ML19332D202
| ML19332D202 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/31/1989 |
| From: | BECHTEL POWER CORP. |
| To: | |
| Shared Package | |
| ML19332D198 | List: |
| References | |
| NUDOCS 8911300152 | |
| Download: ML19332D202 (40) | |
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{"f' ; PLANT UNIQUE ANALYSIS REPORT .
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E.I. MATCH' NUCLEAR PLANT UNIT 2
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# BECHTEL POWER CORPORATION GAITHERSBURG, MARYLAND i 1 + 1 .;q 3, , REVISION 2 l ?. s DECEMBER 1989 l
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-E. I. HATCH NUCLEAR PLANT UNIT 2
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Appendix B ? 8.1-1 2 Title page. -
~8.1-2' -
NED0-24569_ title page - 8.1 -3 ' - Signature page - 1' 8.l_4- - 11 - 8.2 - iii- - 5 8.2-2 - iv - 8.2-3. I v - ;
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7 - Title' page ' - 8 - Y4 3.8B-1 - 9 - J3.8B-2 - 10 -
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Appendix D. 33 - 12 - 13- .- Title page- 2-i
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- TABLE OF CONTENTS (Continued): 4 s
,p l 4 j' < Y 6.4.3l Torus-Attached Piping and Supports- 6.4-8 j
, 4 6.4.3'.1 Analytical Model Description 6.4-8 j 6.4.3.2 Design Loads and. Load Combinations- 6.4-9 l 30 6.4.3.3 Design Allowables 6.4 q 6.4.3.4 : Method of Analysis 6.4-10 g 6.4.3.5; Analysis Results 6.4-12 !
.'6.4.3.6- Summary of Results 6.4-12
,6.'4.4 ' Valve-and' Pump Operability and Functionality. 6.4-13 ]
6.4.4.1 . Analytical Model Description 6.4-13 ,) 6;4.4.2 Design Loads and Load Combinations 6.4-13 l 6.4.4.3 Design A110wables 6.4-14
, 6.4.4.4 Method of Analysis 6.4-14
, 6.4.4.5 - Summary of Results 6.4 .6.5 References 6.5-1 7.0, SUPPRESSION POOL TEMPERATURE EVALUATION 7.1-1 7.1
Introduction:
7.1-1
~7.2.> Transient-Events Evaluated 7.2-1
]' , :7.3 Model Description l 7. 4 : Analysis Results and Conclusions 7.3-1 7.4-1 -3t 7.5 Suppression Pool. Temperature Monitoring _ -7.5-1 7.6 References- 7.6-1 -;
- 8. 0 SOURCES;0F CONSERVATISM 8.1-1 ,
i8.1L Structural Analysis Techniques '8.1-1 8.1.1 Fluid-Structure Interaction 8.1-1 , j 8.1. 2 Modeling 8.1-3 8.1. 3 Buckling 8.1-4 8.2' Load Definition 8.2-1 v 8.2.1 S/RV Loads 8.2-1
-8.2.2~ CH/C0 Loads 8.2-2 8.2.3 Pool Swell Loads 8.2-3 -
8.2.4 Earthquake Loads 8.2-4 8.3 Load Combinations 8.3-1
- 9. 0
SUMMARY
9.1-1 9.1 Stress Results 9.1-1 9.2 Conclusions 9.2-1 vii e s , - - , .-w. . , -- . - - ~ . - - - , . , -~. . . - . . . - . . . . . , , - - , . -..s
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ii. :'
!' LAPPENDIX A' Short Term Program' Summary-l
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t. F, b. -APPENDIX 1B Hitch 2'PULD-[GE NED0i24569 (Rev.'2))~ . i
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i OW f iAPPENDIX C: L'ongiTerm Ppogram Modification: Summary APPENDIX D !mpact'of Torus Shrinkage on' Torus and Anchorage. System A.4 4 l t
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1 p( (>y l , "l l q y v , U l.0'. INTRODUCTION. l QW m W. .The Hatch Unit'2 Containment-System is.one of the first generation. l LM General' Electric; (GE) boiling' water reactor (BWR) nuclear steam supply-z
,, systems housed:in a containment structure designated.as the Mark I 4 Containment System.L -The original design of the Mark I. Containment I
' System considered postulated accident loads'previouslyLassociated with. ;
- containment ~ design, which included pressure and temperature loads asso- ;
~ciated with a~ loss-of-coolant accident (LOCA), seismic loads, dead 1 e loads,. jet-impingement loads, hydrostatic loads due _to water in the l
. suppressionichamber, overload pressure test loads, and construction l
- loads. However, since the establishment of the original design
- ;< criteria, additional loading conditions have been identified that arise
.in the functioning of the pressure-suppression concept utilized in the !
. Mark 1 Containment System. , These additional loads result from the ~ '!
dynamic effects of drywell air and steam being rapidly forced into the j suppression pool .(torus) during a postulated LOCA and from suppression pool response to safety / relief valve (S/RV) operation generally associated with plant transient operating conditions. Because these i hydrodynamic loads had not been considered in the original design of the i Mark I Containment' System, the NRC' determined'that a detailed
' reevaluation. of the- Mark I Containment System was required.
s A two-phase' program was identified to the NRC in May 1975. The first- ] phase effort, called the.Short-Term Program (STP), would provide a rapid _ 1 assessment of,the adequacy of=the containment'to maintain its integrity -l
- under the most-probable course of the postulated LOCA. The first phase I
- p. :would thus~ demonstrate the acceptability of continued operation during 1 d the performance-of the second phase, called the Long-Tenn Program. In the--LTP, detailed testing and analytical work would be performed to
' define the specific design loads against which the containment would be 1
-assessed'to establish conformance to established acceptance criteria. '
-The LTP would be completed after a plant unique analysis (PUA) had been completed for each utility. The results of the analysis would be sum-marized in a- plant. unique analysis report, which would be submitted to !
the NRC and would demonstrate that the proposed configuration of the plant, including structural modifications and mitigation devices, meets i the NRC. requirements for the Mark I LTP as documented in the Mark I ' Containment Long-Term Program Safety Evaluation Report, NUREG-0661. The PUAR to follow applies to the Hatch Unit 2 Containment System. g Appendix D summarizes evaluations completed to demonstrate that the effects of the suppression chamber contraction do not affect the L structural integrity of the Hatch Unit 2 Containment System. p l; f] 1.0-1 Rev. 2 1' L l
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/1.It-DESCRIPTION ~0F THE HATCH UNIT 2 CONTAINMENT SYSTEM .
#]'Th'e:HatchUnit2ContainmentSystemisapressuresuppressionsystemtha
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~ houses the BWR' pressure vessel, the reactor coolant recirculating loops,; i and other branch connections 'of the nuclear steam supply system (NSSS).
'The system consists of- a drywell, a pressure suppression chamber (wet-well)-approximate 1y' half-filled with water, and a: vent system connecting 1 : the- drywell- to the wetwell' suppression pool. The suppression chamber is
- toroidallinishape and is located below, encircling the drywell. The ,
l :drywell-to-wetwell; vents-are connected to a vent header contained within b) ,
.the airspace of the wetwell. Downcomer pipes ~ project downward from the
. vent header and terminate'below the water surface of the suppression pool.- J The pressure; suppression chamber is shown in relation-to the steel
- t drywell in; Figure 1.1-1. Figure 1.1-2 shows a typical cross section of .j
- the suppression chamber. t 1
In-the~ highly unlikely event of a.high energy NSSS piping failure within d the drywell,-reactor: water and/or steam would be released into the - ' drywell atmosphere. This postulated event-is referred to as a LOCA. As a result.of increasing drywell pressure, a mixture of drywell atmosphere, steam, and water would be forced through the vent system into the pool of - water maintained in the suppression chamber. iThe steam vapor would
-condense in the: suppression pool, thereby limiting internal containment ?
-pressure. . The noncondensible drywell atmosphere would be transferred to thel suppression chamber.and contained therein. Section'1.2 'of this report presents.a complete description of these events. ,
The BWR utilizes the.S/RVs attached to the main steam line as a means of 4 primary system overpressure protection.- The outlets'of these valves are connected to discharge. pipes routed 1 to the suppression pool, as shown-in Figures 1.1-2 and 1.1-3. .The discharge lines end.in T quencher discharge ll -devices. Figure 1.1-4 presents-the configuration of the discharge l; device.
- i. Following the actuation of an S/RV, steam enters the safety / relief valve p 4 2 discharge line-(S/RVDL), pressurizing the line and forcing the air and ,
L water initially contained in the line into the wetwell. -Subsequent to , water and air clearing of the S/RVDL, the steam released by~the S/RV actuation flows through the S/RVDL and discharge device into the sup- ,
- pression' pool and is condensed therein. Section 1.2 of-this report presents.a complete description of these events.
l LO 1.1-1
. . - - . . + m-- .- .-,u, _ _ - , m- . m _--. _~.
~o.- _w -.- , - ..-- . . . . . , . . ~..e.. - , - -*
- e.--=-e .e -r r
J.m eu + t p ' %; " > , , i j r ; f w - , : p%' 4. 0 LOADS AND' LOAD COMBINATIONS s.
'4.1 _ LOADS j h b( N -
f Theiloads: considered and the acronyms used in the: plant unique evaluation
; of Hatch Unit 2 are--listed in Table 4.0-1. For additional information on loads,. as they relate to' the evaluation of the torus. for contraction ,
j, ' effects,:see Appendix D.
. j
', l4.1.1- Oriainal Desion Soecification loads 1 LThe original design specification loads are those loads considered in the " if evaluation of the structural components described in Section 2.0 but not redefined or defined as part of the LTP. Original design specification (
, loads can be' divided into the following two categories: i o ' . Normal < Loads- (N) o Earthquake Loads-(E)
Normal loads consist of the loads summarized in Figure 4.1.1-1. Not all , of the zloads listed apply-to ea_ch of the structural components evaluated. The dead load portion of the normal loads applies to all components evaluated;and,, in general, consists of a 1.0-g acceleration times the mass-of the structural component load applied in the vertical direction. For the torus'shell evaluation, the dead load would include the weight of the i, suppression pool corresponding to the maximum water elevation (101 feet, A M '11-3/4 inches) and the resulting hydrostatic pressure. For submerged structures,'the dead load would also include consideration ~of buoyancy and hydrostatic effects. When used for the evaluation of piping, dead load includes the' weight of the fluid and insulation. The.other components of the normal . loads category (live, thermal, pres-sure, and reaction loads resulting from normal operation) are primarily
- considered when= evaluating piping and piping components. Stresses and
_ reactions resulting from these loads were obtained from the original analyses of the piping systems involved and were combined, where appro- , priate, with the stresses and reactions resulting from LOCA and S/RV loads
' during the normal operating conditions (N0C) and LOCA ' transients.
The second' category of original design specification loads is the' earth- l _ quake loads which consist of OBE and SSE loads as shown in Figure 4.1.1-2. D When evaluating the different-structural components for earthquake, either 4 the original seismic analyses results were used or the original seismic - analyses methodologies were used to recalculate seismic stresses based on modified structural configurations. In either cases the original response spectra curves from the reactor building seismic analysis formed the basis - for the seismic results used. In general, for the torus, vent system, and internal structures, seismic stresses were recalculated based on modified structural configurations using the analyses methodologies developed by Chicago Bridge & Iron (CB&I) who performed the original seismic analyses. For the piping systems both inside and outside the torus, original seismic
, analyses results were used, or where modifications were made, the systems
~
4.1-1 Rev. 2
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* % M @'LearthquakeTresults were combined during 't_he NOC and LOCA tran s~*,' -' appropriate,;with the results-from,LOCA and:S/RV load analyses when per-r 4 iforming the structural l component evaluations 1
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de ( TABLE 4.1.8 i FATIGUE CYCLE' ASSUMPTIONS - SBA' EVENT
-Number.of Load Number of Load: Number of Effective Load Type Occurrences Cycles or__ Duration Load Cycles Containment Temperature - Tg. 1 1 :1 cycle-Containment Pressure - P g 1 O -I cycle-Chugging 1 900 seconds
- o Pre-chug 1 320 seconds '* l i
o Post-chug 1 320 seconds
- i S/RV Discharge '
i
- o Single Valve, First Actuation - 1 Approximately.2 *-
j S/RV Case A1.2 seconds o Single Valve, Subsequent Actustion - 15 Approximately 2
- S/RV Case C3.2 seconds / occurrence l o' Multiple Valve, ADS - 5/RV Case A2.2 1 Approximately 2-
- seconds i
i i Earthquake Load - SSE 1 Approximately 16 10 cycles
- seconds i
i i
*The number of effective load cycles is determined by the structural response of the component being evaluated..
See subsection 6.1.3 for the fatigue: evaluation of the suppression chamber and. subsection 6.2.7 for the- fatigue.-- evaluation of the vent system.
~
4 . 1 l . _ _ _ _ _ _ _ _ _ _ _ . - _ .. . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . -__ __ - .1 . .. 1._,. .-, -
. - - - - ,, . _ - . . ~. 1 ~ ;.u .
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Q , i-r ,e [ ] h L :4.2 LOAD COMBINATIONS i
- Th'e load combinations us'ed for structural evaluation are based on the K A more W N k ;.PUAAG (Reference 4-14) and are presented-in Table 3.2.2-1. specific: summa
. structural components is presented in Figures 4.2.1-1, 4.2.2-l', 4.2.2-2,
'and 4.2.3-1. These figures identify the hydrodynamic 1oading conditions i T resulting from LOCA and from S/RV: discharges along with the normal-and . :
fearthquakeTloads. The figures in this section provide the timing se- 1
; quences of the loading conditions for the _NOC and all three postulated' l
.LOCAs:the DBA,~1BA, and.SBA.
[ ' The lengths of.the' bars;in the figures indicate the time periods during =, l'
. . . Twhich a loading condition may exist. A loading condition such as C0 is j assumed to exist continually during the indicated time- period. For S/RV' '
discharge loads, the duration of the loading is short, but- the loads may .H
-occur at any. time during the indicated time period. Loads are
. considered to act simultaneously on a. structure at a: specific time.if
-'. the loading condition bars overlap at'that time. ^
H
, A further breakdown of loading components with respect to timing for the LOCA'poolLswell (Pps), LOCA C0 (Peo), LOCA CH (Pew), and S/RV discharge -
1
-(SRV) loads was shown in Figures 4.1.4-1, 4.1.5-1, 4.1.6-1, and 4.1.7-2, '
A . respectively. The loading components of the normal loads and the' earth-quake _1oads were shown in Figures 4.1.1-1 and 4.1.1-2, respectively. In general, the structural responses resulting' from loadings'which occur fh'duringthesametime: period'werecombinedbytheabsolute' summation Time phasing of the responses was such that the combined method (ABSS). i state'of the. stress resulted in the~ maximum stress intensity. Any exception to' ABSStis= so stated in Section 6.0 for the particular-component evaluated.
- For additional information on' load combinations utilized in the ;
evaluation of the torus for contraction effects, see Appendix D._ -l l l l l 4.2-1 Rev. 2
't 3 ,
6.0 DESIGN STRESS ANALYSIS.
-'h*
~
- Note: ~ For additional information regarding the torus contraction evaluation as-it relates to the Design Stress Analysis, see- Appendix D.-
6.1 SUPPRESSION CHAMBER. 6.1.1 Sucoression Chamber Shell. Rino Girder. and Suonorts
- 6.1.1'.1 Analvtical Model Description The supp'ression chamber (torus) is mathematically represented by a one-sixteenth segment-(22 1/2 degrees) model utilizing the NASTRAN finite element
- program. Symmetric boundary conditions are enforced at planes of symmetry.
Constraints were also applied at the base of the saddle to simulate the saddle n " rock bolts". In consideration of symmetric loads, additional constraints are applied at the saddle base. The torus shell, ring girder web, and saddle web are modeled using the isoparametric QUAD 4 and TRIA3 elements. Figure 6.1.1-1 shows a portion of the 22 1/2 degree finite element mesh. The degree of mesh refinement was determined by converging stresses and displacements for the governing individual = hydrodynamic loads. Figure 6.1.1-2 and 6.1.1-3 show, in plan view, the refined finite element mesh. L. NASTRAN bar elements represent the ring girder flanges, column flanges, and The extent of O. saddle flanges, as well these'bar-elements as miscellaneous is-depicted saddle in Figures 6.1.1-4 andstiffeners. 6.1.1-5. Additional bar
-elements representing torus shell stiffeners (WT4X33.5) are modeled using offset vectors-to account for the actual neutral axis of the element. The shell stiffeners are placed longitudinally at' approximately 20' inches.
p Circumferentially, the stiffeners extend just above the torus equator. l l- The vent pipe penetrates the torus shell through an extension referred to as the vent pipe stub. The vent bellows mechanically connect the vent pipe to the stub. The vent-pipe stub is modeled using NASTRAN shell elements.-_ Adjustments in the torus shell finite element mesh permit transition between meshes. Figures 6.1.1-7 and 6.1.1-7 represent selected elements from vent pipe stub and torus shell finite element mesh. The suppression pool is modeled with standard elastic NASTRAN solid-elements (e.g.', HEX 20, PENTA, TETRA). The fluid elements are related to the mass density of water and the sonic velocity. The application of the concept is achieved by the pressure-analog relationship (Reference 6-1). The pressure-analog method treats the fluid as an acoustic medium with properties which result in an actual representation of the fluid pressure. Figure 6.1.1-8 and 6.1.1-9 show typical sections from the fluid model finite element mesh. The complete finite element model (Figure 6.1.1-10) includes all structural elements and properties symmetric about 221/2 degrees. The shell plate thicknesses are adjusted to account for actual as-built conditions. Effects related to structures or loads not applicable to the symmetric models were evaluated using the following analytical models. 6.1-1 Rev. 2
'I 3
-[
i The analytical; model used for the seismic evaluation of the suppression N Jchamber consisted of a 360-degree beam model utilizing the NASTRAN- finite - element program.: The model includes the.shell, which acts.as a continuous h' beam;'theseismicties,whichresist-thehorizontalloads;andthesaddle. supports, which resist the vertical loads. Torusisaddle= supports and seismic tie assembliesLate modeled as single beam elements with the stiffness. equal
- to that of the: entire assembly. The' seismic finite element model is shown-11nFigure6.1.1-11.' r The analytical model used to evaluate the effects of asymmetric S/RV.
7 discharge and chugging loads on the suppression chamber'was-a 180-degree i shell model. The columns, saddles, and earthquake tie supports were r
, accounted for in the model by applying appropriate boundary conditions.
6.1.1.2 Desion= Loads and' Load Combinations-The' suppression chamber 'shell, ring girder, and-supports are subject to the-loading phenomena described in Section 4.0 (Loads and Load Combinations), as-well as structural and water deadweight loads. The event combinations
- affecting the major structural components are presented in Table 3.2.4-1.
Examination of the various load combinations was made to determine the , enveloping events. -Sixteen load combinations were found to be controlling. These cases are listed below, a .' SRV + EQ D'+ RO+PO + T0+ SRVTP ty,g + SWBDy,g + E(0) I .b. SRV + EQ D+RO+P0 &TO + SRVTP2,5 + SRVBDg,3 + E(0)
'c. SBA/IBA + SRV + EQ ,
TA D+RA+PA + SRVTP 3 ,g + CRTP(post) + SRVBDy,g + CHBD(post) + E(0)
- d. SBA/IBA + SRV + EQ D+RA+PA+TA + SRVTP2,5 + CHTP(pre) + SRVBD2,5 + CHBD(pre) + E(0)
I e. SBA/IBA + SRV + EQ D+Rg+PA.* A + SRVTP3,g + CNTP(pre) + SRVBDI,M + CHBD(pre) + E(0)
- f. Deleted
- g. DBA + EQ D+RA+PA+TA + CHTP(pre) + CHBD(pre) + E(0)
- h. DBA + EQ O D . R, . e, . 1 , . CH w(,,,,, + CHBD(,,,1) . E(0) 6.1-2 Rev. 1
- _ . - , _ _ _ _ _ _ _ _ _ . _ _ . _ - - . ~ . _ . _ _ . . _ _ _ . _ . _ - - . _ - _ . . _ . _ . _ . _ . . . . . -
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p* , > 7.54 SUPPRESSION POOL BULK TEMPERATURE MONITORING '
>e <The suppression pool temperature' monitoring system is required to ensure
- 1:. that-theisuppression pool:is within the allowable limits set forth in
% .the
- plant Technical Specifications. The numbers and distribution of the I
' : poo1Etemperature sensors ~ are shown in Figure 7.5-1. The sensors can be ,
' grouped into two categories: '!
t '1. Eleven (11) "high" sensors (T48-N301A through'T48-N311A) _ located approximately 1/2 feet below the normal
- suppression pool water surface. . #
y . 2. Four'(4) " low" sensors (T48-N009A D), located approximately R 110~ feet below the normal suppression pool water sur face. l c' Both groups-of sensors are shown on Figure'7.5-1. The T48-N300B sensor
- series arefinstalled spares.
Bulk. suppression: pool temperature is taken as the mean of the Group 1 ., average and Group 2 average. This mean is manually calculated from j H<' '
' readouts'available in the Main' Control Room. This calculated bulk '
4 suppression pool temperature. is used for routine Technical Specification. i
. surveillance.
The Group lzsensors are not required t'o be OPERABLE per the plant's
. Technical Specifications. They:only input visual alarms. They are all Lfed from the same '(Division 1) power supply. . Nevertheless, they are 4
,. available for use-during routine plant. operation.
.I ' N ) -The Group 2 sensors'are the original suppression pool water temperature sensors. These sensors input audible alarms, and are fed from redundant power supplies. In the event that'the Group 1 sensors become '
unavailable,L th'e bulk suppression pool temperature is. taken as the ,
- average of:the Group 2 sensors + 5'F to account for temperature "
stratification whenever the RHR system is not in the suppression pool cooling = mode.= Placing the RHR system in the suppression pool cooling
. mode: produces a'well' mixed suppression pool without any temperature
> stratification, hence no correction needs to be added to the Group.2 average for this case.
. t With the exception of the lack of an audible alarm from the Group 1
- sensors, the plant's Bulk Suppression Pool Temperature Monitoring System meets.the. requirements of NUREG-0661 (Reference 7-1). The lack of an audible alarm from the Group 1 sensors is acceptable because suppression pool temperature is monitored daily during normal plant operatica, and at.5 minute intervals during periods of heat addition to the pool. The Group 1 sensors also input a visual Main Control Room alarm, and an audible alarm comes from the Group 2 sensors. i
- n u
7.5-1 Rev. 2 : I
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NOO9C, - E L 91'-G"-( HALF COUPLIND AT OU SiOE - i
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}4 , TEE'QUENCetER ' EL 800'- 5 83/16" .
- in (AT OUTSIDE '
9 s ' E L 9 4'-O- j SURFACE OF TORUS , A,8 db ' A,B O 4
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THERMOWELL 8'
-E N302
'/,. **/s TEMPERATURE
- I ELEMENTS (TE)'
270* ~ , A,B h 90* (TYPli-PLACES) h
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COLUMNS NOT SHOWN - 3 , FOR CLARITY) QOO90 /' N ' 098 N3 . @ .. -DENOTES ACTUAL ' \/ /t-@g" 4 A,B .7
~
DENOTES TEMPERATURE
/ PHYSICAL LOCATION
/ OF SENSOR '
i /" ELEMENT Q307 g
/
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' N I DENOTES DUAL
- 'N , SENSOR UNIT
- i.N UY
'o EL 91 *- 6" _2*-73/4" *
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180* D '" SUPPRESSION POOL TEMPERATURE
- SENSOR LOCATIONS - ,
PLAN - REV.2
- t. . ., , - . . , . - . , ,. . .
, u ,; ,
J r ( T - j
'N ,
-8.0 SOURCES OF. CONSERVATISM 1
Fo(additional- information on conservatisms utilized in the evaluation of the ~ ltorusifor; contraction effects, see Appendix D. y - D j
>8.1 STRUCTURAL-' ANALYSIS. TECHNIQUES The evaluations performed as part.of the LTP for Unit- 2 include conservatisms !
-which involve-structural analysis techniques, LDR load definitions, and load combination methodology.- The summation of.these conservatisms leads to evaluations which exceed the, intent-of the LDR, the ASME Code, and the NRC acceptance criter_ta.-
,8.1.1 Fluid-Structure Interaction The most significant analysis technique relates-to the' implementation of the m fluid-structure interaction (FSI) effect. The FSI effect can be expressed by-
.the following relationship:
api- AP, - M,ll where:' A = torus shell interface area
. P g. - measured. pressure Pg = rigid wall pressure added fluid mass due to FSI M.
- li := acceleration at the interface.
The magnitude of-this effect is especially significant relative to the S/RV load during the SBA/IBA transients. Fundamentally, it is clear that the Lmeasured pressures from in-plant S/RV tests include thc: FSI effects. While these effects are considered plant dependent, it can be demonstrated that the FSI effects for Hatch are agi negligible. It is also known that the rigid wall pressures generated from the GE computer code,- QBUBS02, bound the flexible wall pressures measured during the Monticello in-plant tests. A
" correction" factor of 1.65 was included in the QBVBS02 code to ensure that the predicted pressures do in fact bound the measured pressures. It would
' follow, therefore, that the QBUBS02 predicted pressures include the FSI e f fects.
The plant unique characteristics of the torus also contribute to the magnitude of the FSI effect. Analytical studies have shown that the
-displacement of the torus shell follows the S/RV forcing function and the corresponding acceleration is 180 degrees out-of-phase. This appears to remain true provided that the frequency content of the forcing function falls below the predominate torus structural frequency. Consequently, i
8.1-1 Rev. 2 r -, - ._ . - , ,n. - r ,-- ,
> vi
(;_ , 's
' t the FSI.ters (M,0) in the above equation adds to rigid wall: pressures N predicted by the Q8UBS02 code. The total measured pressure, therefore, E
)
, Lis the sum of the predicted pressures and the FSI effects.- Since it can l m
Q ' be reasoned that.the QBUBS02 pressures-include FSI effects, then the i 1
;i' analytical solution includes'(twice) the FSI effects.-
Quantitative analytical' studies bear out the essence of.the above argument.
. Coupled analyses:(FSI included) have been performed using the Monticello- ,
analytical models. Two cases were examined. The measured pressure time
~
- histories'and the QBUBS02 pressure time histories were applied'to the '
coupled analytical model. As compared to the measured stresses, the
- calculated stresses were two to three times higher and three to five times higher, respectively. Comparisons between Monticello and Hatch !
demonstrate..that plant-dependent parameters are similar and that the- ) above conclusions are applicable. !
~
l g i 9
. f l.
1 L p l L O 8.1-2 L i .=-. . . . . - .. - - . . - . . - _ . . . - . - . - - - - - - - . ... . - .- - - ... -.. . . .
l fyg<
> l t
- 9. 0 St# MARY 3:
W l~ U 9.1 STRESS RES'JLTS . .i L.. l' The evaluations performed as part'of the LTP for Hatch Unit 2 are conserva-J: .tive,'especially.those involving S/RV during the IBA/SBA transients. Most structural component evaluations, including the attached piping, were governed by load combinations during these transients. The conser-g vatisms in the evaluations involve all transients (NOC, DBA, IBA, and l 58A). Because of the conservatisms described in Section 8.0 and the small margins between the calculated stresses and the allowables., the ' stresses as~ reported.herein do not constitute a safety issue. - g L l,
~ Oi l
l H l i I l O l 9.1-1 l
m ,-
)
a D. 1
+
1
9.2 CONCLUSION
S.
]
4)^ The objective of the Mark I containment LTP for Hatch Unit 2 is the ! establishment of design margins based on the NRC requirements for the Mark I LTP as documented in NUREG 0661 for the newly defined suppression J pool hydrodynamic-loads'. These margins are established by analysis ; through the application of the design loads to the Hatch Unit 2 plant
" unique containment configuration, and by comparison of the resulting 9 responses with the established structural and mechanical acceptance . '
criteria, These' required evaluations have been completed for Hatch
. Unit 2 and are summarized in this document. The results indicate that the Hatch Unit 2 containment system as modified will satisfy all'estab-lished design criteria.
To meet the objectives of the LTP, Georgia Power Company has. performed , extensive modification work on the Hatch Unit 2 containment components. l
. - A summary of the LTP modifications is'provided in Appendix C. These modifications are in-addition to the STP modifications summarized in Appendix A. The:LTP work has been performed over the last 3' years during scheduled' plant outages and during operation when possible. This modification program has been responsive to NRC concerns on containment .!
integrity by providing: timely improvements in safety margins without
- adversely impacting normal plant operation.
The PUAR for' Hatch Unit 2 is: submitted in partial fulfillment of the NRC requirements for the Mark I LTP. The PUAR summarizes the work which h demonstrates-that with the containment modifications described in Sec-tion 2 are satisfied. Therefore, completion of-these modifications will result in conformity with the requirements of the NRC-issued Order for Modifi- , cation of License and Grant of Extension of Exemption to Georgia Power Company as holder of Facility Operating License No. NPF-5 for Hatch
- Unit 2.
- Subsequent review and approval of this report will eliminate the "Unre- '
- solved Safety Issue" designation (pursuant to Section 210 of the Energy Reorganization Act of 1974) as it pertains to Hatch Unit 2.
Additional reviews performed to evaluate suppression chamber contraction effects, demonstrate that the structural integrity of the Hatch Unit 2 Containment System is not compromised, as presented in Appendix D. li, i O 9.2-1 Rev. 2
)
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- APPENDIX D
.v
- , IMPACT OF TORUS SHRINKAGE ON l -- -
TORUS & ANCHORAGE SYSTEM l- , i l, i
'i r
r
- h. L L
i i 1,
+
e I r 8 i t . I I Rev. 2 1; 1 I-l
' - -,,4 l ') ,w., , , . , - . . . , , , - , , . . , .
'~
g' d '- .D.1.0z, INTRODUCTION & BACKGROUND' 4,v. 2 bV As part of the Mark I' tong Term Program (LTP), an anchorage system was. designed to resist uplift loads' due to S/RV discharges and postulated LOCAs. i a This was accomplished through the addition of 256 2"$ rock bolts located at the' mitre-joint' saddle support locations. The bolts were installed through ,
- slotted holes which accommodate ' radial thermal growth. of. the torus during q operating and accident conditions. i
?
-DuringanNRCsite[inspectioninFebruaryof:1988,aUnitItorusrockbolt was observed:to be displaced (bent toward the reactor). A detailed inspection of all Unit I rock bolts revealed that many bolts were displaced towards the-Lreactor vessel. ,
y Due to;the condition discovered in Unit 1, a walkdown was also performed on the bolts in Unit 2. The Unit 2 rock bolts did not show evidence of the
,. displaced condition seen in Unit 1. However, a limited number of the bolts - t appeared to be bearing against the end of the' slotted holes 'in the base plate, indicating'that some radial contraction about the-major diameter of the torus- ,
at might have occurred. 4 ' A detailed. investigation was initiated to determine the cause of the torus . contraction, and verify that the condition does not have a detrimental impact - l on the design or licensing basis of.the-torus'and' anchorage system, as-documented in'the PUAR. As;was. found in Unit 1, the root cause of the saddle base plate
~
- h. mo'vement toward the reactor pressure vessel was weld induced shrinkage'of the
-perimeter of the torus. This resulted from-the implementation of modifications performed under'the Mark I short-term and long-term programs, which caused a radial contraction-about the torus major diameter (Reference 1).
The-intent of this Appendix is to summarize the impact of the torus shrinkage
.as it relates to the design and licensing basis of the torus and anchorage system,- as originally documented in the PUAR.
! 4 L p o,
- O D.1 Rev. 2
~~
%e s ,.
,. .c ]
t 7.. I "g s
,/ -D.2.0- PROBLEM DESCRIPTION -
In April?l989h a walkdown of. the 256 rock bolts on Unit 2 was completed.. This' u (wy~
;walkdown measured the existing gaps between the washer and the inboard.end
.(toward the RPV) of_ the base plate slotted hole (see Figure D.2.0-1). : Also
, recorded were the presence.of any cracks in the grout pad, and any bolt " lean" ct - from true; vertical. . Review of:this information and subsequent field. data ,
-indicated the potential. for 13 of the rock bolts .to be located at the outboard; end of the. hole. This. condition could thereby result.in the application of i
-lateral loads to these-rock bolts.as a result of torus, shrinkage. .This was .!'
not, anticipated ~ in' the'. original design of .the anchorage system, as the bolt - location in the. slot.was' oriented primarily to allow radially outward growth Lof.the torus during operating and design basis. events.' E , LA detailed. investigation was initiated to evaluate the walkdown data with
; respect to: initial construction practices and expected torus shrinkage:
.(Reference 1).
c The observed saddle base plate movements were determined to-be caused.by + '. radial shrinkage of the torus. . This shrinkage was calculated to be 0.34". . r The slotted hole-detail-incorporated into the saddle support base plates and-the original fabrication and: installation procedures provided enough clearance r to accommodate this movement, thereby ensuring'no resultant distortion of the anchor bolt. Additionally, it is probable for all bolt locations to have a - gap between the outboard edge of the slot and the adjacent bolt surface. Although shear loadsL acting onLthe bolts.'are not probable, conservative
- i. J O-
~theefforts were bolt has madetto undergone show some thedue strain acceptability of the In to a lateral load.
impact of the observed displacement on miscellaneous torus components is anchorage system assumin addition, the- ,
, provided-herein.
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, D.3.0 : LOADS AND LOAD COMBINATIONS All loads and load. combinations considered ^ in the evaluation were the same 'as O, s presented in Section 4.0. However, the design basis containment load M
definition contains many conservatisms that,can justifiably be removed due to ! advances in methodology, and a closer consideration of the actual loading phenomena. The loads considered and the acronyms:used below are listed in' ; Table 4.0-1 1 1
' Reduced. uplift loads have been generated'considering the~ following:
L a) Load reduction techniques applied to specific loading phenomena. ) ._ b) Evaluation of the critical load combinations used in the original l-design basis of the torus anchorage system. l ? Load reductions were only applied to torus rock bolt anchorage evaluations. Impact of torus shrinkage on all other components is conservatively evaluated = with' respect to the original design ~ basis loads. D.3.1 LOADS AND REDUCTION TECHNIQUES l L All loads as defined in Section 4.1 were considered in the current evaluation. In evaluating the rock bolts, the following load reduction techniques were utilized in accordance with Referencts 2 and 3. D.3.1.1 SRV loads were reduced based on " wet" vs. " dry" structure analysis results. The Hatch SRV load' analyses utilized rigid wall
-.O Pr*55"r*5 9'"*r't'd fr
- eBusSo2 in conjunction with a " wet" structure analytical model. Section 4.1.7.10 has shown that this methodology is very conservative for Plant Hatch and that loads induced from a " dry" structure ' analysis are more appropriate. A.
l detailed review of the existing SRV. analysis resulted-in a - decrease of approximately 50 percent in uplift loads when the
" dry" structure approach is considered.
D.3.1.2 -Local reaction forces, R , were reviewed to ' o eliminate duplicate loading components o consider time phasing of loads o consider direction, not just magnitude, for those loads that act in a specific direction. l D.3.1.3 Review of load phenomena indicates that SRV actuation cannot occur l until at least two seconds after pool swell (PS) uplift is l complete. Therefore, the PS + SRV loads are not combined l , concurrently. D.3.1.4 During the condensation oscillation (CO) period of a LOCA the drywell to wetwell pressure is sufficient to maintain the SRVDL clear of water. Thus, the air bubble loading due to an SRV l' actuation will be negligible. Therefore, the SRV + C0 loads are not combined concurrently. O D.3 Rev. 2 1' ll :
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,~:,.i, V,j ,l.. .: D. 3 .1 '. 5 , !All' dynamic _ loads in the governing: load combinations were combined '1
,W using-SRSS methodology.- A. detailed review ~of the various loading- l
,, i f _. time. histories,;as presented in Reference 4, indicated that this- . )
.: m techniqueLis appropriate, a
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D.4 Rev. 2 i _ .~. . , _ . . . . _ . . . . . . _ - . . _ _. . . . _ . , . _ _ _ . . - . . _ . . . . . . _ . . . . . . . . - - . . . . - . . . . . _ _ _ _ . . . _ . . - ~ . . . . -
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'D.4.0 ,' ANCHORAGE'AND COMPONENT EVALUATION AND RESULTS A1 '
; A detailed in'vestigation was' completed on the' torus and anchora e. system mq]' -
. design to evaluate components which could be impacted by the we d induced shrinkage. The following components were identified as potentially
_I susceptible'to problems resulting from the shrinkage:. ) o; -Torus Anchorage System j
- o- Suppressicn Chamber Shell and Piping Penetrations:
o > Vent System o -Torus -Internal Structures o' ~SRV Piping and Supports in Torus o Torus Attached Piping and Supports ,
, o. . Torus Seismic Ties The results:of the reviews related to each component are summarized in the- >
following-discussion.- Note that the torus rock bolts were the only component in-which the load reduction techniques of Section D.3.1 have been applied. All other components were conservatively assessed with respect to original- , design-basis loadings. ' D.4.1 Torus Anchorage System The original loads and sixteen load combinations applicable to the rock bolts,
'as presented in Section 6.1.1.2, were reviewed based on changes identified in Section D.3.0. ' Design loads and-load combinations for other components were not revised. Three controlling combinations resulted relative to the rock i bolt re-analysis. These are:
- 1. SBA/IBA+SRV+EQ ;
D + R, + P, + T, + SRVTP ,,i + CHTP po,,3 c + SRVBD 3,n + CHBDi p,,t3.+ E(0)
- 2. DBA + EQ 1-D + R, + P, + T, + COTP + COBD + E(0)
-3.. DBA+EQ L D + R, + P, + T, + PSTP + PSBD + PSFI + E(0) l l
These correspona to load cases c, k, and n presented in Section 6.1.1.2, respectively. Maximum bolt loads for each load combination are shown in Tables D.4.1-1 thru L 3. Original design basis loads are identified along with reduced loads L utilizing the techniques of Section D.3. The maximum uplift load per bolt, using the above-mentioned techniques, is 142 kips. O D.5 Rev. 2 I-g
. -- . - . . . . . , .. - . - . ~ . _ -. . . - - , . - --. --..--._.. ___. .. - .
i s I i n A limited ' number (13) of the. Plant-Hatch Unit 2 torus mitre joint rock bolts
~
. have, potentially, experienced sufficient lateral load strain such that it is a prudent to consider.a reduction in their tensile capacity. Reduced capacities (T
~
for these bolts have been calculated, and new design allowables are shown in Table D.4.1-4. The maximum uplift bolt load is less than the minimum reduced , design allowable of 177 kips per bolt. : D.4.2 Suppression Chamber Shell and Piping Penetrations Penetration loads were reviewed based on the calculated shrinkage displacement of 0.34". The displacement was found to increase penetration loads for about half of'the penetrations. Review of the penetration analyses indicated that all stress levels remain within allowables. Therefore, the penetrations remain _ acceptable. As only 13 bolts have the potential to be loaded laterally, no significant stresses will be developed in the torus shell. The small stresses which may develop will be relieved with a nominal increase in torus . temperature and the resulting expansion. Thus, these stresses will not act concurrently with design basis stresses. Therefore, the torus shell remains acceptable. D.4.3 Vent System The' displacements associated with the torus shrinkage combined with the maximum design displacements were found to be less than the allowable
- displacement of the expansion bellows (Reference 1, Sh. 33). Additionally, it
. was found that the external loads in the drywell and suppression chamber at-the vent line due to the shrinkage, plus the maximum design loads, were well O' ' below the allowables (Reference 1, Sh. 39). Therefore, the vent system
. remains acceptable.
D.4.4 Torus Internal Structures The catwalk has bolted connections and supports which can accommodate the + torus shrinkage displacement without creating any additional significant stresses. Review of the monorail analysis and construction sequence indicates that the monorail will not see any significant stresses due to the torus shrinkage displacement. All conduit routing is flexible due to bends which allow necessary displacements with no significant stress increase. Therefore, internal structures remain adequate. D.4.5 SRV Piping.and Supports in Torus L The contraction of the torus will cause a maximum secondary stress increase of 4.9.ksi in the SRV piping. This is less than 10% of the piping allowable stress level and will be reduced due to expansion of the torus under operating conditions and is therefore acceptable. It also follows that the supports I will not see significantly higher loads, and remain acceptable. l The SRV quencher supprt beams were also reviewed. These beams utilize slotted holes in connection details on one end, allowing sufficient gaps for the calculated displacements (Reference 1, Sh. 37). Therefore, the SRV quencher supports remain acceptable. O D6 Rev. 2
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1 g,, ] e t '5 glM 'D.4.61 Torus Attached. Piping and Supports i Torus attached piping (TAP) was reviewed to determine the effects of the torus ; !:' ;(~ p).' . = shrinkage displacement. It was found that' TAP stress levels remain within l
' code allowables. Additionally, .it was determined that any shrinkage-induced ;
o load. on the supports, combined with' original loads', would not exceed design ] ,A- 'allowables. -Therefore, TAP and supports. remain acceptable.. j 1
, D.4.7: Torus Seismic Ties . .
1 j The seismic ties'are comprised of vertical plates with a 3/4" space between - the plates. This. space exceeds the calculated design basis displacement. , l[i l Field-information'also confirmed that sufficient gaps remain between the- I L,
- vertical plates (Reference 1, Sh. 36). Therefore, the seismic ties remain !
acceptable.< 1 I 1 1 1 p-1
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l I 1 1 1 l l l Il O ' D.7 Rev. 2 l l
c, t L TABLE D.4.1 1 a ({} LOAD CASE 1 D + Ri + Pi + T, + SRVTP ,,3 + CHTPcreati + SRVB0 3,, + CHBD<toits + EQ(0) Load UPLIFT LOAD PER BAY (Kips) Component"' Original Reduced Design Basis Dead -472- -472 R,- 282 193 SRVTP 3 + SRVBD i ,,+ 2113 1623
' CHTP c p,,,,,3+ CHBD c ,,,,,
EQ(0) 47"' 26 SRSS Reduction -4708 -4 41") Total Uplift 1923 929 Load per Bay N O Maximum Load / Bolt- 234 .142 1 (8BoltsEffective) - (1) Pressure (P i reactionswb)enappliedtothetorusshell.and temperature Thermal (T,)from reactions loads do not provide subsystems are accounted for in reaction loads.
-(2) Original analysis conservatively used SSE EQ.
(3) SRSS.of seismic loads with absolute sum of other dynamic loads. l (4) SRSS of all dynamic loads. l 1 l I l O :
, Rev. 2 3
v i TABLE D.4.1-2 : LOAD CASE 2 D + R, + P, + T, + COTP + COBD + EQ(0) [r r F Load . UPLIFT LOAD PER BAY (Kips) l Component" Original Reduced Design Basis ; Dead -472 - 472 !. R, 186 78 ! [' i COTP + COBD 1611 1611 EQ(0) 47"' 26 ! SRSS Reduction -47"' - 28"> . I Total Uplift Load 1325"' 1215 , per Bay Maximum Load / Bolt 120 105 () .(8 Bolts Effective) * (1) Pressure (P reactions wb)en applied to the torus shell.and temperature Thermal reactions from(T,) loads do not provi, subsystems are accounted for in reaction loads. * (2) Original analysis conservatively used SSE EQ. (3) SRSS of seismic loads with absolute sum of other dynamic loads. (4) SRSS of all dynamic loads. : (5) Load reduced 471 kips due to elimination of SRV loads from previous ! governing load combination 1 (reference Sect. 6.1.1.2).
-Q Rev. 2 3
- ,_ - . , - . . , . . - - . . ---,-...a,,n, .,. .a, - ,-n-, -
r t. , I t [ TABLE D.4.1-3 :
.x LOAD CASE 3 1 (f-D + R, + P, + T, + PSTP + PSBD + PSFI + EQ(0) c, Load UPLIFT LOAD PER BAY (Kips) e L,- Component"' Original Reduced Design Basis ! , ',: i Dead -472 -472 i i
R, 320 234 o PSTP + PSBD + PSFI 342 342 : EQ(0) 47"' 26 i
; SRSS Reduction -47"' -28"' ,
Total Uplift 190"' 102 Load per Bay Maximum Load / Bolt 62 52 ' 20 (8 Sa't> trr ct've) (1) Pressure (P ) and temperature (T,) loads do not provide net uplift , reactionswben.appliedtothetorusshell. Thermal reactions from - subsystems are accounted for in reaction loads. , (2) Original analysis conservatively used SSE EQ. (3) SRSS of seismic loads with absolute sum of other dynamic loads. (4) SRSS of all dynamic loads.
- l (5) Load-reduced 471 kips due to elimation of SRV loads from previous governing combination o (reference Sect. 6.1.1.2).
1 I i 1 O Rev. 2 L j
G F TABLE D.4.1-4 ROCK BOLT APPLIED LOAD VS. MINIMUM ALLOWABLE LOAD Calculated Vo11ft Allowable Bolt Load ") 142 urs 177 up, m on (1) Applies only to 13 bolts which could potentially be laterally loaded. The allowable load on the remaining bolts is unchanged from that shown - in Table 6.1.1-6, (2) Conservatively assumes a bending strain associated with a flexural displacement of .378 inches prior to application of the tensile load. (3) This is 2/3 of the allowable tensile load. O l Rev. 2
(c D.5.0 SOURCES OF CONSERVATISM Conservatisms listed within Section 8.0 remain applicable except that (3 conservatisms associated with wet structure analysis have been removed for the w/ torus rock bolt evaluation. In addition, the following conservatisms are applicable to the rock bolt analysis. D.5.1 Bolt allowables are based on bolts assumed to have bending combined with axial strains equal to 50% of ultimate strain. There is no . physical evidence and very small likelihood of any bolts being loaded e laterally (Reference 1). I L D.5.2 The total resistance to uplift is shown in Table 0.5.2-1. Considering all torus rock bolts and reduced capacities, it is evident that significant margin in uplift resistance is available when compared against the applied loads. l N D.8 Rev. 2 l h.. - .j
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, TABLE D.5.2-1 I" Pm TORUS UPLIFT CAPACITY Minimum Allowable Number Uplift Resistance Reduced Design Basis of Capacity (1) Uplift Load Condition Bolts (Kips) (Kips)
;t..
All Mitre Joint 256 45,312 19,440 Bolts Critical . Torus Bay 16 2,832 1,215 Mitre Joint Bolts (1) Conservatively using allowable bolt loads from Table 0.4.1-4 for all
,, bolts.
b Rev. 2
r - f ! D.6.0
SUMMARY
AND CDNCLUSIONS Data obtained on the position of the Plant Hatch Unit 2 torus mitre joint rock
/7 bolts, relative to the saddle base plate slotted holes, indicates that the U. torus has undergone a reduction in its major diameter. Due to the observed location of 13 rock bolts in their slots, the potential for these bolts to be loaded laterally was conservatively assumed. A review was initiated to ;
evaluate the condition. The root cause of the saddle base plate movements is ' weld induced shrinkage of the perimeter of the torus. This resulted from the r sequencing of the modifications performed under the Mark I short-term and long-term programs. - This review indicates that the slotted hole detail used in the saddle support ; base plates and the original fabrication and construction techniques provided
. sufficient clearance to accommodate this movement. Although only 13 of 256 rock bolts could possibly see any 1.ateral load, the bolts were conservatively i evaluated using a reduced capacity due to an assumed strain associated with a ;
lateral displacement. Considering this unlikely condition, and reducing loads by removing some conservatisms, the bolts were shown to be within the reduced design allowables. Additional evaluations were made which determined that the torus shrinkage did not impact the integrity of other structural and piping components. Therefore, the existing design basis for these components as documented in Sections 1.0 thru 6.0 remains valid, and there is sufficient margin of safety in uplift resistance. , 4 b k i I l l
]
l 0.9 Rev. 2 l l
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i REFERENCES n 1. CBI Services Report, " Phase !!, Root Cause Report, Weld Induced Q Shrinkage Displacements, Suppression Chamber E. I. Hatch Nuclear Plant Unit 2", 10/89.
- 2. GE Report EAS-71-1088, "Evaluatica of Conservatisms in Hatch Torus Uplift Loads". Rev. 1, 2/89, i
- 3. Letter from R. P. Daly (GE) to D. R. Quattrociocchi (Bechtel), G-GPC 140, dated 6/7/89. -
- 4. GE Report PED 47-0989 " Square Root Sum of the Squares for Combining Dynamic Loads for the Hatch Torus Support Components", 9/89.
i 4 o : I l l l 1 l 1 l D.10 Rev. 2 l I 1
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