
+
h+ + +$g 4 i - +D4 ' ll 1 (V) , s
?
...s.
.,.7.p
- p. ....5..
. 'p y 2
j
'a
? $+ + fk.licauiL cMl s@bauteruel-d g $. $m 1 7..;...y@ rmTamsma R g@[nmUrar.- sq- ..,
e 'l
?....dy -$
- 7. 4 g, T j
. 4 . . . . .-. .
. . .g.... . . 3. . .- ,y r ,
d-S1ETHEIEf5 + '* *2+ + + i + +5+ -
', [ y y -
y-gungag@g . . .J . . . ..,g.... , pmg t... 3d. . . . ... . ..... .
. 2 L.. 3 e he1ou t u -M _Ec um aem -1 L
- 4. .
7.
. gigyirja
.o. . . . . . .
wyFj u j.yr.ifR rqsg
. . . . . . . i. {F
.4 n, d G '1 3 $ w f. a '
s...
. . a
. y . . . . . . ... '[% ; g y q...
- n. . . . . . . . :. . [. y :
! )-5BElnY .a.'a. +. + : + i s. '
(E - J'
;;;D37;'r.-- T }-@LTLEUTE &s ! she w un1 aunt. .lve t.*ith %-f.
5... 35n r. t u m r. mqyrmaini g-g r. 7 . .g . . g . . . . . ... .
,q /'
7
! ,- / 1 . . . p . . . . .,
- 9. . [. ',
jp i y x y . ,;c m
. . . s . . ..
. . .. .. .- g ;
n
~, g . . . . .- .: .2. .
E ; w RMHEHHrdh L.g.m .e e M aM .n o m.ii!. hs
.9
]
?
.I V , w
_ p- _-_._ . ; u m
; D .. . - g
.GE e? 9 - St .J,i c . L.
t l _ _ _ _ _ _ _ _ ~. _____---_L-_____. .-__.---- i- L -
~
O w
- c. . _ <}
=,
u _ c.
- /' - . : -
I
.~, ,
e M he na Nay, A L k 3 ":,. . o e sr t n e m 8 ,, e M l o g _ s f n n-9 o. U H a a I D a r m ml : WWI D.$-\ :. i
, r r
E 5 E. L s A n..b N,
. l l
e "5 - m 4 a r s. v i o s F g f o. n E ~ 4 i bn a nD f 9
'-, ~
u 8 u,/ yAb B. N l e t a c o l o
~ .a x
- l L
e ez.\ x.,
. u d _
T - ;4 o _ o M _
~
n s o
, t g n i
l _*F ~,, K D fv;' f OD B K, 3 e d e m i J w i
"8 1' 4 - e 6 - b Z
~n r a l o5 E
2 m :
. n o -
f$ , t 3 oNgu,.ly/ p. pC' D 9 C 'e f f o '3
. ,I!
w c r u g F E R 1 _
'6 F c
~4
& _ 4< o io
;0 5
i . c.i
- \ y% iAi .
{ e
,sI
- fu t y ,.k $ 30 OE, 1
q - l i.
*4 f
'w.
- r. -
a, l- .
- I s,
J C
'O w
a ; o i_.
- - O' a"
[ , O a W n' i
~,
1 f m i-m - 1_. 2 i_ d - il!! l LLIL f,
I (*
+
3"' 5 00' n /
' 3 l
-3'd
- 2. caoss pieces (3/@ ARE , 375 ST K . X 2" d (tocATe As saow 4
- i. wate s @ p @ ARE 2k"N2k"xk"
-~
._ s :
-7 7" L REF, =
- 7' O" "
=
_ g_ g " '3 m i 31 ,
/ O ,V (~rYP./o PLACES 2 -
4 / <
) / (TYP 4 PLACES
/
7 99M _M N I, \ Q c 2.c.co d OUT5IDE FLANGE 6
.., u. i n
d
.5'O T Y P, j hd I 2 11 6.5 d LG' TI !
APERTURE 2 '-3"d LG, CARD o-Y -
- Also Avallnble On l Aperture Card l s
Figure 3-2a Locating Frame Details l Sheet 1 of 9 Revision 1 (April 1977) _ _ -_ - T~~ L
I l
-n 2'-10;
$ (
o g e 50
, /. .
A c' TYP.(4)) k/ . 5
- 2. CRO SS ' PIEC'E5 @ d 6 ARE.37 5 STK.
6.ocArc As saoww
- i. ANGtes @ mo @ Aae z ye x zye x 3
' ' ~ ---_ _ _ -_ - -
P 7 s hT-7, REF.
-- .)
k k-O ' d
= ,
-d J
y-g e s- } 5"d LG. 3 '-6," d LG, k2-) 5 la I
\ . .,w ,
g,
/-
nd l 3. o 4 b ' 1
. i < .
N : (N e f 5,00 A c l TI 2LYL /G PLACES APERTURE i N CARD
'o l 0' " d j I
Also Available on .' Aperture cara
~
c3 ; L 2 Z2 0 4/36-0Z. K *[# Figure 3-2b Locating Frame Details
'I 1 4 's f
4
- Sheet 2 of 9 q Revision 1 i (April 1977) j
N. A W + 4 . I d ' ouTSIDE FLANGE 2.0.00 1
.E0' TYP. -
I M 2., CRO SS PIECE S /6 ARE , $75 STK. K 2.' ( LOCATE As SRowN
- 1. ANGtts @ (@ An 2 h'x t yt x
i 6 '- 3 "
- w l L8" 9-~- l
_ na 2'-3"gLG, l'-6.d' L6, (2) ; 2 6 3 5 , 5.CD' \ i i o f j f ,fe .
~
PLACES (WP. 4 _3a e e, / TYP. (, PLACES
- 4'-9 v g A
'f- O" ' =
TI 7'-Y"REF' " APERTURE CARD Also AvaHable On APerture Card
% e 222643(o-o 3 )
Figure 3-2c Locating Frame Details - [ Sheet ~3 of 9 [ Revision 1 d (April 1977). ]
- (
. . . .. -f
~ ~
(
=
id-Z.0"REF. pt '. o" ' ( e
= 8'-9"'
= 5'-3"' =
2.
< (7,75 " =
' 1Q TYP. to PLACE 5 g_gh i ge 7 .
- ~g e 9 7f f f
)
yh. i t
'I -
l . _ z'e 6 (zf> 7
- = +
_. 5'-9. 7 5 " - , 10{'sTK.xt'dx2
- 3. LOCATE h MIDWAY BETWEEN! @(@ ik"
- 2. LO CATE CR.OSS PIECE A5 SHOWN
- r. mms @ @ @ @y@ mEzozr w I
WN senEmp
. _ _ _ _ _ _ - - - _ - - - - - - - - - - - - - -a -
=
ece -
~ 2'-0% +3'*
4. 3 7 ]
- N K ,. h, ~[I 'N q.co# oursiot nAnar we.
9 ,
- so d TYP.'
(, ' - 3 3"*
-ide -t = 4 "d
LE. (4 @ "srx. x 2"d x isas"'te.(z) n j TYP, APERTURE C S 14 PLACES CARD Also Available On Aperture Car i SUzz2b436 -o V Figure 3-2d Locating Frame Details 1 Sheet 4 of 9 ' Revision 1 (April 1977) _ _ - _ - - _ _ _ - - - _ - _ - - - - . i
~w--- v--__.-.,._,,
[
- --- .. - l g.g , ; g " g 4 . _ _ _. - - - .. a. $ ' ,0) , .
+
5 '- 3" ' I TYP. 2 TYP. 8 Pt. ACES) Q TYP.14 ) h, ' r ~3\ Tr K1 A > .
%;[/_[ jg ^
J LZl 039Ji o.25 REF. \. _
/
u i
/
j i
/ -
/ .
l-4.00 * - I ~~ ~ t5
.e..
(1) - 2yo a
^
- .... s >. g >' S .. . . .
t g 3n .. a
- 2. CROSS Pi,ECES i(7k8]' ARE 73 STK. K2 (LOCATK AS SHOWN )
- i. .
ANGLES @ AND /[ ARE 7.'/t 42Iz"XI4-
~
./
=
Lr. .
.__.. .. g '. o " ' :
m I'-7dLG [1 ~
,,, ,2'-6'/LG.(4)-
4 l i (e-~)6(,8). 7 3 i 3- s , - j p' 5 y GG Vi \ / k ee l'-8.25 si d OU T.61DE 1:LAN6C _ - 1
- g. .- . _ _ _
\\L ., ,
'r TYP.
i ! L i I (M !
.50 TYP. + 3.00 d(2)
TI o - -3 L,9.75 ,, e * - APERTURE a -- __ c '.lo'*(gh---- . . u CARD Also Avaihble On Aperture Card E , 2 2 3 6 G lo ~ O S Figure 3-2e Locating Frame Details Sheet 5 of 9 Revision 1 (April 1977)
~ -. ,
c
= ---
16'2.2 5 "1 b) , ._ . . _ . . .. 1 S'-V ? 5 '- 3" ' 3 h TYP 144-; j- ~' TYP.S.j 9 'F A TYP.8Pt. ACE 5) 1 s r p1 6.00 REF. - - y j'
.e
_i.;a. 030 \ e M. _ -- F- -
/
u v :
//
\ , ..: r
. \l i n' *9* * '1 QQQ *- % .< , C
( ~ 2'-1 0 " *- + 1,
- - s '- s '"M --- - ~ ~I I
h
- 3. LOCATE h MIDWAVBETWEEN h(hi{"
- 2. CROSS PIECES (k)/( ARE h STK. x 2"d (LOCATE AS SHOWN ) -
AN G LE S p~,J. @ AN D -[' A R,E 2 /t XZ/z"XM4 l= p .- . L
hE.Y, =
. _ _ . g '. o " ' -
~ " l'-T'dt.G. (2h7
/
3 2'-C LG.(k
/
b( 3 m / g ,
-r .
5_ . . l'-8.25" 0.UTSIDE ' FLAN 6C ~ ~
. - u r-A - %OSCD .-. .
, W p*
I \l
' d H 100 d ('d .50 TV P.
- e. ..y.9.75 ' b c
c'.id*(2) --H TI APERTURE CARD Aho Available on dperture Card g 22zo43k~ % Figure 3-2f Locating Frame Details Sheet 6 of 9 Revision 1 (April 1977)
- m. _ _ _ _ _ - _ _ .
w, d .. 4'2.25"0
-( .
G . . . . . . . . . . _- ..
. . . rc
. _ _ . .. g'_9 " t
= -
5 '- 3" '-- 3 TYP14)4pO , r TYP. 8 Pt. ACE 5 7 y/7 TYRh 4 o s x f7 ~ 030 o / C- 4.00 -- - g
@y/
Czi __ tee , .h
.s' '>
.- 5'- 8 ' T2k -- + !
- 3. LOCATE h MIDWAYBETWEEN Z. CROSS PIECES / ARE h/h d " STK. X 2"d
*(
.(LOCATE As SHOWN )
E 7
- 1. ANGLEp @ AND Q Agt 2pg xzfz'xg*
L rx_-_ -_ ___-__ _ - - - -
a f t l'p. = h._. . ____ _ .. g '_ o " c- = l'I"d(2.(2)i L 2 '- 4. ' dLG.(h i I ^ 11
- i (3 (o
@7
\
~
\ i
. 1 i~J fp h I 'M g.y. _ . . . a l'-8,25 " OUTSIDE FLAN 6E-7
- sA ._.,,,, y _ _i 9 TVP, I T' - 3' W\6((g #
+ 3.00 d (2) .50 M ,
~
- .. 3L 9.75 L +
.. v.;d*(g _ _ _ _ .
TI J APERTURE CARD l-Aho Available On Aperture Card s 'zz204%-o7 Figure 3-2g Locating Frame Details i-Sheet 7 of 9 i! Revision 1 I' (April 1977) l e
16'-2.%5" R (
=
, gtg a 9_
5 '- 3" '
)
TYP 14) y ' TYPh-TYP. 8 R. ACES 4 4
- 1055 REF,\ $
f[ /] ' f
.030 " / l/ l 2
~
9 l A.00 *+ + 5 H (1) ~2 lo" L *-+-
= ;
s'-s"T2) =- I i
- 3. LOCATE @ MiowAY BETWEE.N h.hif" d " STK , X 2"d I
Z.(LOCATE. CROSS AS SHOWNPIECES h/h)ARE
< t. AkGLES @wo O ARE 2't/"x2/z"Xl/' 4 l
l 1 L-~ )
.A t
=
$F. ^- .
=
I4 '.- O " ' l'-Y dL6.(2 2'-6' dL4.(4) 3 le 8 7
- 12. ,
N / 1 H
- %] '
ee yl'.8.Z5 as . 00TSIDE : FLAT 46E- [e -
,s 9
N P, . ., . ' l:' ' '- J. g t' 4 \6 .
+ 3.c o (2) d
# 0' " I 4 .
3-9.75 1 #
=
G' ld*(f = TI APERTURE CARD Also AvaMahle On Aperture Card P\ u zO4 Slo 43 Figure 3-2h Locating Frame Details Sheet 8 of 9 Revision 1 (April 1977)
% w i
I4'-2,25' REO e
- 5'-10.5" * =
- I E 8d3 +3 to"
- r - - +g 5 it 2
~
~'
/
tdi//C.12 l e WR(lo)4 if ~ v ~ ~3,a g A// 4-*-a + si F
-= .g s,5" * -
- lgte l=
=
se e - ;
=
l .. l . ysw.xzd x 2s %.(e @ l 1.LotATE CROSS PIECES 4'h A5 SHOWN
< i.
ucs@ @ @ wo @ m eurg
', =
d 7'_o 6
=- .
D 7 = 3'-ts" *- 9 I'5TK. x2'dx1535"d to. (1) j/W .
- 17. OUTSIDE FLANGE .TYP.'
- j. ,
l s 3od 6
'e 4
. 60,TNP.
= 48.5" * =
1 TYR 64'))y p 4 i TI ! 14' 0" ' = APERTURE CAltD Also Available On
' Aperture Card
$)IM lb :bb Figure 3-21 Locating Frame Details Sheet 9 of 9 Revision 1 (April 1977)
N.M
- g. .
^ L:
= 'r2.!8"I~REF.
'50 STK. 4
= 5,f l
6 0 .
'I
.t. A .06 TYR 2 m r
r I
- i. }
\
4 PLACES
/ f [1k ' '
W
-A-i s!
a- sw
/
L_. _ j Allh" !R (6 l a n 00 5.844 _+:03 q. ' (O l f jL
- i
.so'sTx.(t)
D
~
12.187 I$ TI APERTURE
, CARD 9 Alm AvaBahle On Aperture Card
*0bb ll 0 *N$ .
p' .005 o .005_ E l 2 z2.043/o -/0 Figure 3-2j Cruciform Detail Revision 1 (April 1977)
~~~'""~
I f
..j.
l .o , ,,
\
. *.p ' l
. _~
4 _ s. _ ;_. ' a1- r __ y $' d : ,
.h
(_ l
. ?g4 -- g1 o .__ _ _,.. __
. 6_ _. a L c- ,
m .;
! $2" {; 3 fi 0 ^
M ~;
- ' p'; I e
p g ,
=
f\g,, gvy ,+m w-{ C;__y. r n.. w ,:
.=
4 ,. p g; 2 -. J. P i Sa:.g'- gqx *d~I g llrt
.I r- m-a6 I w ;
i si i ff . ~4v '
'f c- m o
_ Y' _ gIs &v _. # V'1 '
- o A
h
/'
"""[ j {lt [h-
.f f ~
v -
/ w.:I a
- - - . e a
d e g -
. h me &J i
'! 01 '
w 3,%j .*e c
' ~
!.,p
'r N
JhJge. 4 d; .J f, g V.. .. , e'/
... X m\
N t" . R< 3 N-M
- a. i
$a, X
du i 0 S ,31.w$
. g J.
l3 < l bg f\ , . \\ ' l']il , b*
- r;t \,hr+ +h4 g A - pg 8 i
i W -s w> D
,~
ry .j@3 g
. p. -
fs a mNw --- w./ - ., \, v 4:" 3 , e, . y'
.a n ,5 . _"
dg ggw -[,
~. -
i-
- */ 1 l- y g, a a
+ y( gq _
- P % .)5 y
\ e ..,
s .y _ n' ;
..Gp.-p Yp d, - m
.e 1
,g 4= .N
'I a s
m ig- ; I . g 1
~ '
8 g- h oo .
-9 1. u .._ - E .' '
l
- ) "
l .- .l , j % . s! . ? j Q . j, -
~
. J); {
i t y ,. nl e .~ _. ; _ _
' gign egr.. g_..
mh t. mz .~ m _~_ ,-, ,- , - - w/- @; J ;\. .I ", -. e i
- g. i..
e g ~ . / - < w'
-t- -
. .cz.z--7m & s
- l. .
.. . - - - -. - - e y ~ .. . ,
' '5 .,,>,lc /' so .- [ _l' _ 'E ~ :,
~ ,
- , y _)
3 I
/ 4
\. i
., ' I ,j- T, 'j '
y5 1: - ** l ,s s 4: 2 7-1s g
- 6[, .a .
e '9 !. O A i W f'., "' 3 y f ta 0, yi l' L .- !. ,Ji , l: 5' , ( -
- i l '. g
, ,t s
)
-_.i _ _ _ _ _ _ _1.___..__ _ _ . _ . _ _ . .
( L - 7 7' 's.., i . j e 4_ . ~ ' ' I l 7~
/ ~
i - _ i . _ [4 -
, 1 D _~
C - /' _
. ._ Z_ .
N . hc,h p f W _ } f f- g l fx y_, 0,0 ' /
/, ,'
- /
- /',, s N
c I Q N < h X,lpd. Q h g j y-b I b - D
,,0 ga s
s k W
- l ,
)*0 G,, f , %
.s a_
3' / . a } r e
]_ ,
)
g 4 j.,
- f ;
fj p l ie 4.
, F 1
8 J
. ! i
( q'- Q _ N I' '
/ \
/
E LW g D o P
\
l
& _ p4 fI :
,(. ' " -
l')- M y gE,.,
,. X X "T* ,~ : - -1 raw -
~
*O S
E R es
,
- t X_i:7XX* T' i C
t t D 0 4 NA F
*- 1 _ C ^ h~ -- ~- 1J v " T 1
"' A3
' ' p/X a
a a c X X i? ^ I T
+
~- ]
y a FE R NJ O I S n o i T C E P S m f'lb T o d" X XX* ' i v ? , Ci ti= T c E TC ( e e7-6 - S E a
, ] i f ~ ~ - :~ iJ v
- S
)
h t u t r
^
c o
.i , X, s
XX* I ~b~ t t
~ -
4"-
]i v
- n E
T M t t a fd
' c
{ w . X ea 5 3 7+ ~ rE _ :h s
- t T
o s 5 *L
- os -
a
="
C - m***_ a' c ' . (~ m6
~
D s.'
- h. s s
e. e s'
- s. C -
- n
- t n
_ E . u a A. y F
~'*g,mml c o c
t s. j '
, Ll iI l Ii'
/
Y, q.l1,f ' &t m,; 4 PI i O
,ll l ( l l I W
P .s I _
;i l
_ i W i ' ;6 j' i .5 es D Wj h{gl i
'I
-A vg[ .
. l g
,Q!' d;l_ !
,i
_ y . l
'(
, iII l
~
; l 3 . . g' . i a-. hy'VNl I
E L
- ( il .'
1 - l. i
\, ,C" 4 k!
J ! l l ,
, l l
iif lM ^ t I_. k' S
- i-
}1l o,
,[
e 1 . i -Ii
; L l lIl
+
R.
.T j j i' I l : ,
Y,
- S
.c E nf
_7P
* )i!i C
b
' {.
.<= \
- t Z
n o. 1 , b / i , Iir
,il T
Y P s
. Ii lr ,
j \' ' d L dI t i l l e
;4 , i I
s
- 3-
~.E - 2 1
.f T
i
.s
- . ' * - )
.m W ; -
i i!I b 5 i t t fL-I l
.c e.c
( u
\
l ,! f
,'I
' l
/
/d l-N. '--i ._
_yt f g
- 3 a
- ' j i j [] ' / f C
,s l/' ; '
!"v ; .
- i .j 1. }
_m m w a.e.e6_ '05
,/
t l ,' H
/
f l al j
/
- - j ' ,
/ '
M [ j ji I l
.$ / g
' A
- g. - bd@l/qmaI ikd,4!
E
' =
R s ') T u t
- a i{ -
F - d
. e
, p n 6 l- ,
y 4 s, # f ! l fI) ( k g
- A Lt 9
Q.'m" . [ g g'y! ' g
, . 5 t
o t.m - . ~ a C 1 I n 4
- 1 [. hll> < i l
b a, 9_ gl,1[j:,, '
' < l
!i/ g'
\j% _
T
., /-,
- 1 n
E ( t E
- ?y\
, t C I
l ' l l l , j- i
'1( ,
E
$ ( .
3 l l ( . .l l ;
]
l ; +/ l
\u '
l l _ / h 6 t I
, l 4
l ' t j ; 1 I. ,i I ey' [ !' I OA Tg u i , f
' : ll l !
A TTk c Nl
; j '/.< SI nV DEO l ; l,, ,
' , j a!
ji r
;g; i
,\
s l I i ' j l ( ( l j
}
L
%RE Na g
l I l ! l f l lil ,- 1t T (m
,Qo l ii iL hq' t , I I l,l l !
i i I l I( l I f l
?s A
SiT E 3 E
*A.
h.
- .l' [!t
e
., O -
4 ~ [ Outer Coll: Boundary. u J. 51 e 6.650" = l -
- 9. 3:i5" -
i l' 1 1 ,.
% +
0.1875"
, 0.267"/
304SS/
.- - J - -
is'ter Fuel Assemt ly Region I _1.9825"
-: 8.4;.6" 5 HO 2
4.480"- i-HO - 4.6675" : 2
^
13.3" = ' \ l Figure.3-5 Reference De' sign f: l i
k/ 1 8.426 AEF > 416 SPACES AT 0.496 = 7.936 -> FUEL ASSEMBLY 0.245-> 4- 0. 530 ->
- g g0
-> 4- O.122
}
g 8gO""O"""""8 8 g888-
- 0. 058 -> p O O O O O O A
' ' ' "' ' ^ S 8 ' '
'ND CONTROL R00 0+ 0.496 OO O O 0 0 0 0 30 g' O - O 0 0 0O PtTCH h -> 1 488 4. 0.068
- o k
- g O O as- 0 Og g .O O4 e e g O 1
o + 2 976 -> O O O ' 0 O 0 0 0 0 00 0O O O O O O OO O O Q O O O g g g@ O c>U I DE Q Q Q O o o j THIMBLE s O O O 00000000000000000 00000000000000000. t 8.466 00000000000000000 0000000000000000 o
] 040 .
TYP O O O O QS, 9,,/, O O Tye 1 O O - p O O O O O O O OQ O
.] .p O OO O O 00O i O Q O O 1 O O O O O y
]
O O ::03 -
- -O O OO O1 v-O O O N nSTRuMEnrATi0n i
g 'h
' SHEATH O O O O l l \ b FUEL ASSEMBLY 00000000000000000 0000000000000000Q WITHOUT ROD k b CONTROL CLUSTER CLUSTER CONTROL ELEMENT FUEL ROOS 264 REQ'D 00
- 0 374 2.f 7. h}l p' l l
CLAD THICKMESS = 0.0225 CLAD NATERIAL - ZlRC-4 &.
'2 b-f;q L
k E q. t[7
. f[- ud O
Figure 3-6 Fuel Assembly Cross Section 17 x 17 l i
-i O . 96 -
. 95 -
y. i l
.94 -
4
.]
f 1
'1
.93 _
1
.)
%./ %
w
. 92 -
e a .. 4 Group PDQ-7 'I l
.91 1 KENO II O
. 90 -
KENO II 95%. Confidence Interval
.89 -
-- .)
-j
'l y
'l l
O 88 12 13 i 14 ,. ii Cavity Center-to-Center Separation in Inches- . Figure 3-7 Criticality Analysis Results ; 1
7. i l l l
' 13.300" r I i i N.
o l 9.335"
~
u I Fuel - 8.426" i Assembly 13.300" Region
- 8.96" -
i j 1.9825 '
-> + HO 0.1875 "
2
'2 .1[" i U0 304SS
~
11.13" d 2 11.3175" > v ) 1 l t l O i i j i i 1 l 4 1 1 1
.I l
i Figure 3-8 Quarter Section of a Sixteen Off-Center j Fuel Assembly Cluster j l 4 1 I o
\
l 1 I 1 l I [ . - _ _ . . _ _ - _ _ _ - - _ _ _ _ _ _ - - . - _ . . _ _ _ _ . _ _ - - _ _ _
, i
[
'% )
13.240" t H 4. 2'.!0" H HO 2 9.020" 1 3.845' ' l
\
" 8.424" i I
~ - 0.1875" <
.l Fuel (minimum * #
- Assembly assembly p r 304SS dimension)' I HO I 2
f 9.3950" rj O Figure 3-9 Cluster Geometry of Four Off-Centered Fuel Assemblies i U
s , . 1
'l O
\j l
.93 -
.92 -
. 91 -
l
. 90 -
1 w
, O g W
o .89-
.88-
.87
.86 i
.85 I i 1- 1 0 5 10 15 .. 20
% Voids Figure 3-10 Reactivity. Void. Ef feet, ' 200'F-l.
l
4.0 SAFETY EVALUATION o o l The safe storage of irradiated fuel depends on maintaining the integ- J rity of the fuel cladding as the primary barrier against the release j of radioactive materials. The protection of the fuel cladding is, .} therefore, a basic design requirement for the SFP and associated j l systems at Trojan. j i The design considerations of the new spent fuel storage racks are i I almost the same as for the existing storage racks. The installation of new spent fuel racks was compatible with the previous storage facility, which was designed in accordance with General Design Criteria 1, 2, 3, 4, 5, 61, 62, and 63, as discussed in Trojan FSAR I Section 3.1. The applicable NRC regulations, NRC guides, and industry standards which are met are 10 CFR 20;. Regulatory Guides 1.13, 1.29, g 1.92, and 3.41; Standard Review Plan, Sections 3.84, 9.1.2 and 9.1.3; 1 ANSI N45.2 and N18.2; and ASHE Boiler and Pressure Vessel ' Code, Section III, Subsection NF. Safety analyses pertaining to possible ; dropped spent fuel assemblies and failures in cooling and cleanup {'v j capabilities are presented in this section, which will replace the ! safety avaluation of the spent fuel storage facility described in l Trojan FSAR Sections 9.1.2.3 and 9.1.3.3. A radiological evaluation j for refueling and normal SFP operation is presented in Section 5.0 of 1 this report. 4.1 SPENT FUEL POOL AND SPENT FUEL POOL RACK SAFETY EVALUATION $ i i The design of the new spent fuel storage racks is such that it.is ! impossible to insert the spent fuel assemblies in other than prescribed ; locations. Suf ficient center-to-center distance is maintained between f i adjacent spent fuel assemblies to ensure a k egg <0.95 even if unbora-ted water and f resh nondepleted fuel, enriched in U-235 to 3.5 wt%, are in the SFP (see Section 3.1.2 of this report). ) i The possibility of accidental criticality was investigated for two : i' [] cases of dropped fuel assemblies, one in which an assembly is dropped
.I 4-1
. Revision 3 (December 1981)
I l
1 on' top of the spent fuel. racks, the other. along the ' periphery of the - h 'l rack array. Because-the depth of the water between the top of assem-
.blics already.. inside the cavities and a hypothetical dropped assembly s laying on top of the rack is approximately 9 in., no adverse reactivity
'l 1
ef fect is calculated to occur f rom dropping .a fuel assembly during fuel , handling on top of a fully loaded storage rack. ,i It is also possible to drop a spent fuel assembly parallel to an assembly in the storage array because of. the 2.5-f t unobstructed space ; between the. periphery of the storage array and the side walls 'of the l SFP (see Figure 4-1).. A conservative analysis was performed for an assembly, assumed to be dropped during handling, which lodges parallel to an off-center assembly in an outer cavity. The analysis assumed ~ l j reflective boundary conditions in three directions out to the three > j zero-current boundaries. A fourth boundary, the'zero-flux. boundary, was used as noted in Figure 4-1. Including the dimensional and posi-tional tolerance effects, temperature effect, and model uncertainty, as .! ! discussed in Section 3.1, the annlysis shows that a 6-in. minimum separation between the dropped assembly and the nearest outer storage - J array face is acceptable to maintain the overall subcritical k gge below 0.95. The summarized results follow. Drop case with dimensional and positional tolerance, keff 0.920 Temperature effect, ok 0.002 Model uncertainty, ok 0.008 i Total 0.930 t The structural design specifies deflector plates on the outside modules-i to preclude a dropped assembly from locating closer than 6 in. to the nearest stored fuel assembly, as shown in Figure 3-4 and described'in g Section 3.1.2. Revised versions of cask drop accident were submitted in Trojan FSAR Amendments 1 and 5 to resolve NRC questions. The design' of th'e 'spenr. -- i ( s 4-2 Revision 3 j (December 1981) __ _ ______ __ _ _ _ _ _ - _ - _ _ _ _ = _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . - _. _ _ . _ _ _ _ _ _ _ - . _ _ _ . __:________A
fuel handling equipment is summarized in Section 9.7 of NRC Safety Evaluation Report for Trojan Nuclear Plant dated October 1974; it concludes the design was acceptable, g The layout of the fuel handling area is such that the spent fuel casks will never be required to trnverse the SFP when spent fuel assemblies are in the SFP. Physical obst ructions identified as wheelstops will be in place and evidence that they are in place determined prior to any movement of the SFP building crane. Wheelstops prevent the crane from D physically moving over the SFP and provide compliance with Technical 1 Specifications regarding movement of heavy loads over the SFP. 'Thus, it is considered incredible that a spent fuel cask would ever fall into the SFP. Nonetheless, all the fuel rods in one assembly are assumed to be damaged in a nonmechanistic manner in the fuel handling accident analyzed in Section 15.5.9 of the Trojan FSAR. The proposed modifica- g tion does not affect the assumptions and calculated consequences of the acc iden t . This conclusion is based on the fact that it would take four independent concurrent failures to allow a spent fuel cask to fall into. [' the SFP, These four failures are: 1
- 1) The mechanical stops must fail.
- 2) A mechanical device on the crane must fail, eg, the crane hook.
- 3) The safety slings must fail.
- 4) Administrative procedure control must be violated.
The heaviest load that can be carried by the Fuel Building bridge crane is a loaded spent fuel cask for rail shipment; it weighs #195,000 lb. If this cask was lif ted to the maximum height and dropped, the impact would not damage the SFP. This is accomplished by providing-in the b
~
design of the spent fuel storage facility a separate cask loading pit with a suitable gating arrangement to facilitate underwater transfer of. fuel from the storage area to the cask, as shown in Trojan FSAR V 4-3 Revision 3 (December 1981)
Figure 9.1-1. ' The dimensions of the cask loading pit preclude .the tipping of 'the 'apent fuel cask while it is in the pit. In the unlikely-event of a sideways or end-first cask drop during the _. . 2 rom the l loading pit, the loading pit floor could f ail, but the SFP structure , would not be affected. If the loading pit floor failed, then.only the water in the loading pit would be released since the refueling gate g would be closed during all periods of cask movement. This volume of: water would not result in the flooding of any safety-related equipment. A diagram showing the path for spent fuel cask movement is shown in Figure 9.1-10 of the Trojan FSAR. Initially, the spent fuel cask is placed in the wash pit for cleaning. Nex t , the cask .is moved to the
^
SFP cask loading pit where spent fuel can be transferred into thel cask. Finally, the loaded cask is returned to the wash pit for decontamina - tion prior to exiting the building. Cask handling througho'ut1this evolution will be performed by the Fuel Building. bridge crane. If a spent fuel cask were dropped onto the concrete floor-above a CVCS holdup tank, the tank could be damaged. However, the content of the tank would be safely contained in the watertight concrete enclosure around the tank. For this event, the safe shutdown capability of the plant would not be affected. Any airborne contamination from a damaged g tank would be removed by the Fuel Building Ventilation Exhaust System.
~
If a spent fuel cask were dropped to the ground above.the buried-diesel' fuel oil and service water supply lines, no breaks would occur in these-lines because of the depth they are buried in the concrete. The mnvement of loads over or around the SFP is limited in accordance with Technical Specification 3.9.7. PCE has Plant procedures in the-form of Administrative Orders regarding the Fuel Building bridge crane h and Fuci Handling Procedures for the use of the SFP bridge crane, which will limit loads and lif t heights'over the SFP. There should normally be no movement of heavy loads in the vicinity of the SFP other- 0 v than the spent fuel casks. Heavy equipment that is to be moved through m u /"% the Containment' equipment hatch, which is accessed through the Fuel 4-4 Revision 3 ' (December 1981)'_ , s
Building on the same elevation as the SFP, will be trar. sported along a set of rails that run perpendicular to the hatch centerline. The g nearest rail is 11 ft from the SFP at its closest point. Thus, heavy
- ioads moved in and out of the Containment should come no closer than within 11 ft of the SFP. In addition, since the load would weigh less than that of a spent fuel cask, accidental dropping of this heavy load C would not affect the SFP integrity.
Nonetheless, administrative actions consisting of briefing and training riggers and secondarily involving design aspects such as safety factors used in struts, cables, hooks and rigging gear were implemented. In C addition, it is expected that such work will be supervised at least by Plant staff personnel, although there may be some outside craft people involved in the work. To provide further assurance that the spent fuel rack modificar: ion does not increase the consequences of an object striking the spent fuel, w f reshly discharged spent fuel (that discharged less than approximately
- 1 yr) will be stored no closer together than in every other storage position in the new racks, except for the following short periods of time when the storage distance may be reducedi
- 1) Initial reracking of the SFP. Following reracking, the spent fuel will be located in at least every other storage cell if such work is completed within approxi-mately 1 yr af ter the reactor was shutdown for refueling.
O v
- 2) Removal of a rack module to facilitate repair of a liner plate leak may require storage of the freshly discharged spent fuel assemblies closer than in every other storage location. Following module reinstallation af ter repairs, the spent fuel will be located in at least every other storage cell, if such work is completed within approxi-mately 1 yr following reactor shutdown for refueling.
l l O 4-5 Revision 3 ' (December 1981)
- 3) Discharge of a full core into the SFP. t!ost freshly (7 discharged assemblies will be stored in every other storage cell. Seventeen spent fuel assemblies will be , !
w stored in the diagonal storage locations between the other freshly discharged spent fuel assemblies at a center-to-center distance of 18.8 in. 1 i This storage configuration will result in the f reshly discharged fuel ! being stored on a nominal center-to-center spacing of 26.6 in. compared l to the previous nominal center-to-center spacing of 21 in. Spent fuel I older than approximately 1 yr is stored in the positions between those 'l i
^
occupied by the freshly discharged spent fuel. Any damage to the older apent fuel will have negligible impact on the offsite radiation dose consequences since the isotope of concern, 1-131, will have decayed at l 1 east 45 half-lives. Thus, the consequences of any object .criking the SFP was no greater with the modified racks than with the existing .i racks. I 1 The process of changing the SFP racks did not endanger stored fuel. \ The sequence that was developed for changeout of the spent fuel racks i did not require carrying any new or existing spent fuel rack over actual spent fuel storage positions. The plan was developed to minimize risk in handling heavy loads over spent fuel and required first that the fuel be located in essentially one corner of the SFP in m u the existing Westinghouse racks. When sufficient existing racks were removed for placement of new racks with sufficient storage capacity, the fuel was moved from the existing racks to the new storage racks. That sequence of removal was essentially the same as for the initial Westinghouse racks. In other words, we continued to remove existing
'.cks
- without carrying them over the actual spent fuel locations.
The layout and design of the Fuel Building bridge crane, including the use of interlocks, meet the requirements of Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, as stated in Trojan FSAR h Section 9.1.4. The later more elaborate single-failure design features p which were identified in Regulatory Guide 1.104, Overhead Crane Handling U 4-6 Revision 3 (December'1981)
0
' ' , ia ,
- m. ,
Systems for Nuclear Plants, are not applicable to : the Trojan Nuclear Plant as noted in'the distribution of Regdatory Guide 1.104'and its Value Impact Statement '(NRC letter f rom Guy ' A. ' Arlotto to Distribution g dated February 12, 1976). 4.2' SPENT FUEL POOL COOLING AND DEMINERALIZED SYSTEM SAFETY EVALUATION' The SFP Cooling and Demineralized System l's 'not required for the safe , shutdown of the plant. . If in? t he event of a -loss-of-coolant accident,
^
preferred power is available, the SFP cooling pumps:would continue"to operate U they were operating before ' the accident. - If.only: standby power is available, the pumps' would be shed ,f rom the' power supply buses. Although the Seismic Category II portion'of the. system may not be operable following a design basis seismic event, a Seismic .CategoryfI. a makeup water supply from the Servic'e Water Sys' tem (SWS)
- i 's available.to-maintain the desired water level in the SFP as shown in: Trojan FSAR g-Figure 9.1-4. The only requirement to assure adequate cooling for.the spent fuel is to maintain the water level in the SFP so that the spent:
fuel elements are not exposed. , To prevent any possibility of loss of water f rom the SFP due to failure of inlet piping, a siphon breaker in the. form of one 1/2-in.-diameter hole in the inlet pipe has been provided.. The' outlet piping is connec-ted above the minimum permissible SFP water level. to render impossible m ; w 3 a loss of SFP water by failure of outlet . piping bhlow this level.< ' ; Also, since the SFP cooling' pumps are inspected once a shift, the longest time interval between a pump failure and'its detection is 16 hr.. ., This maximum time interval will occur only if.the first inspection is-performed at the beginning of the shift and the second inspection is performed.at the end of the shift. If the SFP water , temperature C limit is exceeded as a result of the ' failure of SFP cooling- pump, this - g! failure will be detected sooner.as a result of the~ investigation'. .! performed following receipt of the ' alarm. The..setpoint for..the SFP-water-temperature alarm is 135'F.
\
4-7 Revision 3 (December-1981)
- m. - _ . . . 1m.. -.-_.------.---.--m---_---m-l------.--- A
-L 4 1 The four conditions that are discussed below were analyzed to deter- ;
mine the ' limitations of the' spent fuel' storage facility to cool the - spent fuel stored in the'SFP.' 3 i Condi tion ~ 1' - For this condition,- a calculated maximum 1 decay heat load 4
~
mv-w/;\
~
of 4'.12 x'10 7_. Btu /hriis used. To. perform this calculation, sevenD $ - s; spent fuel regions are assumed in the SFP, each with an average burnup - l- of 33,000 mwd / tonne-U. A complete core is also assumed to be unloaded; 1 A. with an average burnup'of 22,000 mwd / tonne-U.- The' regions are assumed' , d,; i to' have been unloaded annually for 7 yr; the full core is assumed :to - have been placed in the SFP 150 hr after r' actor e shutdown. lThe decay heat factors -used are consistent with the decay heat factors tak'en:f rom gj NRC Standard Review P.lan, Section 9.2.5 (revised ' November 24,- .1975) . - "'R If a full reactor core is unloaded into the' SFP at a time =when sevent ,
. regions are already stored in the SFP, .then one RHR train ' aligned with a the SFP would keep the temperature below '140*F. The SFP temperature i would be above 125'F for only 11 days after reactor shutdown. .The RHR.
train and SFP cooling subsystem cannot be used-. simultaneously to cool' i p t the SFP because the pipe diameter for the inlet pipe restricts the flow rate to the SFP. However, the SFP purification subsystem can be used 1 with the RHR to continue to maintain SFP water purity.- 3' Condition 2 - For the condition where' the SFP is loaded with seven ' regions with a decay heat load'of 18.9 x 106 Btu /hr, a thermal analysis was performed for various cooling configurations. ' The seven l l regions are assumed to each have an average burnup of 33,000 mwd / l tonne-U; six of these regions would be . stored for 1, 2, 3, 4, : S and ' # 1 6 yr. If the seventh region is discharged 125 hr af ter the reactor is shut down, then both SFP Cooling and Cleanup System trains aligned with ' , the SFP would keep the temperature below1 140*F and under 125*F after 20i $: (- . days . - If one of the two SFP cooling pumps fails, the remaining. pump and - . j two heat exchangers would maintain the SFP water temperature below 145*F. However, under this situation, the refueling cavity and .the RCS-temperature could approach the Trojan Technical Specification limit'of 140*F. One RHR train aligned with~the SFP would keep the temperature l A-f 3 U . 1
~
4-8 -Revision 3 (December 1981).
.__-________-__-_--.____.--_-A.-____.-__ -_-_l._--.-... -_2. . . . _ _ . - - - - __. _ - __-xa._.s-_, -- ._ ___ _-.__---.a -- _-]
}
below 140*F. The SFP purification subsystem can be used with the REIR train to continue to maintain.SFP water purity. Condition 3 - For the condition where all forced flow to the SFP is 7 lost, and SFP boiling and fwl temperature analysis of the maximum *~ power fuel assembly was performed. It is assumed for this analysis that the water temperature begins at 140*F, that the maximum power fuel assembly decay heat load is 3.03 J 105 Bru/hr, and that the water level is maintained at the siphon-break elevation of 83 ft,11 in. The evaporation rate from a boiling SFP is approximately 90 gpm and the SUS makeup capability is greater than 200 gpm. The ElYDBPOOL computer program was used to analyze the heat transfer from the maximum power g fuel assembly. EiYDBPOOL is an NAI-derived computer code f rom Control
- Data Corporation computer code ilYDROB, CDC-84004400. The program evaluates the physical properties of water in the pressure range from.
15 to 50 psia and includes the effect of static pressure changes on the saturation properties of water along the fuel channel .. Evaluating the Dittus-Boelter correlation, using steam properties, the ilYDBP00L program determines a lower and thus a morte conservative heat transfer G coefficient than would normally be expected for a boiling condition. The steam and liquid velocities are assumed equal for the void fraction predicted by the llYDBP00L program. This assumption tends to overpre-dict the void fraction for liquid velocities in the iange of 1 to g 3 ft/sec. " The results of the flYDEP00L analysis of the maximum pover fuel assembly 3 yield a void fraction of 0.23 and a steam quality of 0.034 percent in the top 1 f t of the maximum power fuel assembly where net boiling occurs. The coolant temperature is calculated to increase to boiling in approximately 3.5 hr assuming no heat loss from the SFP water. The calculated maximum fuel clad temperature is 249'F, and the calculated 7 maximum centerline fuel temperature is 254*F. The integrity of the
- spent fuel 1 assured under these conditions.
An examination of the thermodynamic state in the space between the spent b 1 fuel racks was also performed, assuming that 10 percent of the gamma 4-9 Revision 3 ! (December 1981) l
;g ,
+ >
- i 4
l energy must be . removed by natural circulation between 'the racks.'- A small . amount ' of. boiling resulted at the - top ' of the ! racks , ' but" not' at' elevat ioni, opposite the ' fuel . assemblies. -A void fraction'of-0.40 and-an exit steam quality of 0.075 percent were calculated .to; occur in the h space adjacent tol but 'above 'the elevation of , the' maximum power f uel l i assembly. , Condition 4 - If al'1 forced flow to the SFP is lost, af ter the spent fuel assemblies have ' decayed for some time interval,: then the time :taken for'the SFP to reach.a boiling condition is extended beyond the 3.5 he calculated for Condition 2. j l Table 4-1 summarizes the maximum heat loads'for.two ' design cases
- I (1) return of the plant' to power operation following a refueling when.
.the SFP has been filled with spent fuel, and (2) following.a shutdown j when the full core has been ' discharged to the SFP. Ileat loads for both-the existing and modified rack designe- are provided.
Table 4'-1 shows that the additional SFP cooling system heat, load because of the rack modification is approximately 14 and' 6 percent, l respectively, for Cases 1 and 2. The increase in heat load is small'. a because the decay heat rate from spent fuel decreases rapidly. , For ~j U ! example, the decay heat rate from spent fuel decreases by approximately ! a factor of 10 from 150 hr after shutdown when the spent fuel,.is. I discharged to the SFP and 1.yr after. shutdown when the next refueling is assumed to take place. Table 4-1 also shows that the rack modification will not significantly I affect the Component Cooling Water (CCW) System operation. The CCW. heat load increases approximately 2.3 and 3.6 percent for Cases i and - 2, respectively. The maximum heat loads are well below the CCW design ; heat load for one CCW heat exchanger. Two CCW heat exchangers are l I provided.
+,
4-10 Revision 3-
+ (December 1981)-
1f _. _ _ _ :_ _ - _ _- a- -
The plant heat discharge rate to the river increases approximately 1.3 and 2.9 percent, respectively, for Cases 1 and 2. In both cases the heat discharge rate to the iiver is below. the NPDES limit. 2 Therefore, it was concluded that the existing cooling systems were adequate to accommodate the small increased heat load which resulted from the rack modification. O l 1 O 4-11 Revision 3 (December 1981)
f+ '
, 5l q> , )
}
q r .
"\. .
-TABLE 4-1 R J
MAXIMUM HEAT LOAD'
SUMMARY
f]
--(106 LBtu/hr) '
;:l Normal Operation ( } _
Shutdown ( } . M (After 30-Day Shutdown) L(150 Hr After Shutdown)' y 1-1/3 Cores 3-1/3 Cores 1-1/3 Cores 1 3-1/3 Cores '):
.SFP Heat Load 9.8 ~11.2: 38.7 41.1 o
.J Total CCW Heat Load 60'9 . .62.3 66.4 68.8L j
- l. - Design CCW Heat Load .87 '87 87 87' l (1 CCW Heat Exchanger) o '
j Discharge Heat to River: n 1 Service Water Pump .29.6' 29.7 Operating 1 . 2 Service Water Pumps 60.3 61.1 82.2 84.6: Operating O o D NPDES Limit 79 79 '160(3)' 160' l i Summer Design Conditions: 91*F dry bulb', 33%' relative' humidity, 75'F-river water temperature. (1) Normal Operation: Operating at power 30 days after shutdown for refueling where 1/3 core was discharged toLSFP. No steam' generator blowdown to river. Cooling' tower at 8 cycles concentration. 1 Full design service water heat loads.- d Boric acid and radwaste' evaporators at 50 percent duty., 1 service water pump flow = 20,000 gpm. 2 service water pump flows =L37,000'gpm. CCW flow through'1 heat exchanger = 11,500 gpm. (2) Shutdown: 150 hr after Shutdown. Full core discharged to.the'SFP with 1/3 core in:SFP from previous annual refueling. No steam. generator blowdown.4 . Full. design service water heat loads. CCW flow through 2 heat exchangers = 20,800 gpm. (3) At river water temperatures.below' *F, the cooldown heat limit.is 240 x 106 Btu /hr.' Above GD*F, the limit isl reduced to'160Lx'106 Btu /hr. (4' d Revision'2'
..(August 1977):
C___ __ _ _ _ _ _ _ ____o_______=__1________ __ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _
I
/
v , i Zero Current Boundaries
;. l I: 9.395" :I I
r Zero - ' 596"
# Flux l Boundary /
$1.9225 l
l .
/ __ _ .
. 1875" l ; ; 1.9225" 304S l~ T 1.9225"
/
39.72" Dr p d f [ j 7.000" Fu 1 ~
'Of u d er[d - I
/
D 596 i 13.24d s em y/ ss m y/.~
/
(3) ~3.845-i
<x / j 4 1 .
p8.424"q _ 8.424" y j J l v o 5
)
1
/ l HO 2
u l 1 U Zero Current Boundary j: 11.3175." ;je 13.240" ;j [a] D = minimum separation distance between a dropped fuel assembly and the outside j i wall of a canister in the outermost row of the storage array. l Figure 4-1 Geometry of Fuel Assembly Drop Accident
/,, i I
\
%J .
'. j i
_ . . ]
I n
., j ,
5.0 . RADIOLOGICAL EVALUATION . r 1 X. I dN)-
- Radiation ' source terms and radiation doses associated.with more fully utilizing the storage capacity' of the SFP are presented in' this section. ;
Included in.this evaluation-are the radiation doses to both' plant-per-l
-sonnel-and to individuals offsite. To provide a better perspective on the increases in radiation source terms and radiation ~ doses as .a result
']
of the mortification, calculated releases and doses are provided and ' g compared fur both the~ existing SFP capacity and the modification. l l Common assumptions are used for both sets of calculations', i
'1 i
j 5.1 SOURCE TERMSJ , The source terms include the radioisotopic content of. spent fuel, reactor ., coolant, refueling water, SFP water, purification system waste,'and -j ventilation system air. Source terms for tritium are considered separately. 1 Design basis assumptions are made throughout the evaluation to place l an upper limit on the radiological impact'of the modified SFP storage racks. A] L 5.1.1 ACTIVITIES IN SPENT FUEL Fission product inventories in spent fuel have been derived utilizing. the equation: l .. dA dt
" ~
M+ A F where 1 A7 = activity in fuel R
- fission rate Y = cumulative thermal fission yield for U-235( )' 3.
i l T l 5-1 Revision 2.. (August'1977) ' y
A = radioactive decay constant ( y
.p y a = thermal neutron cross section l 13 4 = thermal neutron flux (3.5 x 10 n/cm /sec).
The thermal flux and fission rate are assumed to be constant throughout the core. The thermal power rate is assumed to be 3558 Mwt. Each 1/3 g l of the core is assumed to be irradiated for 292 effective full power days
]
l per year. Equilibrium burnup under these conditions is 35,000 Mwd / tonne-U. l Since the average burnup expected for spent fuel is 33,000 Mwd / tonne-U, m 9 . the calculated source terms are conservative. One-third of the core is ] assumed to be unloaded into the SFP every 365 days for seven fuel cycles, j i and at the end of the eighth cycle, the entire core is assumed to be
]
unloaded into the SFP. The cumulative inventories of.significant fission f products in the 3-1/3 cores that are assumed to be present at the plant after the eighth fuel cycle are given in Table 5-1. These inventories have been corrected for an additional 4 days decay after the eighth
.q shutdown for refueling.
1 As stated in Trojan Nuclear Plant FSAR Table 4.3-1 the total weight of g l the fuel (UO2) in the core is 101.1 tonne (222,739 lb). l 5.1.2 ACTIVITIES IN REACTOR COOLANT AND REFUELING WATER 1 Reactor coolant that remains in the reactor vessel is dispersed into. the reactor cavity when the cavity is flooded for refueling. Peak concentrations of fission products and corrosion products in reactor coolant during steady-state operation, f or conditione of 1-percent fuel defects, are derived and presented in Section 11.1 of the Trojan FSAR. Following power, temperature and pressure reduction spiking phenomena can lead to increased rates of release of volatile fission products from fuel into coolant. For this reason, reactor coolant fission-product activities, given in Table 11.1-2 of the Trojan FSAR, have been increased by a conservative factor of 64 . l /3 S. 5-2 Revision 2 (August 1977) r __ . . _ _ ._U
Movement of fuel f rom the core cannot commence until 100 hr af ter 3-7N shutdown. ~ During this approximate 4-day period, the RCS (12,250 f t volume) is assumed to be processed at a rate of '120 gpm through the CVCS mixed-bed and cation-bed demineralizers. A decontamination factor of 10 is assumed for all isotopes except for noble gases, which. are assumed to be completely stripped from reactor coolant in the volume control tank. This decontamination f actor is - consistent with , the values assumed in NUREG-0017. To compensate for increased levels of corrosion products in reactor coolant due to crud burst phenomena that may occur following shutdown, no removal of these isotopes is: assumed to occur by purification or-radioactive decay during the 4-day cooldown period. Following removal of the reactor vessel head at 4 days af ter shutdown, the reactor cavity is flooded with approximately 350,000 gal of water from the refueling water storage tank. Rapid mixing between the reactor cavity and the SFP (390,000 gal) is assumed. Resulting concentrations l of fission products and activation products in this refueling water due O (g to dilution of the 35,000 gal of reactor coolant remaining in the-reactor vessel are given in Table 5-2. Fission products may continue to be released from defective fuel asnemblies during refueling and storage. The average temperature of .< spent fuel assemblies 4 days after shutdown is estimated to be less l than approximately 300*F versus an average temperature of 1350 F during i i full-power operation. In the absence of fission, which can lead to
. l the release of fission products f rom fuel by recoil and knockout mechanises( 0) , the only mechanisms for release of fission products b from spent fuel are diffusion'and/or corrosion. The rate of release of volatile fission products by diffusion follows the Arrhenius law:
i D' (T) = D' (1673) exp [ - f ( - 1673)3 . l l u 5-3 Revision.2-l (August 1977)~ < _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ ___ _ _ _ ___ _ _ _ _ _ _ _ . . 1_ _ _ ________________________J
if I-where- ' n l] D'(T) = diffusion coefficient at. temperature T Lsec ~2
'Q
~
i
'1
~
~1 D' (167 3) = 1 x ' 10 .11 sec , diffusion coefficient :at 1673*K E = 82.'kilocalories/ mole,Lactivation energy; "i
~ '
'R = 1.99 x 10 kilocalories/ mole - *K, gas ' constant' ,
- 1 l <
T = temperature,.*K. i This and. similar relationships are not. valid outside of' temperatures': in the'800-1100*C(10)' range. For the' core temperature' dis.tribution- >- E 1
. v.
given in' Troj an FSAR Table' 15.1-5, the 'dif fusion ' coefficient evaluated ' with Arrhenius law is'4.4 x 10 ~12 sec . At'.800*C it' is a 1 factor of
-1 '
approximately 4 x 10 lower. ' For this evaluation, the dif fusion' t coefficient. of fission products f rom spent fuel is'as'aumed to be only . Lj l a factor of 10 less than at operating temperature.- This is~ analogous. k- to assuming that the average temperature of the spent fuel is 1000*C.
~
l Neglecting radioactive -decay, . rates of releaseof- fission products from fuel are directly proportional to the diffusion coefficient.- The' measured escape rate coefficients given-in. Trojan FSAR Table 11.1-1, divided by a factor of 10 ,3have, therefore,1been used to conservatively estimate the release of fission products into-refueling water according: ' to the relationship: R = A7 . f!v i where -l [ . -i R = fission product release. rate-Ap = activity in fuel [ ' 5-4 .
~ Revision'2ji (Augusi 1977)i 'i
. ei
E
\
f = fraction of power produced by fuel with clad defects fh 5-- V= escape rate coefficient.
'One percent of the ejected fuel is assumed to have clad defects.
The concentration of fission products in refueling water due to leakage out of defective fuel .is given by: ; dC dt
' ~(+ R p 1.
where i C = concentration in water . ! p j V = volume of water a Xg = purification rate constant. The mixed-bed fuel pool demineralized processes water from the SFP at , i a rate of 250 gpm, and is assumed to provide a decontamination factor of 2 for cesium and 100 for all other isotopes. These decontamination {} . f actors are consistent with the values assumed in NUREG-0017. Result-; ing concentration in refueling water at 4 days after the eighth refuel . [j ing shutdown are given in Table 5-2. Corresponding concentrations to the 1-1/3-core storage capacity racks are shown in Table 5-2a. These values assume instantaneous equilibrium between release and loss of fis-sion products from refueling water. l 1 During refueling operations, water in the refueling cavity it assumed to be processed at a rate of 120 gpm through the CVCS mixed-bed and cation-bed demineralizers, and water in the SFP is assumed to be processed through the mixed-bed fuel pool demineralized at a rate of 250 gpm. For uniform mixing between the refueling cavity and SFP, the resulting concentrations of isotopes in refueling water at the ' f, ( ,) end of a 2-week refueling period are given in Table 5-3. 5-5 Revision 2' (August.1977). ___ __ ___ _ _ _ _ _ o
5.1.3 TRITIUM A plant activity balance has been performed to estimate levels of tritium in water of the refueling cavity and SFP. Design basis rates of tritium production in reactor coolant of 1588 Ci during the initial ig fuel cycle and 1501 Ci during subsequent fuel cycles are estimated to apply to a 3558 MWt plant . 3 The TRIT computer code has been used to compute tritium activities as a function of time ( . Principal parameter utilized in the computation . 3 is given in Table 5-4. Tritium concentrations in reactor coolant, refueling cavity water and SFP water are given in Table 5-5. \ Evaporation rates of water from the refueling cavity and SFP have , N been estimated from the relationship (13) . E = 1.18 x 10 (70 + 0.7 u )(1, - 1,) where E = evaporation rate, gal /ft day u = air velocity across water surface, mph i = vapor pressure of water, mm Hg 1, = vapor pressure of air, mm Hg. The average air velocity across both the refueling cavity and the SFP is estimated to be approximately 1.0 mph (88 f t/ min), based on the methodology of Reference 14. The average water temperature has been 3 assumed to be 125'F and the average air temperature 70'F at a relative humidity of 43 percent (corresponding to an annual average outside air temperature and relative humidity of 51.3*F and 80 percent, respectively) . 9 l 5-6 Revision 2 (August.1977)
3; a' ,g 4 5.1.4 ACTIVITIES 1N SPENT FUEL POOL COOLING AND DEMINERALIZED SYSTEM f i
\m / Activities of the filter and mixed-bed demineralized of the SFPCDS include separate contributions from isotopes that are present in reactor coolant prior to refueling and isotopes that are released I from spent fuel in storage. It is conservatively assumed that all reactor coolant origin isotopes are recovered in the SFPCDS. The SFPCDS is assumed to operate continuously throughout the year and to accumulate l activity at the rate:
dA I D -At ~
~dt p R ^D i where !
A = activity of filter and demineralized j D C (0) = concentration of isotope in SFP water at p t = 0 due to leakage f rom spent f uel l
)
[~'h V = volume of water in pool (_s/ I l A = purification rate constant R t = time from start of refueling. 4 Isotope concentrations initially present in SFP water from leakage out of l 1 spent fuel are given in Table 5-2. Table 5-2a shows the corresponding SFP - 73 l w h water concentrations for the existing racks. The accumulations of activity 1 on the SFPCDS filter and demineralized af ter 1 yr of continuous operation . u are given in Table 5-o. Expected activity levels with 0.12-percent f ailed i fuel (NUREG-0017) are approximately eight times less than the levels shown in Table 5-6 for 1-percent failed fuel. ,s l S l Since the activity levels in Table 5-6 are substantially less than the shielding design basis for the filter and demineralized compartments (see Troj an FSAR Tables 12.1-3 and 12.1-4), the increased activity levels should (O %.J' i 5-7 Revision 2 (August 1977)
i l not result in more frequent filter and demineralized replacement. There-gg fore, the volume of. solid waste f rom the SFP purification subsystem should
\~ / not increase as a result of the rack modification.
5.1.5 ACTIVITIES IN VENTILATION AIR Isotopes present in refueling cavity and SFP water can become airborne. by evaporation and by partitioning between water and air. Separate consideration is given to noble gases, tritium, and halogens / particulate. All of the noble gas isotopes that are released from defective spent ) fuel assemblies are assumed to be immediately released f rom the rurface of the pools into the exhausting ventilation air. Concentrations of these isotopes in the ventilation exhaust are thus given by the ;; equation: A f" -At
=
F C e 7 V f% g where l C = concentration of isotope in air Fy = ventilation exhaust flow rate. t Values of A at the start of refueling (t = 0) are given in Table 5-1. 7 Resulting average noble gas concentrations in the refueling cavity ( and SFP ventilation exhausts are given in Table 5-7. Tritium concentrations in ventilation exhaust air are given by: C E
~
- A F y
1 The evaporation rates (E) given in Table 5-4 have been increased by a factor of 1.52 to account for a possible 140*F ' water temperature during
-s refueling. Resulting sirborne concentrations are given in Table '5-7.
l sv/ 5-8 Revision 2 (August 1977) t i
-----------.___m_ . _ _ _ _ _ _ _
a A Particulate'and halogen concentrations in ventilation exhaust air are
/~} given by:
CA =KC p 1 where K = partition coefficient, (pC1/cc-air)/(pCi/cc-water). Based on measurements of radiciodine in refueling water and' refueling. cavity exhaust air at Monticello during 1975, the value of the parti-tion coefficient at a water temperature of 75*F is estimated to be no l
~
(
- greater than 5 x 10 .
It is not anticipated that changing the maximum SFP water temperature from 125*F to 140*F will significantly change the release rates of iodine from the SFP area. The temperature' increase is relatively small and of short duration. As discussed in Section 4.2, if a full core is discharged into the SFP containing seven regions, the pool water temperature will remain above 125*F for only 11 days. In addition ~, based I_ (j\ on data in " Calculation of Iodine-Water Partition Coefficients"' (Parsly)(17) , it can be expected that the iodine partition coefficient will decrease
^
j as the SFP water temperature increases. For example, Parsly(17) shows O that, for the stable iodine concentration and pH of the SFP water, the partition coef ficient decreases by approximately 40 percent when the -tem-perature increases from 122*F to 176*F. No values for the partition ~! coefficient are provided for 140*F. On these bases, the iodine release rate was calculated assuming that the iodine partition coefficient does not change significantly over the SFP water temperatures encountered. Halogen and particulate concentrations in refueling water ;(C ) pat the start and end of the eighth refueling period are given in Tables'5-2, 5-2a and 5-3, respectively. Resulting airborne concentrations are given J in Table 5-7. 5.1.6 ENVIRONMENTAL RELEASES Ventilation air from the refueling cavity and the SFP is exhausted to (.
'(- the atmosphere through HEPA and charcoal filters.(charcoal for SFP only).
5-9 Revision 2 (August'1977)
- j q
l t Contributions from isotopes . contained in reactor coolant at the start fg of refueling and f rom isotopes that subsequent 1y ' leak f rom defective i
.f V) fuel have been included in estimates of annual atmospheric releases.
Releases attributable to isotopes initially present in reactor coolant were calculated from the equation: l
-l (1.- f p) F yCA(0)
Q= (x+x) (1 - e_g43)) R where fp = fraction of isotope in water at t =.G # rum leakage ; out of spent fuel-(see Table 5.1 -2) Q = release rate of isotope to atmosphere A = purification rate constant for combined CVCS R and SFPCDS operation 1 t = time since start of refueling (2 weeks) V Fy = ventilation flow. Releases attributable to the. continued leakage of fission products out of spent fuel were calculated from the equation: fy yF A C( Q= x (1 - e-At) where t = time since start of refueling (1 yr). A combined exhaust flow rate (Fy) of 34,375 cfm was assumed for the' reactor coolant contribution, and a time-weighted average flow of 19,952 cfm was assumed f or the spent f uel contribution. Tritium and - ., i nobic gas releases were handled as special cases. b I -p G , 5-10 Revision 2-(August 1977) l l
.._____m__ _____.___.___._______________________.__..____._______.___________._..__._.__._____._____.__m_. _. _ . . _ _ _ _ . _ . _ _ _ _ _ _ _ . _ ________._m.._______._.__.__.____._.______ _ . _ _ _ . _ _ . - . _ _ _ . _ _ _ _ _
J The' area of the. Fuel-Building housing the SFP is exhausted by two 19,375-(~% cfm SFP Exhaust System fans, one of which is normally in operation. The SFP Exhaust System, described in FSAR Section 9.4, contains HEPA a'nd char-coal filters. Potential additional filter loading due to the rack modifi- l l i cation, is not expected to increase the frequency of charcoal and HEPA filter replacement. Decontamination factors of 100 and 10 were assumed for HEPA filters and charcoal beds, respectively for releases from the SFP. A decontamination factor of 100 was assumed for particulate releases from the' Containment, q Resulting annual release rates to the atmosphere are given in Table 5-8. For comparison the calculated annual release rates to'the. atmosphere for 1-1/3-core storage capacity racks are also given in Table 5-8. O v . The expected change in radioactive gaseous effluents f rom the SFP as a result of the proposed modification are provided in Table 5-8a. Presenta-tion of expected releases is consistent with Appendix I of 10 CFR 50. The expected change was determined by subtracting the values in Table 5-8 for 1-1/3-core storage f rom those for ' 3-1/3-core storage. This differ-(S i ence was then corrected to 0.12-percent failed fuel for. fission products. f Use of 0.12-percent failed fuel for expected release rates is consistent ; with NUREG 0017. Since tritium release rates are not affected signifi-l cangly by the failed-fuel' rate, hte expected tritium release rate is the . difference between the release rats for 1-1/3-core and 3-1/3-core storage. ., l The values in Table 5-8a include the increased tritium releases due 'to ! I increasing the SFP water temperature from 125*F to 140*F for the short time period following the discharge of spent fuel. l During normal operation of the SFP, there are no liquid effluents released from the SFP to either the liquid radwaste system or offsite. Th ere fore , l the modification of the SFP rack design will have no effect on total liquid releases from the Plant. s 5-11 Revision 2. (August 1977)
1' , 5.2 RADIATION DOSES N,_/ Radiation doses to plant personnel have been evaluated ~ using the radiation source terms derived in Section 5.1 for. spent f uel, refueling I X 1 water, SFP. water and ventilation air. Radiation doses to offsite individuals have been evaluated using toe atmospheric release rates derived in.Section 5.1. 5.2.1 DOSES TO PLANT PERSONNEL 5.2.1.1 Direct Radiation Dose From Spent Fuel Assemblies H l The direct radiatioa dose from gamma rays emitted by 3-1/3 cores that are stored in the SFP has been evaluated at the outside of the pool l valls and at the surface of the pool. Fuel burnup history is described in Section 5.1.1. Decay heat relationships have been used to compute the fission-product inventory in the fuel 100 hr after the ef(ith shutdown. Dose rates outside the SFP walls are as follows: East and north walls 0.1 mrem /hr' i West wall 0.3 mrem /hr 1 South wall <0.1 mrem /hr The dose rates outside the north and west walls are substantia 11y'less than the design radiation zone level of 2.5 mrem /hr. The . dose rates outside the east and south walls are substantially less than the design radiation zone level of 1.0 mrem /hr. [ The dose rate at the surface of water in the SFP due to radiation ' emitted by spent fuel in storage does not significantly contribute to the total dose. The ' conservatively assumed dose rate at the surface of [$ the SFP (normal water level) during handling of a spent fuel assembly due to radiation emitted by the assembly is 1.0 mrem /hr. This is below the design radiation zone level'of 2.5 mrem /hr. t L 5-12 Revision 2 I (August-1977)'
y o 5.2.1.2 Direct Radiation Dose From' Activity in Water
- 7-~g
'\~ / The width and depth 'of the refueling cavity and SFP are much greater =
than the. relaxation length of all expected gammas that are emitted by isotopes contained in the water. Under these1 conditions , the direct radiation dose frem these isotopes at the surface of the pools is given by: D =~1066-C E p Y where i I D = tissue dose rate at poo1~ surface, mrem /hr-l u l ! C = concentration of isotope in water, PCi/cc l l ii = average gamma energy, MeV/ dis (8,18,19) , {} Y O (, ) Dose rates on the fuel handling bridges and along the edges of the posis have been reduced by a f actor of 0.7 to account for attenuation -) i by approximately 1/2 in, of steel between the pool and operator'onLthe : bridge, and for geometrical attenuation along the edges of the pool. This factor reflects recent measurements at Oconee( } . !
]
m l Based on the refueling water activities given in Tables 5-2, 5-2a and t3 f 5-3, the resulting dose rates from this source at the start and end i 4 of a 2-week refueling period are given in Table 5-9. 5.2.1.3 Dose From Airborne Isotopes The external (submersion) dose from noble gases and the internal (inhalation) dose from halogens, particulate and tritium have been l determined for the previously derived concentrations of these isotopes in ventilation air.
'\
5-13 Revision 2 (August 1977)
m . g ,
; . c. ,
J5 , a
- i. b t. , si e 4, q_
1- q The submersion' do'se from noble -gases has been ' evaluated for semi- ' p infinite cloud' geometry with the equation: 5\ 1
._ __ U Do x = 0.229 C .A
.(Eg:KSx .+-E Ky;_yx) !l 4 ,:" 1, where- . h i
D = dose rate at Ltissue depth x,f rem /sec x ; C =. airborne concentration, pCi/cc A i ' i = average beta energy, MeV/ dis (18) g j l g y .l E = average . gamma energy,' MeV/ dis (18) Y ,
'l K =l beta depth dose factor at tissue' depth x, 0* , rem-tissue / rad-air i q
j K = gamma. depth dose factor at tissue depth.x,.' j T* ' rem-tissue / rad-air i 9 "J
'l 1
0.229 = semi-infinite cloud conversion factor giving .) dose to air. ;
.{ .s Depth dose factors for tissue depths of 0.007 cm and 0.1 em were used: ,
^ l to evaluate skin and whole-body. doses, respectively(,21,22)- . Based on . S the airborne ~ noble gas concentrations given in-Table 5-7, the .)
resulting submersion doses at the otart and:end of a 2-week refueling period are given in Table 5-10. , q Inhalation dose rates from airborne halogens, particulate'and. .
'l 1
tritium were determined with the equation:- n - D =.(def) b C A ; O . . 1 4 5-14 ' Revision 2-
.(August 1977);
;l
, a- 1
- q 1
where (g D'= inhalation dose rate, mrem /hr 4 def = inhalation dose conversion fact'or mrem /pc1 i b = occupational breathing rate,1.25 x 106cc/hr t C = airborne concentration, uCi/cc. A 1 1 Dose conversion factors have been derived f rom the maximum permissible ; q concentration and maximum permissible dose recommendations of the i International Commission on Radiological Frotection . Based on the g l airborne radioactivity . concentrations given in Table 5-7, the resulting . ] i inhalation doses at the start and end of a 2-week refueling period are given in Table 5-10. If concentrations of airborne isotopes were to :{ i actually approach the values given at the start of refueling in Table i 5-7, respiratory protection devices would be utilized to reduce personnel p inhalation exposures.
]
\ :
5.2.1.4 Miscellaneous Sources of Exposure l Miscellaneous sources of exposure associated with refueling and fuel storage operations include activities associated with SFPCDS filter and i i demineralized operation and replacement, and refueling operations such as refueling cavity fill / drain and reactor vessel head removal / replacement. l 1 l The design basis accumulation of activity in the SFPCDS'after 1 yr of continuous operation (see Table 5-6) is substantially less than the shielding design basis for the filter and demineralized compartments (see Trojan FSAR Table 12.1-3 and Figure 12.1-14) . Therefore it is l anticipated that the increased radioactivity collected in the puri-fication subsystem will not result in more frequent filter and- , 1 demineralized replacement. Demineralized backflush and subsequent j solidification operations are conducted remotely behind shield walls. Filter cartridges are handled remotely with a specially designed shielded -
%./.
5-15 Revision 2
. (August 1977) .j
/r transfer vehicle. If the spent f uel rack center-to-center spacing is-n reduced to 13.3 in. .. the above considerations, together with their in-kh frequent nature, ie, once per year for filter and demineralized replace-
,ments, are not expected to noticeably increase the dose to plant personnel.
During refueling operations in the empty refueling cavity,. personnel can be exposed to radioactive crud that has deposited on the walls and floor of the cavity. It is not possible to accurately calculate the expected crud accumulations on the cavity surfaces. Operational experience at Oconee indicates radiation levels of 200-350 mrem /hr on the floor of the cavity af ter refueling, and ' levels of 25-35 mrem /hr and 100-125 mrem /hr in the vicinity of the reactor vessel studs before N and af ter refueling, respectively(20) If the spent fuel rack center-v to-center spacing is reduced to 13.3 in. , the dose rates would not be expected to significantly increase. 5.2.1.5 Plant Man-Rem Doses p Estimates of personnel man-hours required for a normal refueling i) (,j (replacement of 1/3 core) and for the unloading of a full core are j given in Table 5-11. Conservative estimates of total plant man-rem doses from refueling and fuel storage and inspection operations are also given in Table 5-11. Corresponding man-rem doses for 1-1/3-core storage capacity racks are given in Table 5-11a. The dose contribution , from refueling activities is based on previously derived radiation dose rates occurring at the start of the refueling period. The dose contribu-tion from fuel storage and inspection operations assumes 400 men-hours j of exposure at the previdusly derived radiation dose rates occurring at I the end of a 2-week refueling period. l ! i j Comparison of Tables 5-11 and 5-11a shows the estimated increase in man-rem
]
as a result of the rack modification. The estimated increase in man-rem g I exposure for a normal refueling is 0.21 man-rem /yr. This increase is ) insignificant compared to the approximately 45 man-rem exposure received - at Trojan during 1976. .i i
/~'T, ] ,
(./ ] 5-16 Revision 2 (August 1977) l I J _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ Q
i 5.2.1.6 SFP Modification Exposure O Exposare estimates were made for~three separate cases.. The factors used in developing these radiation estimates are shown in Table 5-14. Each case, with its associated man-rem estimate, is listed below:
- 1) Installation of new racks prior to 0.4 man-rem refueling.
- 2) Final installation of the new spent fuel 17.7 man-rem storage racks is delayed until af ter refueling, but all the preparatory work described in our letter to the Atomic Safety and Licensing Board of Septem-ber 20, 1977 is completed.
- 3) Final installation of the new spent fuel 50.3 man-rem g
storage racks is delayed until after refueling, but Phase 2 of the preparatory. 9 work described in our letter to the Atomic Safety and Licensing Board of September 20, 1977 is not completed. The estimates displayed a severe radiation exposure penalty associated with the installation of the spent fuel storago racks af ter refueling. The new spent fuel storage racks were designed for installation in a dry spent fuel pool. The amount of time required to install the new racks af ter refueling was considerably greater than for other plants' racks which were designed to be installed in a water-filled spent fuel pool. In actuality, doses were substantially lower than the 17.7 man-rem projected during licensing hearings. As recorded in Table 5-15, the total whole body exposure was 1.4 man-rem. 1 0 5-17 Revision 3 (December 1981)
' The lower doses occurred because the SFP water had -a lower radio- .)
; . activity level, than was anticipated and used for the origiani estimate.
3 , This lower radioactivity level was due to the determined effort to , purify. the SFP water prior to and : during the ' rack replacement.- 1 1 5.2.2 SITE BOUNDARY DOSES Radiation doses have been evaluated at the north' site boundary (662 tn),' the offsite location having the highest annual average-atmospheric j dispersion factor. The submersion dose from noble gas releases'and the inhalation dose from halogen, particulate and tritium releases are. evaluated with the same models as described in Section 5.2.1.3.. Continuous occupancy at the site boundary is assumed, and 'a breathing-rate of 20,000'1/ day is applied.- Atmospheric dispersion and deposition models presented in Regulatory Guide 1.111 together with onsite:meteo-
~
rological measurements yield a .X/Q equal to 4.7.x 10 sec/m and J p plume depletion factor equal to 0.87.' <
.V Site boundary doses corresponding to estf uted atmospheric release a
~
rates in Tables 5-8 and 5-8a are given in Tables 5-12 and 5-13, respectively. 1
]
5.3 ' DISPOSITION OF EXISTING RACKS l 1 L l Although no spent fuel had been stored in the SFP, the SFP was used to' :li n temporarily store liquid radwaste in July 1976 prior' to its processing S ]1 in the radwaste system. As a result, low-level radioactivity was'. l present on the lower 80 percent of the existing spent fuel racks. The . radioactivity levels ranged f rom an average 70,000 dpm/100 cm2'to a' 1 2 I maximum of 180,000 dpm/100 cm . The predominate nuclides were l Co-58 and Co-60. General radiation levels in the SFP were less than 3 0.1 mrem /hr. 5-18 Revision 3
'(December 1981)
~
A high pressure water spray was used to partially remove the radioac-tivity from the racks. However, it was expected-that'80. percent of the' existing spent fuel racks would have to be disposed 6f as low-level solid waste. 9 v The packaged volume .of the racks .is not known precisely as this evalua-- tion has not been completed. Ilowever, based on experience at other-plants, it is anticipated.that the final packaged waste volume _should not exceed 1500 ft . l l l i t 1 l i 1 1 l 3 5-19 Rev'ision 3 (December 1981)..
.f.
l O TABLE 5-1 CUMULATIVE FISSION PRODUCT INVENTORIES IN FUEL AT 4 DAYS AFTER 8T11 FUEL CYCLE Activity Isotope (Ci) Kr-85 2.71 x 10 6 Zr-95 1.45 x 108 Ru-103 6.88 x 107 Ru-106 9.07 x 106 Sb-125 6.11 x 105 Te-129m 2.58 x 106 I-131 4.70 x 107 Xe-133 9.39 x 107 Cs-134(a) 6.03 x 106 Cs-137 2.46 x 107 Ce-141 1.27 x 108 Ce-144 1.26 x 108 [a] Includes contribution from activation O. of fission-produced Cs-133. O
v - 7 1 J
'l 1
1 (/ TABLE 5-2
. , 1 l
PEAK CONCENTRATIONS OF FISSION AND' CORROSION PRODUCTS IN REFUELING WATER AT 1' 4 DAYS AFTER 8TH FUEL CYCLE [a,b] ig .
-d From Reactor From Spent Coolant' Fuel Total- 1 Isotope (pC1/cc) (pC1/cc) (pCi/cc) 1
- 3-1/3-Core Storage .13. ]
l
]
Kr-85 0.0 0.0 0.0' l l Zr-95 2.27 x 10-6 1.47 x 10-7 2.42 x'10-6 y Ru-103 1.03 x 10-6 6.90-x 10-8 1,10 x 10-6 j ( Ru-106 8.97 x 10-8 '9.28 x 10-9 9.90 x 10-8 ! 6.24 x 10-10 6.24 x'10-10 Sb-125 - j Te-129m -
'1.62'x 10 _6 1,62 x 10 6- -
I-131 6.10 x 10-3' 3.57 x 10-4 6.46 x 10-3 Xe-133 0.0 0.0 0.0 , Cs-134 8.64 x 10-4' 9.92 x 10-5 9.63 x 10-4 1 Cs-137 4.50 x 10-3 4.05Lx 10-4 4.91 x 10-3 l Ce-141 2.19 x 10-6 1.27 x.10-7 2.32'x-10-6. l l (q
"/
Ce-144 Cr-51 1.11 x 10-6 4.39 x 10-5 1.29 x 10-7 0.0-1.24 x 10-6 4.39 x 10-5 .j Mn-54 3.63 x 10-5 0. 0. 3.63 x.10-5 q Mn-56 1.37 x 10 0.0 1.37 x'10-3 'l Co-58 1.18 x 10-3 0.0 l'.18 x'10-3 ] Co-60 3.49 x 10-5 .0.0 3.49 x 10-5 j Fe-59 4.72 x 10-5 0.0 4'.72 x 10-5 _ H-3 2.10 x 10-1 0.0 '2.10 x 10-1 [a] That is, at the start of 8th refueling. (b] 1.0-percent failed fuel. D^ l .i 1
.j Revision 2 L Q] (August 1977).
.!q k___._,.-_ _.n--- __ ---- _- - - __ - - - - - -
.. -_. _ ~ . . . .. .
f .. I p
%. TABLE 5-2a PEAK CONCENTRATIONS OF FISSION'AND i.
i CORROSION PRODUCTS IN REFUELING WATER AT' 4 DAYS AFTER 8TH FUEL CYCLE [a,b]
-From Reactor From Spent.
Coolant- Fuel Total y Isotope (pCi/cc) (pC1/ce) (pCi/cc) I 1 a 1-1/3-Core Storage i Kr-85 0.0 0.0 0.0 , Zr-95 2.27 x 10-6 1.47 x 10-7 2.42 x 10-6 Ru-103 1.03 x 10-6 6.87 x 10-8 1.10 x.10-6 ' l
.Ru-106 8.9 7 x 10-8 7.93 x 10-9 9.76 x 10-8 Sb-125 -
3.77 x 10-10 3.77'x 10-10 Te-129m - 1.61 x 10-6 1.61 x 10-6 1 I-131 6.10 x 10-3 3.55 x 10-4 6.46 x 10-3' Xe-133 0.0 0.0 0.0 l Cs-134 8.64 x 10-4 6.25 x 10-5 9.27 x 10-4 Cs-137 4. 50 x 10-3 1.60 x 10-4 4.66 x 10-3 Ce-141 2.19 x 10-6 1.27 x 10-7 2. 32 x 10-6 (~ Ce-144 1.11 x 10-6 1.18 x'10-7 1.23 x 10-8 (h) Cr-51 Mn-54 4.39 x 10-5 3.63 x 10-5 0.0 0.0 4.39 x 10-5 3.63 x 10-5 Mn-56 1.37 x 10-3 0.0 1.37 x 10-3 Co-58 1.18 x 10-3 0.0- - 1.18'x'10-3 0 Co-60 3.49 x 10-5 0.0 3.49 x 10-5 Fe-59 4.72 x 10-5 0.0 4.72 x 10-5 H-3 2.10 x 10-1 0.0 2.10'x 10-1 (a) That is, at the start of 8th refueling. [b] 1.0-percent failed fuel.
. (') Revision 2 (August 1977)
()
d q I fx: L ,)-. i TABLE 5-3 -; u PEAK CONCENTRATIONS OF FISSION AND l l CORROSION AT 18 DAYS PRODUCTS AFTER 8TH FUEL' IN REFUELING CYCLE [a,b WATER ] 1/3-Core Storage 1-1/3-Core Storage , Concentration Concentration Isotope (pci/cc) (pC1/cc) ~ l Kr-85 0.0 0.0 i Zr-95 1.27 x 10-7 1.27 x 10-7 Ru-103 5.40 x 10-8 5.37 x 10-8 Ru-106 9.04 x 10-9 . 7.72 x 10-9 6.18 x 10-10 3.73 x 10-10 Sb-125 q Te-129m 1.22 x 10 < 1.21 x 10-6 _ I-131 1.07 x 10-4 :1.07 x 10-4 d ' Xe-133 0.0 0.0 a Cs-134 9.79 x 10-5 , 6.17'x 10-5 Cs-137 4.05 x 10-4 1.60 x 10-4 p1 e Ce-141 9.47 x 10-8 9.47 x'10-8 V Ce-144 1.25 x 10-7
- 1.14 x 10-7 Cr-51 1.94 x 10-9 1.94 x 10-9 Mn-54 2.20 x 10-9 ' 2.20 x 10-9 Mn-56 0.0 0.0 Co-58 6.46 x 10-8 6.46-x 10-8 Co-60 2.18 x 10-9 - 2.18 x 10 Fe-59 2.39 x 10-9 2.39 x 10-9 H-3 2.10 x 10-1 2;10 x 10-1 ~
l [a] That is, at 2 weeks after start of 8th refueling period.
^
[b] 1.0-percent f ailed fuel. Id . gg. Revision 2 (August 1977) canu-a-. - - a-- -- - - - - - - - - _ - - - - - - - _ _ - - , . , - - , - - - - - - _ - - ~ - - - - - - ------ _-_--,,---__a---.------..s- ---_a- - - - - - . u - ---.---d--- -----s ---a.- - - --.
3 < TABLE 5-4: PARAMETERS USED IN TRITIUM ANALYSIS Parameter- Value Volume'of water.in RCS, CVCS holdup tanks and primary water 299,000'- storage tank, gal: 350,000, Volume of refueling cavity,. gal- ;;
' Volume of Spent Fuel Pool, galf 392,000--
Volume of refueling water storage tank, gal 350;000 Volume of water in reactor vessel, gal ]35,000 Evaporation rate from refueling cavity, gal / day 1,160 Evaporation rate from Spent Fuel-Pool, gal / day l',010-Ventilationexhaustfrom'refuelingcavlty,cfm 15,000 Ventilation exhau'st from Spent' Fuel-Pooli.cfm. <19,375 Surface area of refueling cavity;'ft 1;485 Surface area of Spent Fuel Pool, ft 1,300 . Rate constant for mixing between refueling cavity.and Spent 0.38(a)- Fuel Pool, day-1 > O Rate of loss of reactor coolant.from plant by: leakage and. controlled discharge, gal / day 370 Days of full-power operation per fuel. cycle (pCW D[i ) 320-Days of refueling per fuel cycle ) '45
)
[a] Corresponds to 95 percentf of complete mixing in 8 days, that' is des'cribed in Grant, P. J.,'et al, Oconee' Radiochemistry Survey. Program, Technical Paper ROTFL 75-4,-Babcock & Wilcox, Nuclear PowerGenerationDivision,Lynchburg,Va.,Mayfl975.
/
'. ! I j
I 'y
. /. -
./ QT NUd y+ K, v c t , j p
'- \$
(
)'
, 6 3 1: f i
E , X .- ! ( l' ,;
,l y. 'V. >t Ml '
- w. 3
(R,
- \,)
TABLE 5-5 TRITIUM CONCENTRATION IN PLANT WATER SYSTEMS (pC1/g) Fuel' Reactor Coolant. Refueling _ Spent Fuel Cycle No. System Cavity- : Pool l 1 1.13 4.47 x 10-2 4.40 x 10-2 2 -1.71 9.24 x 10-2. '9.09 ic 10-2 3 2.04 1.32~x 10-1 1.30 x 10-1 4 2.22 1.61 x 10-1 1.59 x 10-1 5 ?.33 1.82 x 10-1 1.79 x 10-1 6 2.39 1.96 x 10 1.93'x 10-1 7 2.43 2.05 x 10-1 2.02 x 10-1 8 2.45 '2.11 x 10-1 2.08 x 10-1 (;;) 9 2.46 2.15 x 10-1 2.11-x 10-1 ! l l l l fg 1 l
r 7 l g^s J O,- . q TABLE 5-6 l ACCUMULATED ACTIVITIES IN FUEL POOL PURIFICATION SYSTEM AFTER 1 YEAR ACCUMULATION "l lQ 1
.I From Reactor From Coolant Spent Fuel Total Isotope (C1) (C1) (Ci)
Kr-85 0.0 _g 0.0 -3 0.0 _3
- Er-95 1.3 x 10-6 1.5 x 10 -5 1.6 x 10 -5 i Ru-103 4.8 x 10 ~ 5.7 x 10 6.2 x 10 l Ru-106 1. 3 x 10 2.3x'10[3 2.4x10[
Sb-125 - 2.4 sc 10_q 2.4 x 10_g ;
-Te-129m -
-13 4.7 x 10-12 4.7 x 10 -12 1-131 3.8 x 10 3.9 x 10 4.3 x 10 Xe-133 0.0 0.0 0.0.
Cs-134 *79 100
. 1.8 x 101 2.0 x 101
, Cs-137 1.2'x 10,61 9.8 x 101 1.1 x 102 Ce-141 2.9 x 10, 3.0x10] 3.3 x 10~2-Ce-144 1.3 x 10 2.6 x 10 2.7 x 10 5 O Cr-51 1.4 x 10 ~
5 0.0 , 1.4 x 10 2-Mn-54 4.4 x 10 0.0 4.4 x l'0 un-56 0.0 ~ 0.0 0.0 -2 Co-58 9.6 x 10 0.0 9.6 x 10 ~ Co-60 8.6x10[ 0.0 8.6 x 10 ~ Fe-59 5.2 x 10 0.0 5.2 x 10 [a] 1.0-percent failed fuel. I-O
/'T, l G)
Revision 2 (August 1977)
7 c
/
1 TABLE 5-7. J ACTIVITIES IN VENTILATION AIR FROM ! REFUELING CAVITY'AND SPENT F E POOL DURING 8TH REFUELING a ,g - (pCi/cc) I 3-1/3-Core Storage 1-1/3-Core Storage Start of End of Start of End of. , Isotope Refueling Refueling Refueling- Refueling ! Kr-85 2.2 x 10-7 1.9 x 10-7 9.5 x 10-8 8.3 x 10-8 . Zr-95 1.2 x 10-12 6.4 x 10-14 1.2'x 10 6.4 x'10-14 I Ru-103 5.5 x 10-13 2.7 x 10-14 5.5 x 10-13 2.7 x 10-14 Ru-106 5.0 x 10-14 4.5 x 10-15 4,9 x 10-14 3.9 x 10-15 Sb-125 3.1 x 10-16 3.1 x 10-16 1,9 x 10-16 '1,9 x 10-16 :- Te-129m 8.1 x 10-13 6.1 x 10-13 8.1 x 10-13 6.1 x 10-13 j I-131 3.2 x 10-9 5.4 x 10-11 3.2 x 10-9 5.4 x 10-11 i Xe-133 7./ x 10-6 1.1 x 10-6 7. 7 x 10-6 1.1 x 10-6 Cs-134 4.8 x 10-10 4.9 x 10-11 4.6 x 10-10 3.1 x 10-11 l Cs-137 2.5 x 10-9 2.0 x 10-10 2.3 x 10-9 8.0 x 10-11 C I [q Ce-141 1.2 x 10-12 4.7 x 10-14 1.2 x 10-12 4,7 x 10-14
- D)
Ce-144 6.2 x 10-13 6.3 x 10-14 6.1 x 10-13 5.7 x 10-14 j Cr-51 2.2 x 10-11 9.7 x 10-16 2.2 x 10-11 9.7 x 10-16 l Mn-54 1.8 x 10-11 1.1 x 10-15 1.8 x 10-11 1.1 x 10-15 Mn-56 6.9 x 10-10 0.0 6.9 x 10-10 0,0 Co-58 5.9 x 10-10 3.2 x 10-14 5.9 x 10-10 3.2 x 10-14 Co-60 1.7 x 10-11 1.1 x 10-15 1.7 x 10-11 1.1 x 10 Fe-59 2.4 x 10-11 1.2 x 10-15 2.4 x 10-11 1.2'x'10 H-3 1.9 x 10-6 1.5 x 10 1,9 x lo-6 1.5 x 10-6 . (a] 1.0-percent failed fuel. r (, Revision 2 (August 1977). E______ _ __ _ _ J
O TABLE 5-8 MAXIMUM RELEASES TO ATMOSPHERE m FROM REFUELING AND FUEL STORAGE OPERATIONS " E D (Ci/yr)- Release Rate for Isotope 3-1/3-Core Storage 1-1/3-Core Stor m Kr-85 6.34 x 10 ~8 1 2.73 x 101-8 4tM Zr-95 7.79 x 10 7.78 x 10
*1 Ru-103K 2.62x10[ 2.61x10[
le Ru-106 1.09 x 10
-10 9.44 x 10 -10 Z7p Sb-125 8.18 x 10_7 4.96 x 10,7 33 L Te-129m 3.23 x 10 -3 3.21 x 10, SA. I-131 3.20 x 10 1 3.20 x 10;3 Ed. Xe-133 x 4.78 x 10 4.70 x 10 Tdf Cs-134 1.38x10[q 9.23x10[g 36f Cs-137 6.67 x 10 ~ 3.05 x 10 32L Ce-141X 4.66 x 10 ~
4.62 x 10~ C 2?6d Ce-144 1.39 x 10 1.27 x 10~ O @ A. Cr-51 103 h Mn-54 4.32 x 10" 3.69 x 10
~
~
4.32 x 10~ 3.69 x 10
~
L LA Mn-56 1.35 x 10 ~ 1.35 x 10" llS
~
Co-58 1.18 x 10 ~ 1.18 x 10~ M Co-60 3.56 x 10 3.56 x 10 M Fe-59 4.70x10] 4.70x10} H-3 2.15 x 10 2.02 x 10 (a] 1.0-percent failed fuel. O Revision 2 (August 1977)
(; f l I'N l 1 i TABLE 5-8a , EXPECTED CHANGE IN RELEASE RATES OF l
- RADIOACTIVE GASEOUS EFFLUENTS FROM l SPENT FUEL AREA [a]
(Ci/yr) I 1 1 1 Isotope Release Rate Kr-85 4.33 x 100 Xe-133 9.60x10]2II Zr-95 1.20 x 10 ~9 Ru-103 1.20 x 10 -10 Ru-106 1.75 x 10 -11 Sb-125 3.86 x 10 -10 Te-129m 2.40 x 10 1 Cs-134 5./,8x10[65 Cs-137 4.33 x 10 ~II,
/ Ce-141 4.80 x 10
( Ce-144 H-3 1.44 x 10 1.30 x 101
-9 -
[a] 0.12-percent failed fuel, i j i
.j j
\ ? \ V Revision 2 ,; ' (August 1977) i L _ _. _ _ _ _ _ __ ._ . . _ _ _ _ _ . _ ____..____________________..__:.d
F
, . ' =
' TABLE 5-9/
j DOSE RATES AT. POOL' SURFACE AND EDGES- FROM U
~lvCI ISOTOPESCONTAINEDINREFUEL}N}'
-DURING 8Til REFUELING a WATER (mrem /hr) 1 '
3-1/3-Core Storage 1-1/3-Core Storage' Sta.t.of End of Start of- End of R Isotope Refueling Refueling Refueling Refueling-Kr-85 0,0 -0.0 0.0 0.0 ' Zr-95 '4.3 x'10-3 2.2 x 10-4 4.3 x-10-3 .2.2'x 10-4 ? Ru-103 4.0 x 10-4' 2.0 x 10-5 -4.0 x 10~4 1.9'xL10-5 Ru-106 5 l'.4 x 10-6 5 1.2 x 10-6; Sb-125
'1.5 x 10 2.3 x 10~ 7 2.2 x'10
~7 1.5 x 10 7-l'4 'x'10- 1. 3 x 10
~7' Te-129m 1.3 x 10-4' 9.8'x 10 1.3'x'10~4 9.5'x'10-5T I-131 1.8 x 100 3.0 x.10-2 zl.8 x 100 3.0 x.10-2 Xe-133 0,0 0.0- 0.0 0.0-s/
Cs-134 Cs-137 1.1 x 100 2.1 x 100 1.1 x 10 1.7 x 10~1 1.1'x 100 2.0 x 100 7.2 x 10 6.7-x 10-2 g-v l Ce-141 1.3 x 10-4 5.5 x 10-6 1.3 x 10-4 5.5"x 10-6 ; L
- l. Ce-144 4.7 x 10-5 4.7 x 10-6 4.6 x 10-5 4.3 x 10-6 Cr-51 1.1 x 10-3 4'8 x 10-8
. 1.1 x'10-3 4.8 x 10-8 Mn-54 2.3 x 10-2 1.4 x 10-6 2.3 x'10-2 1.4 x 10-6 Mn-56 1,8 x 100 0.0 .. '1.8 x 100 '
1
-0.0 co-58 8.6 x 10-1 4.7 x 10-5 8.6 x 10-1 '4.7 x110-5< ,
Co-60 6.5 x'10-2 4.1 x 10-6 6.5 x 10-2 ;4.1 x-10-6 ! Fe-59 4.2 x 10-2 2.1 x 10-6 4.2.x 10-2 2.1 x 10-6 1,7 x 10-1 Total 7.8 x 100 3.1 x 10-1 '7.7 x 100 , l' [a] 1.0-percent failed fuel. I Revision 2 (August 1977). I p _ - - - - = =
r. (j' TABLE 5-10 INPLANT DOSE RATES FROM AIRBOR ISOTOPES DURING 8TH REFUELING l i
'(mrem /hr)
Submersion Inhalation ! Tissue Dose Dose Total 3-1/3-Core Storage 12 Start of Refueling Whole Body 0.21 0.70 .0.91 Skin 0.62 0.70 1.32 Lung 0.21 2.58' 2.79 Bone 0.21 0.76 0.97 l Thyroid 0.21 6.44 '6.65 Internal Organs 0.21 1.19 1.40 End of Refueling Whole Body 0.033 0.44 0.47 Skin 0.116 0.44 0.56 ) Lung 0.030 0.78 0.81 g\ Bone 0.030 0.45 0.48 Thyroid 0.030 0.75 0.78 Internal Organs 0.030 -0.67 0.70 1-1/3-Core Storage Start of Refueling Whole Body- 0.21 0.64 0.85 Skin 0.59 0.64 1.23 g. Lung 0.21 1.93 2.14 v Bone 0.21 0.68 0.89 Thyroid 0.21 4.60 4.81 Internal Organs 0.21 1.05- 1.26 End of Refueling Whole Body 0.031 0.43 0.46 Skin '0.097 0.43 0.53 , Lung 0.030 0.71 0.74 Bone 0.030 0.44 0.47 l Thyroid 0.030 0.74 0.77 Internal Organs 0.030 0.66 0.69 (a] 1.0-percent failed fuel. ( \
%)
Revision 2 (August 1977)- L. ___
r
. h m g /
e 595275591 61 n Rl R 706856203 33 i l meu eg n a 810520110 1 3 66 51 un 7. f M f i e 8 e el n ( h rer t
] uo t e l
[ d rf b eer nm e o t) ] l ar a I
- 8 t ri mr Rh f a ef
/
[ 7 7 7 7. {. 77
- - af v-eR 588998580 om o n mo sm o(
3 2 1 d nt r o i - D arf l t a ea et r ui rsh f d s o / a r f eR gr S u eh m n _ T ) o 055065050 00 bt i t _ N e H 527097520 05 9 rc E g - 21 62 214 64 st . ue _ M a n 1 21 d a0 d r _ E r a u i R o M t y+ yd _ I t st t U S i s i o Q l ve vt _ 1 E e rn eag a _ 1 R r ee s cd ce _ - o bM s e u 5 R C m 855665852 - - er rd o E - uf vol
) E W 3 No t o t r
, L O / rco oap ch
- , B P 1 a/
s Q A T N A 3 s t e crd eR rm M ( n a n o ed a d1 G s i re e N k t d e d + _ I c a e f oc o _ L a g r r ooa oe E R n e o lf l v U i g p c yf r f o n _ F w n o e t u b E e n o _ i 3 r id s d a R N o n n nn n i / o . i s o o 1 c i ar a]) t/ t n P i i c e bd aR a e F s t fl iP t P[ e r t eya lS sn c ol vF a F vT e l eye u S w S inoF p d t vo at t p tf e _ O i ot n i s n h em e m d vm erd vn eg t vo vsge0 g n uaeereuai mn or o env t crrot t c ei nbf b cio+ _ i s gsc ocnsi d cd i a
.areb l
e gn aa r ub a l nl f l na l o ssh) so u eial orei pl ee/sesyeg en f sl nl esl e t t R et l t e s er u3 pseg ru aampaob ae _ R euef/ peua rr ort mre _ vf t 1 uvf r rr 8 t enr eo oo ee oees eo rrieeerrt o f c co s ff s s7 ss us so oo( iod aop - td rsaat n ll DD D coenl l cil aopappaae aa t t )) ] - el pre eeru oo ab c RFUTRRRDF TT [ [ [ _ ~ O Y L yiE@ w j 95E" $Z"
'['G.
r -
. h d m . g / n e 585645585 54 n Rl a meu R 706746202 12 i
- . l e eg c n
a 810520110 1 3 66 51 un 6. f a M f i e 8 e f el n ( h r rer t u uo t s _ e ] l rf b nm eo t) l b I eer r e ar * [ I
- 3 t ri mr Rh I 6666.g 66 f a ef t
- - af v a omo
/ .
eR 588998580 on w mo _ sm o( 3 2 1 d nt r i m D arf l t o a ea r et r ui f rsh fd . s o / a r r f eR gr h
)
n _ e u eh m / _ S g o 055065 050 00 b t it R - T a H 527097520 05 9 rc m - N r - 21 62 214 64 st ue E n 1 21 d a0 d r 7 . - o a u i M t t y+ yd 1. )s - E S M t 0 e - R st i s i o ( p - I e l ve vt o U r eag a a Q o rn ce gt no 1 E C ee scd s e u i s 1 R - bM er rd li
- 3 m 855665852 - -
e 5 R / uf . vol o t o t r ue
]v E 1 N o E W rco ch f n L O 1
- oap a/ er
( B P ( t e eR ro A N s n crd a n rm f r b T A s M k o ed a d1 oi c i re e a G a t d e d+ d N R a e f oc o nm I g r r ooa oe eo L g n e o l f l v r - E n i g p c yf r f o ef U i n n o e t u b h F n o i 3 r id s d a r E t s o i n n /o nn n .t h R i i s o o 1 c i ar a] ) t/ x t n P i i c e bd aR E a e F s t f l iPt P[ e m r t lS sn c ol vF a F vP e eya l eye u S w S noF6 p d t vo at t p tf e imS4 O i ot n i s n h em e d vm erd vn eg t vo vsge0 g n uaeereuai t crrot t c - s gsc ocns mn ei nbf or oenv+ b cio i l e i gn d cd aa i a r .areb ub as e l nl f l na l o ssh) so ee/sesyeg u eial orei pl f sl nl esl e en t t R et l td e seru3 pseg ru aampaob ae R euef/ peua rr ort mr vf t 1 uvf r rr 7 t e l enr eo oo ee oeeseo rrieeerrt ff ss7 ssuss o o f cc o s oo( iod aop td rsaat n ll DD D D coenll cil aopappaae aa t t )]
]
]
el preeeru oo ab c , d RFUTRRRDF TT (( [ [ m)'- .
,g;ps w ggm $3 i
l
E' , o ym
\_s i
TABLE 5-12 1 MAXIMUM SITE B0UNDARY DOSES FROM REFUELING AND FUEL STORAGE OPERATIONS [a] (mrem /yr) Submersion Inhalation i Tissue Dose Dose Total 3-1/3-Core Storage l Whole Body 0.030 '0.46 0.49 Skin 0.187 0.46 0.65 Lung 0.018 0.68 0.70 Bone 0.018 0.46 0.48 Thyroid 0.018 0.72 0.74 0.018 0.68 0.70 Internal Organs
'}
1-1/3-Core Storage
/N
\
Whole Body 0.022 0.43 0.45 l l Skin 0.106 0.43, 0.54 lung 0.017 0.65 0.67 Bone 0.017 0.43 0.45 l Thyroid 0.017 0.68 0.70 l Internal Organs 0.017 0.64 0.66 [a] 1.0-percent failed fuel. l l ,- l' . Vl'y
'Rrrision 2
.(August 1977)
I
.2__ _ _ _ _ _ _ - - . _ _
l
-l
/N 1
'I l.
l l TABLE 5-13 l EXPECTED INCREASE IN SITE BOUNDARY DOSES FROM SPENT FUEL AREA [a] (mrem /yr) ; d Submersion Inhalation Tissue Dose- Dose- Total Whole Body 0.0009 0.0275 On0284 Skin' O.0096 0.0275 0.0371 > Lung 0.0001 0.0413- 0.0414. q Bone 0.0001 0.0275 0.0276
. Thyroid 0.0001 0.0411 0.0412 .;
Internal Organs 0.0001 0.0411 0.0412 1
.~
(a] 0.12-percent failed fuel. l q C/ J Revision 2 g) e
'v (August 1977)
i -l TABLE 5-14 l FACTOR FOR INPLANT RADIATION EXPOSURE ESTIMATES . FOR SPENT FUEL STORACE ITCK WORK Dose Rates in 5lork Area General area dose rate in dry spent fuel pool (measured dose rate). 0-1 area /hr -j General area dose rate around dry spent fuel pool (measured dose rate). Negligible' l Ceneral area dose' rate in spent fuel pool after refueling. Calculated based on 10.6 aren/hr'- I 0.12 percent f ailed f uel and corrosion product input f rom spent fuel movement and l rack replacement. Measured dose rates at other plantes Fort Calhoun 10-15 area /hr Prairie Island 20 arem/hr . s- 4 L ! Ceneral area dose rate around spent fuel pool af ter refueling. Haasured dose rates
~
~
1 mres/hr 4 at other plants: j Fort Calhoun 2-3 mren/hr Prairie Island 2.5 ares /hr i Contact dose rates on existing racks prior to ref ueling.(measured). 0 2-0.5 aren/hr l Contact dose rates on exisitng racks af ter refueling. Heasured dose rates it another 10 aren/hr l plants .j Ik Point Beach 10-70 mrea/hr after washdown i Man-hours in Radiation Area (a) Insta11stian of new racks (b] prior to refueling. 2656 man-hr i J O Installation of new racks af ter refueling. but all preparatory work described 5288 man-hr
- j
.O in our letter to ASLB of September 20. 1977 ,1,s,completedic). s Installation of new racks af ter refueling.' but Phase 2 of preparatory work described 8504 man-br l in our letter to ASL3 of September 20, 1977 is g completed.
1 Disposal of existing rocks. 325 man-hr i
. [a] Includes only work involving radiation exposure and does not include tasks such as unloading new racks 1 af ter shipment, etc. 1 (b) Installation of the new spent fuel storage racks involves the following steps: -
l (1) Remove existing racks. (2) Fit and weld embedment stud sleeves (56 pieces). - (3) Inspect embedment stud sleeve welds. l (4) Locate alignment sleeves and embedment module locating frames over embedment stude. Weld esbedment ~! module locating frames to alignment sleeves (11 embedment module locating frames and 56 alignment ' ' sleeves). (5) Locate and weld support cups to embedment module locating frames using template and new rack modules (56 support cups) . . I 1 (6) Inspect fitup of specified tolerances and welds. (7) Locate new rack modules into SFP. (8) Level rack modules and fit and ' weld the top module-to-module connections (40 connections). l (9) Perform final alignment checks and inspections of new racks to confirm correct installation. (c) Underwater work in spent fuel pool is escimated to require approximately three times the man-hr of comparable work in a dry spent fuel pool. .' O V Revision'3 (December 1981)
--- -_ _- ._--_------_- - - - - - - - - - .- -- -- - - - - - - - - -- A
I 4
.l TABLE 5-15 l
( , RADIATION EXPOSURE ACCUMULATED DURING SFP MODIFICATION (mrem) - 1 l
)
Work Description Exposure SFP rack, clamp and bolt removal .124 ~ q Lining spent feel rack boxes 36 I
- l 17 l Diver setup and testing Moving spent fuel and poison rods 18 j)
General carpenter and labor support 261 ' I
.4 l
I SFP rack removal 323 _ m j u i New rack installation (excluding divers) 450 ) l Drag testing new racks 2 d -
. .l Repair of spent fuel handling tool 2 l
,i Scaffold placement for crane stop movement 33 i
^ '!
/ t 5Q Removal of lights from pool 0 .i d
l l Divers in SFP (TLD data) 152 1418 1 l 1 Note: With the exception of the exposure listed.for Divers, all data was obtained by pocket ion chamber which will be conservative for these small doses. . l l I l O :
'i Revision 3 (December 1981)
6.0 NEED FOR SPENT FUEL STORAGE RACKS The existing Trojan spent fuel racks were designed to accommodate the spent fuel discharged during an annual refueling (1/3 core) plus one j entire reactor core. Within a year of discharge af ter each refueling the spent fuel was to be shipped to the Allied-General Nuclear Services' (AGNS) reprocessing plant in Barnwell, South Carolina. Thus, the design of the existing spent fuel racks provided for the perpetual ability to accommodate the discharge of the entire core to the SFP. AGNS has not been licensed by the NRC to either store or reprocess spent fuel and it appears that AGNS will not be able to accept spent fuel from Trojan in the foreseeable future. Commercial reprocessing has been deferred indefinitely in the United States as was announced by President Carter in April 1977( . This Federal policy makes it very unlikely that any commercial reprocessing f acilities will be available as was expected when the existing racks were installed. In lieu of reprocessing, the I Federal Government intends to receive and store the spent fuel from reactors. These storage facilities are to be available by 1985( . Q In March 1978, 1/3 core will be discharged into the SFP. Unless addi-tional storage were provided, either onsite or offsite prior to the next annual refueling in 1979, Trojan would lose its capability to discharge an entire core when an additional 1/3 core is discharged into the SFP during the refueling in 1979. With annual refuelings, the SFP would be filled in 1981. Therefore additional storage must be provided for Trojan spent fuel. Five alternatives were evaluated to provide additional spent fuel storage capacity. The alternatives are: (1) storage elsewhere within PGE, (2) storage at another nuclear plant, (3) storage at a commercial facility, (.4) interim storage at the AGNS reprocessing plant, and (5) installing new spent fuel storage racks at Trojan. Of these, the best alternative is installing new spent fuel storage racks at Trojan to more fully utilize the Trojan SFP storage capability until Federal storage facilities A become available in 1985. l
/ /
v 6-1 Revision 2 (August 1977) ( 1
' - _ _ _ _______.________m..___-
L
' 1 i
Alternatives 1 and 2 - There is no alternate storage capabilit'y within - ' e e PGE. Pebble Springs' Nuclear Plant, the only potential PGE storage site, )a is not scheduled to be completed until at least 1985. Another utility : is unlikely to jeopardize its own future spent fuc1' storage flexibility l j by accepting spent fuel from Trojan. Therefore, alternatives (1) and 1 i (2) are not satisf actory near-term solutions. { Alternative 3 - There are no commercial offsite storage f acilities pre-l l sently available. No commercial storage f acilities are scheduled to be j
^
completed prior to the Trojan refuelings in 1979. Because of planning, 3 l licensing and construction uncertainties associated with commercial of f-- 1 site storage facilities, it is not prudent for PGE to assume that commer-cial storage offsite will be available prior to the Trojan refueling in 1982.
.k Alternative 4 - As discussed above, the AGNS reprocessing facility is l not presently licensed to operate. In addition, as stated in ERDA l 77-25(25) , the storage capability at AGNS also is not available, j Q Alternative 5 - Installing new spent fuel storage racks at Trojan is the j only alternative to maintain uninterrupted plant operation. Suspended f power operation at Trojan is unacceptable because continued operation is ,
I i l required to meet present and future power demands. Replacement power is presently unavailable on an assured basis in the Pacific Northwest; j this situation is not anticipated to change. Secondary hydro power only - l becomes available for a few months during a high-water year. For other Q j sources which might become available, recent comparisons ( 6) show that I
. I the cost, depending on the plant f actor, will be 1.26 to 2.96 times ;
greater than the cost of power produced at Trojan. The total estimated j 6 cost of installing new spent fuel racks in 1977 dollars is $1.25 x 10 , l which would be a savings over'the cost of obtaining pow' re from other sources. l r Thus there seems to be no viable alternative to installation of the new n spent fuel storage racks at Trojan. If the new spent fuel storage racks O are installed, offsite storage will not be needed until the refueling ] (/ 6-2 Revision 2 (August 1977) l l~ \
a: l .
-3 in 1985,~wheniseven regions of spent.fue1~ from seven. annual refuelings '
have been1 storedLin the.SFP. , In the inte. im period tiefore 1985,' five ~ t alternatives for spent fuel disposition may becomeLavailable. (1) National. policy may -be implemented to provide for L spent fuel. storage and disposal.
,]
(2) The SFP at'the Pebble Springs Nuclear' Plant.may;be- -l L ready to' receive' spent, fuel. Gl (3) Current racks at Trojan may be modified in stages, to maximum density racks -(1000.or more fuel assemblies)- .
-1 (4) The Federal decision may be'made to reprocess spent. i fuel. Spent fuel would then be shipped to AGNS for reprocessing 'under the terms of the present contract.
(5) Refueling may be continued until'1988 when the entire-
/ storage capacity at Trojan will be filled. y l =t PGE is aware that a generic review by the NRC of spent fuel storage is [Q .q scheduled for resolution in late 1977'or 1978.. To continue. power oper- lC v.
ation, maintain capability to store an entire ' core and -install the new . , g
.a spent fuel racks at the lowest cost, the new racks' must be Linstalled ' ,. . . ".'
before the March 1978 refueling.- To allow PGE to fabricate and. install-lC v j
.i new spent fuel racks before that. date, resolution ~ of- the. generic ' review -
s and approva'l of subsequent licensing amendments must-be:made by. December 1977. Because of the ' time constraints, participation.in' a JvC7 ., q generic review'would adversely affect the present; schedule.-therefore, m s. s PGE requests a. separate licensing .and Edesign review' for. Trojan.; Figure.6-1 illustrates the present PGE licensing schedules.- - l2. i
-r ;
a is 6-3 Revision-2 4 (August _1977 ) f
- - - - . . - _ _ _ _ . - N.._--_.. _ _ . . - - - . - - - - . . ~ _
l kn a r' : 1.
~' o g*.
l.s1E ^
~
. 9 .g.
}. i
=
6 P - CD w cn '
.s D ~ >
oWa 3 I*5g 1.n5 i -
'c qh w. . (* j n
( !
,=
o* hat..h E SU
- me-1 j y 4 -
r* s- ** )' x55 w W s g'" 3 - >n .: 7 $NWI - sA l.& a5=5Eg$ ;c[- m CJ <C ev, 1
,35W5Nu i
3 I l_" w. L" a L 3 y O QoOt> ' -e y +a.e Ee ( j x t.,.
't a s. !
w v < _ = . t2 3.
. O"g . . _ _ _ _ . . . . . _ _ _ _ _.__ . . . .( ) .
(g g 'j
)
9
, go
- 33. - a ;.g ;
se 4 y l U1
*. i_azO q
g S. .e.
- _ ;g , i- S "
a e t . c2 - m a a w a as in. u - ?-e U .C g W LO~ O' w s.. .m.
.s s~m. ] hm,
+g .$.
u 4 vp 43
)
~
#1 d-g k_. $
= [ n,ee l
I* . s H d ', j e--- .n _ _ _ _ . . . . . . .. ._ _ _ _ _ _ _ . - .o t a ,o . : in g O : _,3 "I d V : sg r 0(A,( i g, ,
':ggc>...______.____________
ud Q N,'\L c" E b s o
- it V fgn A f(W&
_El ,, b iU 4 3 ?*" it w s. s * - i
~
Y > . f2-c
., i-g- te
_ . . . . _ . . . _ _ _ _ _ _ _ _ _ _ _ . . . . . _o g g g,
}e g
- i r<
- 5. ,i*
g .r rn __- _gy c>...........____ _____... . e5 en a
-$ 3 l h i* i
'_~E E -eg !
s
.%wW g ;
j* I T m_ gt (N r Z a
,E te < -
q - _ . . . . . . _ _ _ . . . . . . . _ _ _ _ . . . . . _ _o a
- s M 3
.O 2 W 2
s l . g u A a
-~ .
l l y u
-~ -
IJ
- g e n
- n u i
- 'J '
lWu f M y a e I R r l p
- A I
r r a M l r, 9 b F e 8 7 9 I n 9 1 fo - a J a I c e D I v o N O / - I O t c
/l I I l n.
it S I A p e g u f l u J le
~
n u J I
- y h* 4 M
I a p r h l a A i p r a M I. . b 7 e 7 F 9 i 1 n p'% a 3 J h g n t i c s a n r
' e t c n i o L C A
- y. s
]
-7.0 TESTS AND INSPECTION
'O The spent fuel storage racks were designed, f abricated 'andiinstalled as safety-related Seismic Category.1-equipment. As such, tho. quality assurance program for the spent fuel storage' racks. is governed by 10 CFR 50, Appendix B. To satisfy this requirement, PGE.and the g contractor, Programmed and Remote Systems, Inc. (par), adopted and , implemented quality assurance (QA) programs.that conform to_-. ANSI N45.2-1971. All work was subject to inspection and- approval by PGE's' inspector for conformance to specifications, drawings and quality control requirements.- 7 It was not considered necessary to impose requirements beyond ANSI N45.2. Revision 1 of par QA Program was approved in March 1977, and
' the deletion of the ANSI referenced standards was ' incorporated'into Revision 2 of PGE-1013. The foregoing is consistent' with Section 3.2.3-of ANSI N45.2.13-1974, which does not require imposing QA requirements;- ^
beyond ANSI N45.2. The. deletion of the referenced ANSI standards did not change any of the material requirements. Material test reports' O provide conformance with applicable ASTM specifications. ASME Section II, in turn, endorses ASTM standards. The PGE contract with par to design and fabricate the' spent fuel racks was' signed in January 1976, and the ' first independent inspection-by Bechtel Procurement Inspection, as an agent of PGE, was made in April 1976. Also, the par QA program was reviewed' and approved in . April 1976. PGE verified the implementation of the par QA program in two separate audits, which did not detect any significant findings. G v. During the manuf acture of the spent fuel racks, Bechtel Procurement Inspection performed the tasks associated with " Customer Witness and Hold Points" as well as normal supplier inspection activities. The PGE QA Department also audited Bechtel Procurement Inspection activities relating to their responsibilities as 'an agent of PGE. %O). 7-1 Revision 3 (December 1981). _ . _ _ _ _ _ _ _ _ _ _ .m____. -___
par is.an experienced fabricator, and the contractual and regulatory requirements ' imposed contained suf ficient' controls to assure that the final product me't the design / contract requirements. , Af ter assembly 'of thef spent fuel storage racks, ' tests were conducted _by par with a representative fuel assembly (dummy) at least 0.27 in. wider than an actual fue1 assembly. The dummy' assembly was,placed in each cavity to ensure-that no binding occurs.. Following-field installation and before using the racks, . tests were also conducted by PGE with a representative fuel assembly which more closely duplicates an actual fuel assembly than the dummy used by par. The insertion and withdrawal drag ferces were measured with a load cell and were required to be less.
~
than 50 lb during both the PGE and par tests.. Each guide assembly ,
' w centerline was . aligned so as not to deviate 'by more than 0.125 in.. f rom its true vertical pet the design specification requireme its.-
Equipment in the SFP Cooling and Demineralized System. is. accessible for inspection under normal operating conditions. Maintenance of the SFP water below 140*F is an indication of satisf actory operation. The SFP 5 cooling, purification and skimmer pumps were tested in accordance with - the standards of the Hydraulic Institute and found to satisfy the respective performance requirements. The SFP cooling pumps were tested one at a time for their ability to operate locally by manually operat-ing the respective push-button switches.- Pump. characteristics, heat exchanger capacity, filter and domineralizer' performance, and Contain-ment isolation functions are monitored periodically during operation.. i 7-2 l Revision 3 (December-1981) l'
r;
8.0 REFERENCES
TN
!d 1. SAGS, Static Analysis of General Structures, Structural Dynamics Research Corporation, No. 5729, Cincinnati, Ohio.
- 2. CHEETAH-P, Publication No. 84004100, Control Data Corporation, ,
Minneapolis, MN (1974). .
- 3. R. F. Barry, LEOPARD-A Spectrum Dependent Non-Spatial Depletion l Code for the IBM-7094, WCAP-3269-26, Westinghouse Electric Corporation, Pittsburgh, PA (1963).
- 4. W. E. Berry, Corrosion In Nuclear Applications, John Wiley & Sons, Inc. , New York (1971) .
- 5. G. E. Whitesides and N. F. Cross, KENO-A Multigroup Monte Carlo Criticality Program, CTC-5, Union Carbide Corporation (1969).
- 6. K. J. Bathe, E. L. Wilson, and F. E. Peterson, SAP IV Structural Analysis Program for Static and Dynamic Response of Linear Systems, Earthquake Engineering Research Center, University of. California, j Berkeley, CA (Revised April 1974). l
- 7. M. E. Meek and B. F. Rider, Compilation of Fission Product Yields, j NED0-12154-1, Vallecitos Nuclear Center (January 1974). C v
j i
- 8. C. M. Lederer, J. M. Hollander, and I. Perlman, Table of Isotopes, (s
Sixth Edition, John Wiley & Sons, Inc., New York _(1967).
- 9. R. J. Lutz, Iodine Behavior Under Transient Conditions in the Pressurized Water Reactor, WCAP-8637, Nuclear Energy Systems, .
Westinghouse Electric Corporation, Pittsburgh,'PA (November 1975).
- 10. D. F. Toner and J. L. Scott, " Fission-Product Release from UO2 "'
Nuclear Safety, 3_ (1961), pp 15-20.
- 11. Reference Safety Analysis Report, RESAR 41, (USNRC Docket No. STN I 50-480), Nuclear Energy Systems, Westinghouse Electric Corporation, !
I Pittsburgh, PA.
- 12. J. W. Lentsch, K. Hornyik, and R. K. Turnock, Accumulation and Release of Tritium in Pressurized Water Reactors, PGE-8001, Portland General Electric Company, Portland, OR (November 1973).
- 13. D. K. Brady, et al, Surface Heat Exchange at Power Plant Cooling Lakes (RP-49, Report 5), Publication 69-901, Edison Electric Institute (November 1969).
- 14. American Conference of Governmental Industrial Hygienists, Committee on Industrial Ventilation, Industrial Ventilation, A Manual of Recommended Practice, Edwards Brothers, Inc (1968).
/3 8-1 Revision 2 (August 1977)
f
)
I
- 15. C. Pelletier, et al, Radiciodine Source and Other Radionuclides Effluent Measurements at Nuclear Power Plants (Research Project
/'~T 274-1), Progress Report NP-101, Electric Power Research Institute, V Palo Alto, CA (January 1976), j 16 Nuclear Environmental Services, Radionuclides Releases from Nuclear Power Stations for the Period November 1974 to 'May 1975, 4 Electric Power Research Institute, Palo Alto, CA (May 1975). I
- 17. L. F. Paraly, " Calculation of Iodine-Water Partition Coefficients", l Design Consideration of Reacter Containment Spray Systems, USAEC .
Report ORNL-TM-2412, E (January 1970) . l
\
- 18. M. J. Martin, Radioactive Atoms, USAEC Report ORNL-4293, Supple- i ment 1 (November 1973).
I
- 19. M. J. Martin and P. H. Blichert-Toft, " Radioactive Atoms Auger- l f Electron, a , 6 , y , and X-ray Data", Nuclear Data Tables, 8 I (October 1970), pp 1-198.
( 20. P. J. Grant, et al, Oconee Radiochemistry Suryny Program, Technical. l Paper RDTPL 75-4, Nuclear Power Generation Division, Babcock & j Wilcox, Lynchburg, VA (May 1975).
- 21. M. J. Berger, " Beta-Ray Dose in Tissue-Equivalent Material Immersed in a Radioactive Cloud", Health Physics, ,2_6,6 (1974),
pp 1-12. 3 ( 22. L. T. Dillman, " Absorbed Gamma Dose Rate for Immersion in a Semi-V Infinite Radioactive Cloud", Health Physics, 27_ (1974), pp 571-580.
- 23. International Commission on Radiological Protection, Report of Committee II on Permissible Dose For Internal Radiation, ICRP Publication 2, Pergamon Press, New York (1963).
- 24. President Carter, " National Energy Plan," Energy Policy and Planning (April 29,1977) . '
- 25. Energy Research & Development Administration, LWR, Spent Fuel Disposition Capabilities, ERDA 77-25, 1977 Edition (May 19777.
- 26. Table 9.3-1, Pebble Springs Environmental Report Construction Permit Stage, Portland General Electric Company, Portland, OR.
- 27. A. R. Chandrasekaran, S. S. Saini, and M. M. Mainotra, " Virtual Mass of Submerged Structures" Journal of the Hydraulics Division, Proceedings of the American Society of Civil Engineers, HY5 (May 1972).
- 28. R. W. Clough, " Effects of Earthquakes on Underwater Structures",.
Proceedings of the Second World Conf erence on Earthquake Engi-l neering, Tokyo and Kyoto, Japan (July 11-18, 1960). m 8-2 Revision 2 (August 1977) I L. - - -
l 1 l l O APPENDIX A RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION O O Revision 1 (April 1977) ,
f;i
?
<. 1 r APPENDIX A 1 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION j l
Appendix A provides a summary of the requests for additional ~ informa.- _ tion that preceeded' the' NRC approval for Amendment 34 to Facility 1i Operating License for the Trojan Nuclear Plant. g. l v 1 This Appendix encompasses the supplements submitted by PGE.in support -
.of the initial license change application dated January'6, 1977.
s Listed with each summary is- the applicable PGE document that demon-strates compliance'with each request. I If further clarification is.needed, PGE letters to the[NRC, as listed. i in Amendment 34, may be referred to. i
\
l l i l
'I I
i hs-U A-1 Revision 3 (December:1981).
Summary of-NRC Request PGE Documentation'of-for Additional Information Compliance (PGE-1013)
,\
In - the footnote on Page 3-1, indicate if. Rev. 1, Section 3.1 l Figure 9.1-1 of the FSAR will be replaced . by Figures 3-2, 3-3 and 3-4 of your report.
~l Provide detailed sketches of the locating Rev.'1, Figures 3-2a frames and the cruciforms at the bottom of thru 3-2j, the rack cavities. State the weights of the Section.3.1 rack modules in the table on Page 3-3. Also, , ,
provide the dimensions of the stud sleeves on the embedment assemblies and the bolts in the tie plates at the top of.the' modules:111us-trated in Figure 3-3. J Provide sketches of the mathematical models Rev. - 1, Appendix .B of the fuel pool, the fuel storage rack, and g. the fuel assembly system which were utilized in the analysis. Illustrate on the sketches ! the mechanism of shear and load transfer to ) the fuel pool floor slab. Discuss how'the effects of sloshing water, the surrounding l external and entrapped water, and the ex-pansion allowances in the connections were considered. Also, provide the resulting significant modal frequencies of the fuel racks in air and water and the corresponding mode shapes and participation factors. State the damping values assumed-for the fuel racks. Indicate if the combination of the spatial
? tv'. 1,-Appendix C responses and the modal responses for the' fuel storage rack seismic system is in accordance with Regulatory Guide 1.92.
.i
-i A-2 Revision 3- =
(December 1981). 1 i L-_-_-___-___-____-- - . - _ . - - - -- - - - .- - --- --- -
Summary of . .NRC Request - PGE Documentation of.- a for Additional Information ' Compliance (PCE-1013) " Your reference to the FSAR and the ASME Code Rev. 1 Section 3.1.3-Section III, Subsections NA and NF, regarding the loads, load combinations and acceptance criteria utilized in the design of the racks is not sufficient for an adequate review of the design. Therefore, provide a summary for these items. Compare the most severe temperature distribu- Because of the- dif-tion considered for the structural desi En of. ferences in design the fuel pool structure for both the original documentation, it rack design and the new rack design. Also, was not' meaningful
^
provide the most severe temperature differ- ~.to make the' required g ential through the pool utilized for the-new . comparison. rack design. Provide a summary of the highest stresses, the Rev. 1, Section corresponding safety margins, the locations in 3.1.1,-Appendix B-the rack structure where these occur, and the ; maximum displacements at the top of the racks for the loading conditions considered in the analysis of the rack structure. Provide the details of your analysis' consider- Rev. 1, Appendix B ing the impact of the fuel assemblies against the rack walls. Show how you incorporated the local effect into the total effect on the rack design. A-3 Revision 3 (December 1981) _ ___ _ - __ __ x ...
D1 F- _ 1 n; ,
]
J
". . Summary of ?!RC' Request PGE Documentationlofi * ' l for' Additional Information Compliance ~(PGE-1013)
]
Provide a detailed ' summary of the stress Rev. 1,' Tables l3-41 margins in'the anchor bolt'embedment assem . and 3-5.. ' blies, tie plate components,.' locating frames and the. fuel pool floor for the. loading Because of t the 'dif- 'i
~
imposed by the new rack assemblies for the - ;ferenceslin design 3 critical load combinations. Compare numeri- documentation, l it . cally these results.to those for the. previous- was not meaningful- 'q
~
rack structure. to make.the required , j comparison.-
, l)
.i' -
On Page 3-3, it has been mentioned that the Rev . l, ' Sec tion . '3.1.1" > fuel racks are designed to. withstand the. (1 g n effect of a dropped fuel assembly. State the' assumptions regarding the . masses, the kinetic { energy of the dropped assembly and the ductil- .) ity factor of the target in absorbing the energy of impact, indicating any possible . effect from buckling. Provide the results of- l your analysis. On Page 4-3, it is stated that there is no Rev . ':1., ' Sectio'n 4.1 threat to the integrity of the pool. if a . . i l spent fuel cask is accidentally dropped into '1 t j the pool from the maximum height. State the assumptions regarding the kinetic energy of the dropped cask and the ductility factor of ! the target in absorbing the energy of impact , and provide the results of your analysis. ' l t
, a >
f
' t:.
I v A-4 Revision 3' ' I
"(December l1981) t L___.-- - __
j q ,. [1 ,1 f.
.; e g>,,1:; g o ; i,, ,.s
'f, Summary of NRC Request
,. PGE Documentation of for Additiona1'Information Cocepliance (PGE-1013) .;
?%)
On Page 3-3, it is stated thatf"the rac h and. Rev. 1, Section'3.1.1. their interfacing structures are constructed almost entirely of Type 304 stainless . stet.1". 1 Provide'a list of all materials utilized for , the design of the fuel rack structures, their corresponding functions, and their~ applicable specift ations. State' clearly the codes which are utilized-for Rev. 1, Sections < q the design, fabrication, and installation of '
; M 1.1 ' and 3.1.3 ; -) J/.s H
the rack structure. Specify the codes from , Section 7.0. ] 1 which the maximum stress limits (at appro- . ,, / ,. j priate temperatures) for the materials 'rnd , welds were obtained. 'l i l The utilization of a 25 percent increase in Rev. 1, Section 3.13, Cj v 4 l the normal operating stress limits of , ASME Appendices' B and C. j Section III, Appendix XVII as the acceptance criterion in the OBE load combinations on' t Page 3-12 of your report isunaeeptabIe. The normal operating stress limits of ASME 4 [ Section III, Appendix XVII, without any l increase, should be utilized. .In addition,- W a comparison of Tables B-6 and C-1 indicates , 4 that the combination of the three earthquake I ( i l components by the SRSS method generally yields , 'I higher stress resultants within a rack module 4 than if the resultant earthquake stress' is '- taken as the greater of the absolute sum of \ U
.a the vertical component with each of the 'l l
individual horizontal components. Therefore,- 7 the earthquake components'should.be combined utilizing the SRSS of the three individual
]V components.
s J A-5 Revision 3 l
./
9 .. (December 1981)J
;y0 8 v
Summary of URC Request PGE Documentation of l fs for Additional Information Compliance (PGE-1013) i ) i
\~_/ .3 The load combination results in Tables 3-4, Rev . 2, Tahle C-1 i B-3 through B-6, and C-1 indicate that the l ASME Section III, Appendix VII interaction Equation 21 was utilized for the calculation .j of all results. The code states that Equa- ]
.1 tion 21 shall be used in lieu of Equation 19 only when the applied axial stress is 'less <
than or equal to 15 percent of the allowable i axial stress. Therefore, utilize the appro- l priate interaction equations in these tables. In addition, provide a summary of the resul- ! tant shear stresses within the rack modules. C j The assumed values of "K" (effective length Rev. 2, Appendix C faccors) were not indicated in Figure B-1 and Table 3-4 of your report. In a telephone con-(p) versation with representatives of the NRC,
\_ / ,
par, and PCE on May 24, 1977, it was stated l that K=.5 was utilized throughout the analysis. Justify your utilization of this value, rather l than the value of .65 as recommended in the l Commentary Section of the seventh edition of the AISC Code. In addition, since the built-up funnel sections at the top of the rack modules are welded continuously to the tops of the canisters and adjacent inverted funnels, thereby provf ding for full lateral support, clarify whether gross or local buckling was considered in the calculation of allowable axial stresses. l l l /% v) A-6 Revision 3 i (December 1981)
l Sununary of NRC Request PGE' Documentation of-for Additional Information , Compliance (PGE-1013) .l ( ! i Your utilization of 3 percent additional Rev. 3, Appendix B;- ) damping due to rack submergence is unaccepta- References 27 and 28.
~
ble. Based up_on an-examination of the' current literature on the effects of submergence on damping, indications are that there is a neg-ligible increase in damping due to submergence for objects the size of the rack modules in an . environment similar to that which exists in a fuel pool. Therefore, the damping values should be taken as 2 percent and 4 percent ' for the OBE and SSE conditions, respectively. 1 i Justify the use of a maximum modal frequency Rev. 2, Appendix B. of 15.70 Hz as indicated in Table B-1 of your report. The maximum modal frequency con- b sidered should be 33 Hz unless a lower cut-off frequency can be quantitatively substantiated. I
%/
The allowable stresses presented in Rev. 2,. Appendix B. Tables 3-4, B-1 through B-6, and C-1' indicate ) \. : l that a yield stress of 30 KS1 was utilized, i I l the yield stress for stainless steel at 100*F. Th'e yield stress at the approp.riate tempera-ture should be utilized. I l Consider the possible loading arising from a Rev.'2, Section 3.1.1. temperature gradient through a rack module, (cg, consider the case of an empty set of canisters adjacent to a full set of cani-i sters). Also, discuss the effects of the -! increased fuel pool temperature on the fuel pool walls and liner. r t s..) , e A-7 Revision 3 (December 1981) l i l 1 _ _ - - _ - - _ _ _ - _ - _ - _ _ _ _ _ - . _ _ _ _ _ _ _ -- ._ _ _ _ _ _ _ _ _ _ _ .
i i
- 'I
. Summary of NRC Request PCE Documentation of' /
.. ;f] for Additional Information Compliance (PGE-1013) On Page 3-7 of your report, the discussion of Rev. 2 and 3,
)
the impact of a- fuel assembly is not complete. Section 3.1.1 l The local and gross effects on the rack-modules must be discussed and. quantified .for i the'following three possible cases of a fuel-q assembly drop-
,)
a) A straight drop on the top of a rack f module. b) An inclined drop on the top of a rack { mod ule .
~l c) A straight drop through a can with the g fuel assembly impacting the cruciform at the bottom of a can, fN i
include the kinetic energy and the drop hei8ht considered for each of the three cases. ! Also, discuss the effects of several impacts of dropped fuel assemblies affecting the struc-tural integrity of a rack module. . Impacts on ' the built-up funnels of a module, which are relied upon as structural members, will af fect their load carrying capacity. Assurance that the effects of the incident on the structural i integrity of the rack will be assessed : immediately following the incident would be sufficient. to ensure the safe storage of the fuel assemblies. 'V. A-8 . Revision-3 .
'(December.1981)
----_,_._---_-----a6 -
Summary of NRC Request PGE Documentation of for Additional Information Compliance (PCE-1013) In addition, consider the effects of the loading which will result from a fuel assembly sticking inside a can (this loading is defined in ANSI N210-1976). The upward loading should be the maximum force the crane is allowed to exert on a fuel assembly. In Appendix B, your reference to ANSI Commit- Rev. 2, Appendix B. tee 20.2 draft report entitled " Design Basis for Protection Against Pipe Whip", dated June 1973, for the calculation of the effects of fuel impact on the rack structure during a - seismic event is not sufficient. Provide the , basis of the method and the details of your. calculations. Also, discuss the effects of i the impact locally on the can and on the fuel assemblies themselves for both OBE and SSE. The additional bolt loads out of the plane, Rev. 2, Appendix B. illustrated in Figure B-8, arising from the vibration of the rack modules while containing i the worst possible unsymmetric array of fuel assemblies, should be considered. Provide the water chemistry which will be Rev. 3, Tables 3-6 maintained in the spent fuel pool. Include and 3-7. the boron concentration, pH, and the chloride, the fluoride and any heavy metal concentrations. l l l O A-9 Revision 3 (December 1981)
i Summary of NRC Request PGE Documentation of'
- for Additional Information Compliance'(PGE-1013)'
V You have stated the. average burnup'of'the .Rev. 2, Section 5.1.1. spent fuel in the pool is 33,000 mwd /MT in Sections 3.2.1.and 4.2 of your April 1977 submittal and 35,000 mwd /MT in Section 5.1.1 of your April 1977 submittal. What is the average burnup expected for the spent fuel in l the pool?. i ' l Discuss the expected change in the radio- Rev. 2, Sectional l logical gaseous effluents from the' spent fuel 5.1.5,and 5.1.6, .i ' I pool area because of the proposed' modifica-- Tables 5-8a and 5-13. tion. Include in your discussion the impact of the increase in the maximum pool bulk water temperature from 125'F to 140*F because of f l the proposed modification.
~ b S i I
. Rev. 2, g What is the weight of the fuel in the core in J l %
l metric tons? ,. Section 5.1.1; Trojan i 1 FSAR. Table 4.3-1. .. What will be done with the material to'. be ~Rev. 2, Section 5.3. removed from the spent fuel pool (eg, ' spent fuel racks) because of the proposed modifica-tion? If the material to be removed will be disposed of as solid radwaste, what is the volume of the packaged waste? What is the volume of solid waste generated Rev.>2, Section 3.2.3'. by the replacement of a cartridge filter in the Spent Fuel Pool Cleanup System? For the proposed modification, what is th'e frequency of operation and the expected . flow rates j through the cartridge filter and demineralized.
/
is during a yetr? What is the expected frequency. A-10 ' Revision 3 I
.(December 1981) i i
1 l Summary of NRC Request PCE Documentation of f._s for Additional'Information Compliance (PCE-1013) l I
%.J of replacing the cartridge filter and deminer- :
I alizer, and what is the basis for their replacement?
)
In Section 3.2.2 of your April 1977 submit- Rev. 2, tal, you stated no equipment modifications .Section 3.2.3. were required for the Spent Fuel Pool Cleanup ! System. Explain why the Spent Fuel Cleanup System is adequate to maintain low pool water i concentrations, so that there are reasonably l low exposure levels in and around the spent fuel pool area, during and af ter the modifica-tions of the pool. Provide applicable experi- .j ence from the other plants to support your estimates.
/) \ /
In Section 5.1.6 of your April 1977 submittal, Rev. 2, Sections you assumed credit for decontamination factors 5.1.6, 3.2.2 and 3.1. of 100 and 10 for HEPA filters and charcoal -s beds to estimate releases to the environment, i is the Spent Fuel Pool Exhaust System (SFPES) ' used continuously to filter air discharged l from the spent fuel pool area? What are the design conditions of the heater for the System, and does it operate continuously? Where and I how do you sample for iodine being released ! to the environment from the spent fuel pool area? l I
/'% l f\..e; ;
.l A-Il Revision 3 (December 1981)
\ _ _ _ _ _ _ - _ _ _
I Summary of NRC Request PGE Documentation of. g for Additional Information' Compliance'(PGE-1013).
-Q' .
Table 5.11 of your April 1977.subuittal Rev. 2', Section itemizes refueling operations with the' respec- =5.2.1.5. sq tIve' man-rem dose equivalent that'will be ]
- i received by personnel from each operation during a normal refueling. Describe the. .
a impact of the proposed modification on the man-rem exposure of the ' items listed (eg.1 i fuel storage and inspection operations). I The dose equivalent rate at the pool surface Rev. 2', Tables 5.9 ~q j is given in Table 5.9 as 7.8 mrem /hr. Explain and.5-11. why this value is not in the listed exposure rates of Table 5.11 for spent fuel pool g operations. , l Is the overhead handling system, including Trojan FSAR Section [
, rigging gear, single-failure proof ? 9.1.4; hRC letter l % <
from Guy A. Arlotto -) i l dated, February 12, l
.J ~
1976.. .I l y What heavy loads, other than casks, may be Rev, 2 and 3, I moved in the vicinity of the spent fuel pool. Section 4.1. Describe the travel path of the spent fuel Rev. 2 and 3, cask in the vicinity of the spent fuel pool . Section 4.1. What is the gross weight of the spent fuel Rev. 2 and 3, cask to be used? Section 4.1.
.D
)
A-12 Revision'3 (December 1981)-
l Summary of NRC Request 'PCE Documentation.of for' Additional Information Compliance-(PGE-1013) O Discuss any possible scenarios. by which it is - Rev.-2 and 3, .
' conceivable that a Lspent fuel cask or heavy Section 4.1. j
. load cou'Id fall or'tip into the spent fue l- -
pool. 1 i Are there. any plans for preferred. spent fuel . Rev.'2 and 3,- , j storage configurations in the spent fuel pool? Section 4.I'. j
.I j
Clarify the statement that differences in fuel Rev. 3, Appendix B. , rac'k seismic response usinF a 2 percent . increase-in damping due to submergence and 1 1 added conservatism in the virtual weter mass l 1 are essentially compensating. C li Clarify the effect of the increase in the fuel Rev'3, Section 3.1.1. ! pool nominal design temperature - from 125'F to 140*F in terms of applicable ~ load combination 1 stress limits. i
'l so It is stated that stresses in the' racks due Rev. 3, Section 3.1.1'. -
q l to the maximum pull-out force exerted by the crane on a stuck fuel assembly are well below yield. Clarify whether or not these stresses 1 are within the applicable Code acceptance 1 criteria. It is stated that combined stresses in the Rev. 3, Appendix C. 1 fuel racks for load ' combinations - including i fuel bundle impact inside cavities remain below I yield. Yield stresses.are not the acceptance criteria. Clarify whether or not stresses are within Code allowable values. O l A-13 ' Revision 3
.(December 1981) .
1 1
7 i .g
- Mr .Jl t t
' Summary of.NRC Request PGE-Documentation of,., 3 for' Additional . Information Compliances (PCE-1013)0 i l b
-Describe the effects on the fuel bundle of ,
j impact inside a cavity as previously_ requested. From the referenced description,. it does not. 'Rev..3, dppen' dix B5 appear that torsional conditions due to ; possible unsymmetric' fuel storage in the fuel. j rack array were analyzed for the tie-bolt - design. Justify the tie-bolt design approach with respect to the possible torsional effects. s j How will the failure of either one or both of 'Rev. 3, Sections l the Spent Fuel Pool (SFP) cooling pumps'be 3.2.3 and 3.2.4. n detected? , & 1
. sq . .. !
What will be the length of the longest time. . Rev. 3, Sections-interval between a spent fuel coolin,g pump 4.2 and 3.2.3.- l- failure and its detection? l l The brief description provided in the' Licen- Rev.'3, Section 2.0.- see's letter to the Licensing Board dated / PGE 's ' Answer to September 20, 1977;is not sufficient to gain. Dav'id B. McCoy's a full understanding of the SFP work that is Motion'for Disclo'sure currently in progress. Describe in'detsil the' 'and Cease and Desist, work that has been performed and will be per < dated October 5, formed in the SFP prior to a decision' with 1977. respect to the proposed amendment, and inclu'e.d your justification and safety evaluation as'to-why these activities do not involve an 4 unreviewed safety question or otherwise require prior NRC approval, bh g A-14' Revision 3
'(Decetaber 1981)3
i Summary of NRC Request PCE Documentation of
-w for Additional Information Compliance (PCE-1013) j s N
Describe the effects and forces upon a tuel Rev. 3, Appendix B. assembly at impact on the side bf storage q cavity resulting from a seismic event. The additional information provided in the letter ~i of September 27, 1977 (Goodwin to Schwencer) did not provide sufficient specific date to l support the numerical conclusions. I l Clarify the comparison of calculated to Rev. 3, Appendix B. design allowable stresses for the velds and l top tie-bolts. j The fuel assembly stress results, presented Rev. 3, Appendix B. in terms of ratios of allowable stresses to calculated stresses, were based on unir-g radiated fuel assembly component material (O x / properties at 70*F. Provide information on allovable fuel assembly component stresses for irradiated materials at the operating temperature of 140*F. Clarify the ASME Code edition and addenda that Rev. 3, Section 2. were used for design. What is the SFP water temperature alarm Rev. 3, Section 4.2. setpoint? In PGE-1013, Amendment 2, the Quality Rev. 3, Section 7.0. Assurance Program referenced ANSI Standards, except N45.2, have been deleted. Did the deietion of these standards affect material testing requirements? /'~) '%) A-15 Revision 3 (December 1981)
l I
.)
i Summary of NRC' Request' 'PGE Documentation of . , g for Additional-'Information ~ Compliance (PCE-1013): r '! A I ; What is PGE 's intent with . respect to the 'se-u Rev.'3, Section 3.1.. d
) '
.of the design details'shown in Figure 3-3a of
-PGE-1013? a Refer ' to -the PGE letter to the NRC, Rev. 3, Section-3'.l.1. )
November 8, 1977. Provide"information.on the 5. reference used to establish allowable stresses for the ARMCO 17-4PH stainless-steel material used for the top tie-bolts and module-threaded , ld q J feet components. With reference to existing' plug and seam welds - Rev. 3, Section 3.1.4. in the spent fuel pool floor liner plate, it
, im stated , "The vendor drawing for the -liner -
plate installation'specified that these welds ,. w-
" i were to be ground, but did not indicate the '
~
f final contour of the welds in the cases". - Please clarify the weld grinding' requirements j for the initial liner plate installation. ; How do the pro:edures used to perform the plug . Rev. 3, Section 3.'1.4. and seam weld grinding provide for control of heat generated by the grinding process, and effects of the grinding performed? l-Will PGE repair all plug and seam yeld ground Rev. 3, Section-3.1.4. areas that do not pass liquid penetrant inspection? Are other areas (other than plug nnd seam Rev. 3, Section 3.1.4. weld areas) where the - fuel pool liner plate I is ground also inspected by the liquid penetrant method? O k_.) A-16 Revielon 3 . .
, .(Dec' ember 1981) ,
RJ -
~
]
,- q l
-Summary.of NRC Request PGE Documentation of d
" for Additional Information Compliance (PGE-1013)- 1
-fv. ,
J i Provide a-criticality onalysis to confirm thatl Rev. 3,.Section 3.'l.2. a dropped fuel assembly leaning'against the- j pool wall ~ (with,its , base adjacent to the rack 1 l- bottom) vill result in an array for which keff is less than 0,95, or provide. additional measures (eg, def1-ectors) to' assure that 'a dropped fuel assembly,in any possible.configu-- i ration wil1~ remain 6,in, away from the rack-sides. As discussed in the -Safety Evaluation. issued Rev. 3, on November 11, 1977, minimum boron concentra- Section 3.1.2.1. tion in the SFP water and decay times for h spent fuel should be proposed in the event th a t the re-rad evolution should take place after the first refueling. These measures- . h
.]
would-provide ,nssurance that k egg would i remain below 0.95 and=the potential'radiologi-i cal doses below 10 CFR 100 guidelines in the event a seismic event should cause racks containing spent fuel to be upset. As an alternate, an analyeis demonstrating that 'he-t seismic restraint capability of both old and 1 l new racks would be maintained is acceptable.
- 1 l
l
,J' Provide inplant radiation exposure estimates Rev. 3, associated with the spent fuel storage rack Section 5.2.1.6- .
q l work. ' I l l 8 l LO l A-17 Revision 3
-(December'1981) l
_ _ __ _ _ _a
.1 4
[. p:
.J l
l ( '. i l APPEllDIX B
SUMMARY
OF MODELS P l
)
-I l
Revision'l (April-1977), -l
]
i l APPENDIX B fq. l
SUMMARY
OF MODELS I Figures B-1 through B-5 show the finite element models used to determine module member stresses resulting from static and dynamic loading con- ) ditions. Fuel assemblies are considered to add mass but no stif fness. The peak SRSS resultant module feet reactions under seismic loading were ] 1 used as input to the model shown-in Figures B-6 and B-7 to determine max- l 4' imum locating frame member stresses and embedmont loads. The locating frame plates will easily span the existing leak detection channels and transmit maximum module reactions. 1 Figure B-8 shows the model used to determine top tie-bolt loads. The j tie-bolt loading condition used in the design conservatively considered the maximum " diaphragm shear" in the critical direction of translation to be resisted by only those bolt assemblies oriented normal to each direction of motion, analyzed separately (thus, about 1/2 of total C n tie-bolts available were considered effective in each horizontal direc-i \ ( ,/ tion). This approach was determined in the design to be conservative j based on both direct shear and torsional shear tie-bolt analyses of indivi,Jual rack modules, A planar analysis of the model shown in Figure B-8 was done for the north-south direction (X axis), which produces the highest tie loads because of the shorter model base. These tie loads were then applied to the detc.iled three-dimensional - SAP IV models of the racks in both north-south and east-west directions, thereby conservatively considering the three-dimensional ef fects of the tie loads on each rack. The rack modules and ties were designed accordingly to transmit these maximum loads. Subsequent torsional analysis of the entire fuel rack array under maximum eccentric loading conditions has shown that the maximum vec-torial combined tie-bolt shears are about 1/2 the values used in $ the design approach. Thus, torsional conditions result in tie-bolt loads appreciably less than those used in the design. N B-1. Revision 3 (December 1981) _m .m.___._. _..m----
].s 3
The top-elevation of the modules is approximately 15.5 ft:above the , a e
- fuel pool floor and the minimur water elevation is 24 f t above the '
A floor. The minimum water cover is therefore 8.5 f t and the racks are
-well below any surface wave effect induced by seismic activity. j 1
Therefore, no additional. loads are generated by surface water sloshing.
.I In the. dynamic calculations of the submerged module, the total " virtual" ,
horizontal mass conservatively included all water trapped within .the i module and the _ volume of water trapped between the cavity face and the wall of the pool in the direction of motion. . The added, mns's of the , water is equal to approximately 30 percent of the dry mass of .the fuel k and the module. Only the submerged natural. frequencies were cal'eulated.
-i The first eight modes of vibration were analyzed which encompassed the .;
first and accond modes of vibration in each of the three spatial - m , coordinate directions. For the modified 7 x-8 module (see' Appendix C),- % l calculated modal frequencies in ascending order were: 8.0Hz,l8.2Hz,. 12.3 IIz , 15.6 Hz, 16.0 Hz, 16.5 llz, 17.2 Hz,'and:17.3 Ilz. Results of the analyses show that only the first three response modes contribute - d
/] )
significant stresses in the fuel rack components. In addition to the 30 percent increase in mass, a 2 percent. increase -in i damping was used in the analyses to account for.. changes in response due ~ to the submerged conditions of the fuel' rack ( ' ). The' design-conservatively considered the trapped water within cavities, displaced water between cavities, and' displaced water. between the fuel rocks and pool wall in the direction of motion all to be effective over the full height of the racks for the virtual' water mass determination (about-30 percent of the total weight of fuel and rack structure). A 2 percent increase in damping, based on the referenced experimental data, was also- g , used in the design in consideration of the' underwater response 'condi-
- tions (hydraulic drag, etc, on the racks). The experimental- data indi-cate that the amount of water mass that should be considered ef fective is a direct function of the fuel rack response displacements relative toc i the water mass.- Accordingly, since' the fundamental transnational mode '
/N shapes are essentially linear (supported cantilever displacements), the B-2 -Revision 3
. (December'1981)- ,
- - - _ _ -____-_A_--.-__--_ __ n_ _ _ _
virtual . water mass used in the design is approximately twice the amount suggested by the-references. If a virtual water mass proportional to response displacement is used, and no credit is given for increased damping due to submergence (response equivalent to that in air), calculated fuel rack member stresses would be increased between 10 and . i 15 percent. All materials for the SFP racks have been delivered to the i fabricator, and rack fabrication is proceeding. Material mechanical property certificates show actual material yield stresses varying from' , a minimum of 37,000 psi to a maximum of 51,000 psi in comparison with , l w l the nominal code value of 30,000 psi. Therefore, design allowable ) stresses based on actual rack material minimum yield stresses, with I t temperature corrections included, can be increased at least 20 percent and no changes to member section properties are required. 'l
\
1 The analysis of fuel rack seismic response, considering that no credit is given for increased damping due to submergence, shows that the j l stresses resulting in the welds and the top tie-bolts are within I acceptable code limit. No credit was taken for actual material mechani- [] N,/ cal properties, j l ) l OBE response spectra at 4 percent damping and OSE response spectra at j 6 percent damping were utilized. Comparison of these spectra show that l i in the range of fuel rack response frequencies, spectral values are nearly identical. Therefore, because of lower allowable stresses, the l l load combinations which include the OBE governed design. ! l l The seismic modal and spatial responses were combined in accordance with NRC Regulatory Guide 1.92, Revision 1, as described further in b Appendix C. Load combinations and allowable stress limits used in the l design are in accordance witn NRC Standard Review Plan, Section 3.8.4. Material properties for Type 304 stainlese steel at 140*F were used in the analyses. Specifically, a material yield stress of Sy = 28,000 psi and a modulus of elasticity, E = 28 x 106 psi, were used. l The effects of impact, under a seismic occurrence, of fuel assemblies I p inside fuel rack cavities were analyzed by conventional energy balance
'v/
B-3 Levision 3 (December 1981) L t -- . _ _ - - - - -
sl
? ,
)
.a methods similar: to those' described'in ANSI Committee 20.2 Draf t Report,-
4
.;q " Design Basis -for Protection Against' Pipe Whip" (June 1973). The
. equivalent static force, R , developed by a : fuel assembly'. impacting against a : fuel rack cavity, can be described by the- following equation: .)
l
~ ~
~
Aa 1 m t^ 6 ( p-1) e 1 1 2p - l 1 where
]-.
.j j
M' = the total mass of fuel and fuel rack structure per cavity =
'2638 lb' mass l
A = the maximum seismic acceleration of the. fuel: rack calculated- E f rom the response spectrum when no gap exists between the
' fuel and fuel' rack cavity 1
C N A = maximum cavity bundle. gap (fuel considered to be tight against l a corner of the cavity-~ opposite the direction'of motion, ie, ' j maximum potential energy) = 9.020 in. - 8.424'in. = 0.596'in.
- c. = ratio of mass (weight) of fuel, M , to total mass (weight) of fuel and cavity, M t, = 1616/26$8=0.613-l i
6* = limiting clastic horizontal displacement at module center due to uniform load along its ' height. This displacement was determined in the finite element analysis of the module by applying a static uniform load and increasing'the module j center horizontal displacement until the module' diagonals ' (limiting elements) were at yield: 6 = 0.395 in. e p = allowable material ductility ratio = 10. The equation was derived by summing: (1) the virtual work produced by. the inertia force of the fuel assembly mass acting through the maximum gap between cavity and f uel,'M AA g (here the module ^ peak response acceleration, A, was used considering the fuel as a rigid body); _ and (2) the virtual work produced by the inertia force of the combined fuel
' (q .l v
B-4 Rcvision 3. (December 1981)? m
.I and cavity structure mass acting through the limiting elastic horizontal 1 module displacement, M A6 To preserve energy balance.(conservatively-j t .
~
neglecting energy lost'during impact), this total virtual' work must be- l
-1 equal to the area under the force versus deformation curve, R,(6 -' 6e/2),
representing the internal strain energy absorbed by the ' fuel: rack module (6 - maximum displacement = g6e )'
~ i Numerical substitution into the above equation ' gives R, .= M A(1.150).
t _ . f i. The ef fect of a fuel bundle . impact on a storage rack cavity is then , j i calculated to be equal to 15 percent of the seismic response that would j occur if each assembly were considered to be rigidly. attached to each 1 J cavity. This 15 percent increase was added to the peak fuel rack. ]: seismic response (which considered fuel assemblies rigidly attached to j all cavities) to include fuel impact effects. . It is noted that this I impact force is equal to the inertia force of the fuel bundle acceler - ated to about 90 percent of the maximum loaded fuel rack acceleration. The fuel rack module components were designed to resist these together. m.4 j w A 6 4 with other static loads. *
%)
Probable maximum local effects of fuel assembly impact on a cavity were p) also evaluated. Because of the relation between'the natural.. frequency 4, of the rack and the duration of the fuel assembly. impact, itlis statis- y tically improbable that each fuel assembly would impact, simultaneously .' and in the same direction. By representing the probability' of this occurrence by the SRSS method, the equivalent static' force (representing the impact of a fuel assembly on each cavity) would have to.-be. approxi-mately 60 percent of the maximum seismic response to have the 15 percent overall effect as calculated, ie,
- 1/2
+ (0.57F) = 1.15 f_(F) _
c The localized fuel impact effect is equivalent to a static force of approximately 400 lb at the top of each cavity. . Static force tests. have been made on fuel impact portions of-individual cavities-for-O concentrated loads of up to.20,000 lb without exceeding localized yield
-(] ,.
l
~
B-5 ' Revision 3' .. s..
-(December'1981);
e
\
. . . ~ .
a
s I stresses. Therefore, local ~and.overall fuel bundle impact effects have g .fi been considered and provided for . in the. fuel rack design. " Kf Pursuant to the NRC request for clarification of the effects on the q
-fuel elements resulting from the governing' seismic event (SSE), supple- j
=\
mental analyses were performed in addition to those previously described. i
,1 For the analyses, the mathematical model considered the. fuel assembly to be initially located eccentrically inside the- fuel' rack cavity such I
,a that the maximum gap of 0.596 in., uniform over the assembly he'ight, existed between'the assembly and the opposite rack cavity wall. The appropriate Westinghouse eight-grid fuel assembly mass-and nonlinear stiffness properties were used in the model. Discrete loads, representa- a tive of seismic inertial effects, were mathematically ~ applied to each 0 mass point and iterated in increments up to values resulting in the total impact reactions bounding the results described. Two separate analyses were performed using the WECAM finite element computer code.
i The first analysis was performed for two iterative grid load distribu - h V tions corresponding to a total impact reaction ~of 400 lb (1) uniform i rectangular load distribution, and (2) inverted triangular load distribu-tion (maximum at top of assembly). Results in terms of ratios of ' allowable stresses to maximum calculated fuel assembly element stresses are presented in Table B-1. The deflected shapes for both' cases are
] i provided 1n Figure B-9. ;
l N The second analysis was performed for two sets of -iterative inverted i triangular grid load distributions corresponding to total impact reactions of 850 lb and an arbitrary upper. limit of 1090 lb. For this second analysis, only the inverted triangular load distribution was used because it was considered to be a better representation of the assembly discrete mass inertial force distribution corresponding to the. assembly fundamental mode response. Results of this second analysis j are presented .in Table B-2. The deflected shapes for both loading conditions are provided in Figure B-10. B-6 Revision 3 (December 1981):
L t-The conclusions reached by the supplementallanalyses performed, as O described herein, are that governing fuel assembly stresses due .to-
~ ./ .
sei' .. tic (SSE) induced impact' 1n the fuel rack are below allowable'
~
values by a factor of about two. -The reaction at the. fuel assembly d base nozzle is also less than that required to cause. fuel assembly. i
' sliding, and thus, the mathe:aatical model used is a valid '
representation. j
.j
.i l
1
.i-1 f
C l l B-7 Revision 3
-(December.1981):
)
B C s s + e) r B M t P o S (
}
+ s 7 3
l M s 9 4 5 7 l P e e( r 1 2 2 0 l t 3 : 4 bt S - ai wm d oi e - lL n l i A b S m EB o SL C S ' E0 R0 OT4 s ) TS s - g T e] o SMA r2 ( SU t[ EME S) 's RI S g s TXA } - P e 5 2 . SAC l ( r 7 5 F 1 M l t * - y - E G et S 5 4 7 6 0 s ..
- s. B LTN li 7 6 7 s s BNI b m m 1 2 ,
e s E AED ai r t r e L WNA wL o a t r B OOO o f s t A LPL l i s ~ s T LM l n e e s . AOM A U i n g s s CR t a n e s F O r r i r e OYF - e b d t r LI t p m n s t OBN n d d o e e . s_ I MU e e o e o r m b e n TE l r l r p : - n g A S E' o p b b . s e e a n RSi l m l m l d l l l r i AT m i e i e e o b b b d o h u h u t b a a m n LR C T F T F a m w w e e EO i y o o m b UF d s l l . F~ - a . l l d d g g r f a a e e n n r o t t i i i m m a a e g g d d n n u u l l s n n a a u o m m u u a i i o o i i i c c C d d l l n t x x l l a a - o a a a a a g o o r r - n M M. C C n l l a a d a i - l l e l =
=. = =
d m m u u s p - a r r g g a x 'M B g B o o o n n B E P- P o o L f f a a . ' ; i i i i ] ] n n r r 1 2 U U. T T [ [
~
~
" ~
-d' E <u.* o 1
[9eo*Ek .- ' Z
+ [ ,l t'
g I
- 1 1
?
?
-I l
TABLE B-2 j A 'I ( ) RATIO OF ALLOWABLE STRESS TO. t V FUEL ASSEMBLY COMPONENT MAXIMUM STRESSES- l FOR TRIANGULAR LOADING CASE AT 850 AND 1090 LU l Allowablell} Stress 'Allowablelll Stress ; Limit (Pg)[2] Limit (Pg + PB )I l N Load (1b) Component Uniform Stress (og) Combined Stress (og + eB} j 850 Thimble 4.75 2.05 850 Fuel rod 260 33.1 1090 Thimble 4.75 1.96 !
- w. ,
1090 Fuel rod 260 38.6 [1] Based on unirradiated properties at 70*F. j [2] Explanation of symbols: q
.1 Pg = Maximum allowable membrane stress \-
Pg = Maximum allowable bending stress ) og = Calculated membrane stress - og = Calculated bending . st ress. l l l l 10 V Revision 3 (December 1981) ____---_a
3
-w I
., J k
; i i .
1: d 1 s , h i)'^- l} a i .
' [fl '
i
' i li.
q' . d i ! ,
- l ti
< > l' ;,
, j jj '
l '.'
! "S, j
~ ,' l} z - i 9
, y ,
e OBE 5% DAMPlNG w i Gj .I r 9 I O '
/
! <t ,
3 s j ' l/ 9 s lJ s t Z l .01 ' >' j ,1 I 10 10 0 - i FREQUENCY -CPS - u (,, VERTICAL DESIGN RESPONSE SPECTRA j . .i TYPICAL i ' 1 3 1. ' l{ i e i' l - - l a l4 lj . 2 . e "o , I) z t 9 OBE 5% DAMPIN0'
~
< e w
s- A w .I .- d. i 8 n
') ,e l .OI -
FREQUENCY -CPS HORIZONTAL DESIGN RESPONSE SPECTRA =
- k-p
m. oY s ON .. 3
~
BEAM SECTION PROPERTIES O / e \ lTM3**S of IN!!7!A j
/ pyy - M3A
(.4 . 4, N 'g ' N "73 A1 At af la . II 38 MS NI'tI0t; ,1
% :'(,.l[ ( l/
g i 1 . 116 1.10 ) 4.; 4s..ft 44.404 94.400 ?. 9steltl. . 4
,i 8; 'N ' - 2 . 37, ,$0C 2.01C Olt 44.70s 9,,40 f 0.i s t 's 1 1
I g i
'd 4 q~ >
4 i.0,0 2.000
.s4.
- 1. 0CL
.,u 1 000
..n
. 0 4 ;,
n.e.,
- 14. it e t . ,0 9.74.
. : ,,. .m .
.w lLY = u. Pvuutt I 5 3.I'l .500 1.060 .f ?f 48.942 .42 f IJslat,4 --
l! pi : s i.,,0 ,.0>0
. 000 i.00
. 00<
m.
.s.,
,0,,,
a.,
.o.... ,4 .. ..,.
- 4. ,4
; s 40. 4.... .
l
$o !
s
, i...e ... , .s.0 .0,. .n, ... . <4.i..4 y e -
j
- n . ....,4 ...., . ... ,.4,, ...... .,..; ......,0 j j .0 ...n i.000 ...,, . 00 n ,.... . ....
gg.n q
, J l
[ . l h.; l Y:/J'
' Wf h > .
TL j i 1
/- -
@ APERTURE g3
- < CARD j ..
3 x 1
< Also Avaliable On 1 Aperture Card . j SPENT FUEL MODULE j
R 6 x 7 MODULE 6x8 MODULE 6x8 MODULE 6x8 MODULE 42 C AV. 48 C AV. 48 C AV, ~ 48 C AV. 6x8 MODULE 6x8 MODULE 7x 8 MODULE 6x8 MODULE 6x8 MODULE 48 C AV. 48 C AV. 56 C AV. 48 CAV. 48 C AV. 6x7 M00VLE 6x7 M00VLE 7x7 M00VLE 6x7 M000LE 6x7 M00VLE 4 CAV. 42 CAV. 49 CAV. 42 CAV. 42 CAV, oi Figure B-1 ' General and:Tfp'ical Modules 1 Revision'2 (August 1977). . -+
1 N. M bd hjthalhaaba a AM St L I W kdd d h .h h., 1^ 944&&&&& t s e e 95&&&ded tdkaE wasamaa s 9t s sf < ?t99659f eueeeee 9 eeu ei T4 f' fL6tl 6CTTOM
+ L" L , ,, . - -. . .
I 9
,, . . .e
, ,g . . , , . .
,t Y c <
j' //" o . .
.- b x, y .
/ s
. - Q..v e...f. , e a j y :
/
g- q \y :4 +7
. 3 my .
z t 8 g 3, p g.. e
, .n .
. . n,
+M ;_
l c i ,
- n<f >f-a S
E.
) 9 9 9
1
.,- - w .
e .
,/ , a a a e c s- 4.,_a .- ., + . , . u. , a a -
e ., a @l% 1 e (d.Fb.M ODUG6, O W
%g
..y 1
-c m a. s a . .,
i e e sm i m ~ e.. w,e .s. e TI l(CA t APERTURE s2 3 % CARD 6 6 9w L. g g Ate AvaMahle On
- ' Aperture Card i
- ,, ., . . .. . - . . ., . . - -- . . . . -- - . . . . -s
.7
. . c, . . . ,. , . ,, s 3 , : -
.- 2 j i u o .
= = a ~ ~. = *. 4 l l c, e o .i m , * . - . , . - o , .. . . . , , , , . . .. s_ m u , .. m
, g.s4 '
6- . L*4 3 4 m G 1 & Ia (o ARRAY $_ a , - - . . .. .s ,. , ,, m us . m -- .. . . . . .. . . u, . u w. . **- 4 3 9 3 6 6 l 3 " E * % i f 6 @ s 8 9 3 :[ E 6 6 9 3 a d 6 6 6 s > 9 d v. Fi e , g 3 6 l
- cff .e2-s
- d. .n m , ,u , ..n n ., m 4 . ,, . . s. . . . , t, as n . .
. .] J & ,, e ,
L :3 Q n<g a ' ti
! . . 6 3 6 M-e E f 6 a !
1 6 6 6 9 9 9 9 6 E 6 8 9 9 5 6 6 6 9 9 9 9 E 6 6 [ $_
'*\
/,, "'e g ;
9 3 3 6 G c . = ,,, ,. a n .. .i . = w -
- w a a 3 4r as u .a . .. .
a,2
% Mb 6'O 4 ll 17. 13 i+
G 6x8 SPENT FUEL MODULE q (,, Figure B-2 6'x 8 Module Model l Revision 2 ! (August 1977)
.i.
! t' C-j
.- a e.G. a ah -
- f. s, aasssa
} f~ f&d&d&& gesasse i lc 9&&SSSS u 9 9 e r s a
- U U W U to 6 TOP MN
~4 N/.. , , , , ,
e q 4
' ,j
. , .r 9 .p ** ' . . .2 h
1,
\.3 I. :
i
,Y
,, R .
,B
'[ . m. O
\ i
' ( :,<l '
/'- %<%,5/'ys r $l N&
J t - . .. . . s -' , N 'N D 9 E 6 N C
\ x] =
[ s .,/ .N d
+ m y.
, . , <, \ ~
.. ~
[ - S
\
K g C ' 3 e
'( t .G 9
l'
/ ,
4mL .. .. .. = in a
\
/ e 3 4'g; j >.
b 4 G h-c
,/'~ ,- -,
f,f S, $ 4 nt>
- l. f' ,,> 3 6 X< <~S p/g gu ,
c
/ ,
N , , . . s a c y
.. .. . 1 ,.
,[ 6[W ' Oh W l 3 j! , /0 x 1 M O DU LP_.
i 1 .i
- j r ,
lyw w .
P y
- a. O- n. 9. ma%. 1 is..
6 6,r_. @ @ 6 e>. n .- . TI s e e e e e ,- 555555 e, e. e. .e..e.~ s APERTURE M . s 9 9 s e 8 CARD. w w e & s' d , SOTTCM ftAN e ho Avanable on
,' Aperture Card
. . , e . ,, .. ... . ~ m ~ o ,
g
., . - ., ,, m , _m . ,, ^4 j j ;
t > ' J p <, . . , . ., , .. . . .. , , , - ,. . . \.
. m ,, _ _ , .
l I 61 s. L% u4 3 4 5 6 7 1, x 0 ARRAYS m m me a 2 e ,m ; . , . , . m q m, . ., u m . . . . . , , ,, - . .. s . a v 8 6 6 6 E G E' N mq 9 9 9 E 6 6 3 9 3 E @ 6 9 3 9 @ 6 6 @ G 6 E @ 6 g s w4
+ .
- a m
6 @ m *
,i
.x .... i ., , , ~ .n r n, . . .i,. ~ u . 4
- o, n t- s s - s. u mw 9 is @
S.
+
O 9 9 3 6 G: S 9 8 9_ E E E 9 3 0 E. E 6 9 9 @ 6 6 1 e e t *~ $ ..
~
E' .
;. . SQ 3
e
% *p t-n a <. ,, m wi ., e, .,, e , m o ., . , , . , . mm ,, ,, , , , , , , , , , ,, q le bh,94 'l 10 li 12 i$ e 6X 7 SPENT FUEL MODULE Figure B-3 6 x 7 Module Model Revision 2 (August 1977)
g, . .. .. .. r- t 4 i h e a.e } ^ -. 1.s n na
- b. qlgSDSSB . G G G G @ t
=? 9afdchdah5 ;5 $ 5 I E i
[' k. - '
'944&45d ' ,'$fd$$k E 6 6 %
Oggggd,$
~
i j li- fdddded @G , 3
$ $ 3) lI .
fddn50Ya$d f8 h $59 w w -w w t
. top P'. A N BOTTOM 6i
!\ \ af .; Y 4 ,1 M 1 w e .s 6 41 4 Ad 4) < i k i i-lr g . e i is rs n a 5 *
- w (
/
s s .
/ \ / \
g~ ~
< ,z< <<< ,
1-n -u $5 +: 2 x
$ <$ < <, 1< .<
*N
<cf , , < ,
s ( f ' ,, L
'c
( l s
/ ,
. ++. . . . . . . ..
\
/ s s\
, D D G G 6
emst [ \
\
N
@5 , ,.
3
]. {6 D e e e # e
/ , p . , .\ . . ..., ,. .
/
6 & i
#
- 6 b y fruf y .
- - ' > e e i / s
- ( q<< -
/
, p 7 ,.
3 3 [ 1
/ -
- / N s
/
e e
.h f 1 x1 Ni C DULE. O d -
- [-- ,.
'Oe***..N__
x _ x_. .. .-.. _=__ _a_x_ _
l 2 TI APERTURE CARD o Aho Available on ( Aperture Card
/ \
s, u a a a u
\
,s m ,,, .u w, ~ ,,s a .n ss m m m m sa u n sa
\e. \n. 7 / \m 7 7 1 L L A \ A L A \ A
\, . . .. . . _
\,,/ _ _ _ .. .. \ .. _ ... ._ \.. . .. .. _ - \. . .. ..
/ /\ / / /
/ \ / \ / \ / \ / \
w, w. m m m as w . e a m x a su u o, 91 .is m u w u+ u ro v 3
.1 x 1 ARRAYG 4 S L
..e 1
m
, ., e n w m . . - - - , m, w = , -t- c- w w w . , n- e r N 3 9 6 6 0 ,,
4 39 f 6 6 8 9 9 [ 6 6 6 3 3 [ G E 6 3 8 0 8 6 e 9 e 6 a , 4 9 4 0 9 e ., n. n m
- m n. . = -m 4 ,s ,, e u 6 w, w . m a ,,, n9 r m e, 4
6 / a 6 E 6 6 @ 8 3 3 9 @ O @ 9 9 0 E 0, 6 9 3 3 G E @ G @ @ 6 6 6 T
- p
@ 9 6 e 683 G1 -
10 ll 12 13 14 7x7 SPENT FUEL MODULE l Figure B-4 7 x 7 Module Model Revision 2
}{122 h (August 1977)
g-l .jg w a ( e. a A,n.m.nA 3. a a;s GEJiMil -
-O e e <
k . (; n Y g!333 'S E 5 i
$% 'y Pp ?deddddd I,%$sE$
? ~~
~A Qdsddddd 6 6 % .(
pAsc ?dd&d&&&
' I,'z - '5 5 $]
'####5## N$i
. )~ QSf { <<, S.
k<
- f , Q,-h, Y
, --TOP P'LAN b oTTd
', ;i p. g.
i -
/ l' 'I ,
l / W,
/
l / Q '' ~ ~ - -~
\
h!ild.j lW ? XQ Up
, e e > s
*gs l
Wl'/.. N,l N \L; ,
/( !
\ \ ./
-Q@
- d ap # .. ,,. . , -
/
I Nfidb h
$hfliph j / %f~Jp
\, , ... ... .. - . .. _
g <
\
pf 8: 1xb MODULE 1
'" ' ? "'
',. u, . i= = . + se
,_. k .
4 w 3 3 3 D u i w. u.
. s p .. .* . u. += $= '; 3
,- h b h ,9 h b b h h h' b !
104. ds.' ,
,, ,. ,,, ,, a 4 3 w. .,- @ .,,, ,,..
a-o- ,. 4 g
,,, ,,, ,,, .. , , s. ,. ,, ,, ,, ,,,
3 1 j w
, e e q c g .
g
- - - - - - - 4. . r.g_ X, ,, .. .. , .. ,,, ,, . . . . . , , ,
g S u. h $ o* G ln ' h w
- 1. y v v]g g ][-
% (. s L ww
- _ = _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _-
a .a a. a .. i , .e, .e..
.e..
G E., 3 . ..E..6, MM4 T1 ! $$ $ APERTURE g 33y CARD i -eee e ! PLAN Also AvaMahle On Aperture Card 4 ,, . . . . .c .. ,,s ,, ,,s ,,, ,, ... .. . ,a i. 4 - i i. <. - - i ..,s. u N
\. .
4
\ \."/.., .
4, ,., ,, ,. ,, , ,, ., . ... ,, ,,, ,,, . ... ,,.. .,, .. ,,, .., .. . ... .. ..i
. i. , m..,
4 i l l q r i a,.,.....,. ... .., s .1 .., ,. ..:.. ,,, a . ., s ,, ,. .4 ou su l.. , .. . s 9 S 4 5 & ; 1x1 A R R AYS 2 , . .., ., 4. . u . - . . .. .. . . . . . . ~ ... ,, ., . ., ., ,, kW 3 3 9 @ @
~.
9 9 8 9 9 9 a @ a 9 3 s a a 9 9 9 9 9 9 9 0 8 9 9 9 9 9 9 9
- N E 6 e k a3 s,.,4
-y . . . . , . . . . . . .. .
., sge- s
@e 3 O
- i. 92-49 m
D 9 9 9 9 9 8 9 9 9 9 9 9 9 9 0 8 9 9 9 9 9 @ @ 0 9 0 @ @ @ ,f,,e 9 @ @ g c ... .. , . . . . ,,. - .. ., . ,. - . . . - , . ... ,, ,, . - . 4 17 lb l4 7x8 SPENT FUEL MODULE Figure B-5 7 x 8 Module Model-r Revision 2 lk N br Q " (August 1977)
I l nY 24 % ; r p q t:..q A ( -
,i!,,g#'C> +23 I g- MODELED AS BEAM & RIGID ELEMENTS TO-REPRESENT THE PLATES TYPICAL' 4 20 s. > 22 o'21
! $ > t h.. h.
*18 *19
+++ +++
g . -. , G.. , 8 1 t
+
16 9 17 +15
,I
.> s , v4 _s > + ;
.)
lO,8_., ,q 14. l 1 j0 g- A -l t,.13 P> l (J r t',df. t .
- 3 r .
[ ? .,j)',;. ,
; 4 i@
q + 12:e
.. 11 ) l I8 + 103.19 ~ ~ '
+
.+g i
@ e6
**7 9 ,
+++ 6e+
4+ > .
+. .
>Q>+ 4 t 4 Wg* 5,* - (JOINTS) .-
b y,,,./p2 , l 'fY i g, . p, (SPANS) X
~ Figure B-6 Computer Model; 2 Module Floor Truss PLOT NO.1 X-Y PL ANE O) r U Revisions 2 (August 1977) t E_=-__------ - _ .-. _ = - - _
y: AY . 18 17
++
,5 f *. +
* +/f '
/;,ji'#
SN$bl ' +M,% .* 'if
*$[.' !
;t ,
, D' + + +7.;)/,/RY$.,
;4k.bl%%,1,-4 g.. . , .$ <* ,.O. ; . / ,, .
s to rep esent the
.*+..r. *, * $ . y,9*1,14.fh.,
3.. . bi,c,w,
., ./
,~,9,,, i..!"9'd ...
m
- , z. .- %..s,.f.
- r.. . : - s.,y.4,,.
,;cs . . y , ,, plates typica1 top & bottom
&l. ,,i...c..'7,y ,Mn0EA)ib$$$$$b:d$,,.:.. .. 4(G
+ 11 -* + a+ +n+ . + .. N :.
+ + + + + + 13 e + <$hRh);.);.bl:,%ye
+ +A+ + + + 4 + + 12
+++ Q') QG)
+ ++ +
+ + +
+ .++ +
+ ++ +
+ ++ +
+ +6 +-
. .. +
. +O,, +(9
++ +
Ol'2 ++ +
+
+ ++ +
+ ++ +
+ D+ +
+ ++ +
+ +4 >
\ 494+++++++++4+c v++++b++&+&+*+4+++10 h +++
+ ++
+
6
+ >+ >
+ ++ +
+ ++ +
+ 4+ +
+ ++ io (SPANS)
.. ++
Q .
.s + q
+
+ ++ +
+ ++ +
+ ++ +
+ + +
+ +t +
+7 , 8
++++**+*-** + + + + e * + + + + + .+ + +Qu + + + + ++ ++ + ++
6
.,M,.;F.,*4j'f,'['f.~,QA'J;h,
~
(FfM3% .f '.f".* Q. ,.,; ('4 :',.
- M ,:;*q.i.*U , '. r , ? .; 91 ,-
l0Eihr.c,
,vr, 6& 3 + s '*'+ + 5 + 4 e
* ' M*N,A $s. ,+ + 4+ + + d" ,pg' .n-6 ,. O . Af .
gs.pg/ ) :+ ' ' & 'I. 6 ~ i + ds + +Yi4 ?! 2 4- (JOINTS)
% '+ \ (,;;;\+
Xy' Figure B-7 Cotuputer Model p PLOT NO. I X-Y PLANE 1 Module Floor Truss -3
\ Revision 2 (August 1977) l
'i
. - _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ }
1l lil! i n E re
/ * ) 6 m_
c l' 'k i f c !! I!l il i! i
/
iS or
! [ ,! l :
i:/ -
-G
,> ji-[
h ( i
- @c p
W ! E 7 4 5
! ' j; 8 s.;/i e' I i!
d s a I R !' .i!
/ i ef;/
T' Q 6' l g t
.S
'c i
!! . i!. lI .i lle t @lI!8 -
S @ S ' O' 6'
, @w
/.l l! .'
' II!g T ili-
;. !l* i 8f! :
6 i!! gse
/ E )
a R' 6 u
! / i li:i' lii , Ii
- l O!iil e b R ! .Ii ,:i 1: 3! !
i: I, v. G s
)
6 i,. Gig!hl; D\ : !I jN!
-9 Js
/j S 9 \s @ n i:- E i ; .j ! :' i -}! -
e 4 / I A '
% ~ '
@i I ,!: j !llll!!
.* i-D 4e li:/ @
e 6 .! ! ;ll / ' Gi
,lI!ijt Y u .\ il
^
A S @. Q- .! il!
' ' .v E v.
9 e Ii!' i* h! S llil r t C 6n
,- G
; i Il ;
- l ; ,i v
6 9 6
/ /
/ O S 9
- I!:i!.Ii
\ f :
A 4 s s $ a i.
@ j/ b t A
( '
.h I
@/ / l*
G x r_ iG' gl
<. l3 1
l .
\,' i, w _
@ l/ 'Q m C 4I tE. -
e E a a .ea xoP exh~ ! E E1" O S0"HEv lfllll ll
(: 4
-l
'\
TN TN l- N 3 RID I . GRID I ~ i
.l g-
*] .
s- . N-GRID : 2 -
~ GRID 2 ~ -
g N 1 ia l GRID 3 - g GRID:3- - j N , q N' q
-GRlD 4 -
GRID ' 4 - ' d 1 g % l 1 GRID 5 - GRID 5 - N
\
O GRID 6 - N GRID 6 -- N' N, ' N-
.\
GRID 7 -- GRID 7: - '
\ 1 N 1 N .
N GRID 8 ' GRID 8
\
BN BN O O5 1.0 0. 0.5' ~ 1.0 m a) Triangular Load b) Uniform Load i l
'i
.r Figure B-9 Fuel Assembly Deflected Shape Revision 3 400 Lb Reaction Load- (December.1981) ,
.l a_____.___._____________________.-.__ _ _ _ _ _ _ _ _ ___m
.; <1 I, f
'i
/3 TN % TN-3 U GRID I -
GRID I -' N l l l :
- i. 1 1
! l l
L
% i' q
\ GRID 2 - GRID 2 5 l l l N 'l ' GRID 3 - GRID 3 - N' N GRID 4 - GRID 4 - N j t l - % 1 i 1 i GRID 5 - GRID 5 - l N ' 1 I GRID 6 - GRID 6 . l
~
N l
- t. GRID 7 -
. GRID . 7 - -
N s GRf D 8 GRID' 8 [ %, i BN BN~ a) REACTION FORCE = 850 lb b) REACTION FORCE = 1090 lb Figure B-10 Fuel Assembly Deflected Shapes for Revision 3 Triangular Load Distribution '(December.1981). l^
f O APPENDIX C SUPPLEMENTAL SEISMIC ANALYSIS OF 7 % 8 RACK O O 1 (April 1977)
{I ~ 1
) a
')
APPENDIX C
\
* .l SUPPLEMENTAL SEISMIC ANALYSIS OF 7 x 8 RACK HODULE l l
The initial analysis and design of the SFP racks were performed using" < the criteria specified Lin Trojan FSAR Sections '3.7 and 3.8 for lo.2d i j combinations and allowable stress limits, and for methods of combining modal and spatial seismic responses. Subsequent reanalysis was per-' 1l formed to assure compliance with load combinations 'and allowable stress limits specified in NRC Standard Review Plan, Section 3.8.4, and , compliance with NRC Regulatory Guide 1.92, Revision -1, for combination. 1 of modal and spatial seismic responses. 'This appendix summarizes- i results of the reanalysis. Rack component member orar.erties are: the same for all modules (see - beam section properties, Figure B-1). The 7 x 8 module, which is the largest and mest heavily loaded, was found in the initial analysis ' G j 1 to govern design of the most highly stressed module members. The j A reanalysis was therefore performed for the 7 x 8 module and for the locating frames. Model characteristics used are as' described in-Appendix B. Flinor structural modifications of the 7 x'8 rack were l found to be necessary as a result of the. reanalysis. Stif feners were J added to the module legs (Figure B-1, beam section properties, Section 9), and gusset plates were added to the base 12-in. channel (Section 5). The channel with added gusset: plate is referred to as Section 10. These modifications have been incorporated on all of the rack modules, ie, 6 x 7, 6 x 8, 7 x 7 and 7 x 8. In the redesign, member slenderness ratios in accordance-with the AISC Code, 7th edition, recommendations were utilized. Table C-1 summarizes the atress results for. the reanalysis of the modi-fled 7 x 8 rack. Ihe combined stresses in the fuel rack members for.the governing load C combination, which includes fuel bundle impact, are within' code allowable
\./ i C-1 Revision 3 (December 1981),
m __--.-.-L_. - - - - . - - - - - - - - - - - - - - - - - - - - . - . - - - - -- - - - - - -
ry i ku 4
' t i
a limits. The combined stress interaction ratios are illustrated in' Table C-1.
. Fuel bundio " rattling" impact stresses in the fuel rack' members have 1 j been accounted for in the design as. described in _ Appendix B~. ' The -
effects of this impact on . the fuel bundles do not. otherwise enter into . the desf gn of the fuel racks and, therefore, have :not been analyzed as 'l part of the modification design. However, comparisons show that . transnational mode . response accelerations are. nearly: the same 'in the - ! new racks and' existing racks.:and " rattling"' impact effects.on the l fuel should be essentially the same in either of the rack l l configurations. 4 Generic impact analyses and test data for Westinghouse' prototype--17-x 17 1 fuel bundles (identical' to Trojan's fuel), per Westinghouse Topicel-l- Report WCAP-8288, show that fuel' bundle component. stresses and impact ~ l forces due to a simultaneous LOCA and seismic event (having a peak horizontal acceleration of 'up to 0.4 g) indicate that the fuel bundle m' t w 1 design is structurally acceptable- based on established allowable e design lisaits. 'i I L l Fuel bundle impact forces determined.in the analysis of the new spent j
'\
j fuel racks for the Trojan SSE, having a peak horizontal acceleration' of l 0.25 g, are much less than those used in the Westinghouse generic h l impact analyses. I l Unirradiated material properties at 70*F were used in the analyses-because they are the more conservative values; ie, they result' in lower allowable stresses. Radiation affects the zirconium alloy so as. to increase its yield strength. There is some attendant reduction in ductility that occurs due to irradiation; however, with the. calculated stress levels being only about 1/2 of the allowable values and' within . the clastic range, a . reduction in ductility is .not of concern.. The allowable stresses used'in the stress ratio comparisons presented' t are considered to be conservative. .Unirradiated properties of zir-conium alloy were also used to establish-conservative allowable C-2 Revision 3; (December 1981) Q_____:_____-________-___--__
l l stress limits for normal operation and/or worst-case combined seismic l ( i' and blowdown fuel assembly stress analysis as described in Trojan FSAR ^ v j j Section 4.2.1.1.2. i l l l l 4 I i l l l l l 1 l l i (~s
\s _-)
1 1
/~'s I )
kJ C-3 Revision 3 (December 1981) L. _- ___ _ _ i
]
c 0 0 0 6 2 7 5 8 7 9 2 1 T 9 7 5 3 5 5 5 0 5 6 6 6 5 6 8 7 9 5 8 8 N <1_ I f e o d s o s E 9 7 4 7 6 9 2 8 7 6 C e z 2 6 2 1 3 3 4 1 r L+
+f b422 93 3 3 4 1 3 2 E
M t s S D A d ( l e
) g i S s 4 3 2 6 9 1 4 4 9 9 n y E sy 9 5 6 3 3 7 7 3 1 4 i D eb d m O rf 2 5 3 4 9 1 1 6 1 n u M t 1 1 e m S b i 8 n d . d i T e F n m S n
- a R i 7 5 2 7 6 0 4 5 6 0 0 a I b 3 8 4 l F
( ma . 3 6 6 8 116 44 6 8 1 i a s a of 4 2 2 2 1 C 1 t x h D a a O 7 H e d - T l e r E . b n e M 0 0 0 0 0 0 0 0 0 0 a i b z 8 8 8 8 8 8 8 8 8 8 w b m S b o m u S F 6 6 6 6 6 6 6 6 6 6 l o N R ) 1 1 1 1 1 1 1 1 1 1 l c S i a n p s r o 1
- - k c o) i C ( i f1 t E s 2 c E L 0 0 0 0 0 0 0 0 0 0 a a e L U
- y 8 8 8 8 8 8 8 8 8 8 B
A D O b eF 6 6 6 6 6 6 6 6 6 6 b et mar
.ldS un-T M l 1 1 1 1 1 1 1 1 1 1 h cr o b t ao0f 8 a pf2 X
w o sm ein ,de l m o9 s 7 l 2 0 2 2 0 0 5 3 2 0 i l i1 u A 2 3 2 2 3 3 1 4 2 3 t et R a . ucsl O F 6 6 6 6 6 6 7 5 6 6 0f ana F 1 1 1 1 1 1 1 1 1 1 0 roi geir S 1 nt t e E i nat S sdi ua S l u qm E 1 al eE R t 5 uch e T a - qnt ,h S rB ei 1 t D me ub e 3 5 5 6 9 8 1 3 6 4 sco5 t f E mmr 2 8 6 2 3 4 2 4 9 3 si 1 a N i eu 1 1 1 2 2 2 3 3 2 2 emt2h I xMg rsl2 t B a i t i u-M M F sesI w O ( seI o C e . rVh lE X s .i bB e aOhh ss w t pt p
)
1 oe as lh sre0
- l t i gt0 nrB } a a 0, oe s .rl ib e eiTal5 t mr 1 2 3 4 5 6 7 8 9 0 h NPi3 cuu eNg 1 TEI M S i g
))] ]
F ab c d ( (( [ [ fg*S w [gm[e O _}}