ML20235B828
| ML20235B828 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Bodega Bay |
| Issue date: | 07/20/1964 |
| From: | Crane P, Sibley S PACIFIC GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20234A767 | List:
|
| References | |
| FOIA-85-665 NUDOCS 8709240240 | |
| Download: ML20235B828 (39) | |
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- f PACIFIC GAS AND ELECTRIC COMPAN'l DOCKET NO. 50-205 t u.o cc:.'l ( ': W
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AMENDMENT NO, 8
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BEFORE THE UNITED STATES ATOMIC ENERGY COMMISSION In the Matter of PACIFIC GAS Docket No. 50-205 AND ELECTRIC COMPANY Amendment No. 8 Now comes PACIFIC GAS AND ELECTRIC COMPANY (the Company) and amends its above-numbered application by sub-mitting herewith Amendment No. 8. This amendment sets forth additional data concerning Unit No. 1 of the Company's Bodega
. Bay Atomic Park (the Plant) and is responsive to the Commission's questions set forth in letters to the Company dated May 19, 1964
? and July 8, 1964.
The Company's independent experts have advised and the Company believes that within the foreseeable future and i well beyond the life of the Plant, any movement exceeding a l fraction of an inch is extremely unlikely, and any movement exceeding one foot is practically impossible, on any of the j fractures (or so-called minor faults) in the granitic rock at the site of the Plant.
Such fractures as exist at the site were probably caused during or shortly after emplacement of the magma.
Those fractures in the rock at the site which are nearly at r3.ght angles to the San Andreas Fault were probably caused
's at a time when the rock at Bodega Head was under strain 1
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], caused by its rounding a curve of the San Andreas Fault some eons ago when such rock was moving along the Fault many miles south of the Bodega Bay area. Such strains in the rock of Bodega Head caused changes in the dimensions or shape of the rock as it moved around the curve of the San Andreas Fault in that other location. The strains were relieved by the formation of the minor fractures noted. Today the straightness of the San Andreas Fault in the Bodega area precludes the occurrence of either new fractures of this type or any measurable displacements along the old ones in the foresee-able future. Furthermore, because there is no evidence of any vertical movements in the Bodega area in the geologically r,e-cent past, they cannot be expected in the foreseeable future, 7'
as displacements which have caused earthquakes are not produced haphazardly.
In the foreseeable future strains in the earth in the general area of Bodega on both sides of the San Andreas Fault will be relieved by differential ground movement along the known zone of weakness in the area, i.e., along the narrow belt representing the most recently active portion of the San '
i Andreas Fault, some 1-1/4 mil'es from the site, where ground '
movements have occurred during the past centuries and can be expected to occur in the future. The rock of the Bodega Head site has not experienced any differential ground movement exceeding a fraction of an inch during at least the last
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..; 40,000 years, during.which there have been hundreds of large ground movements in the San Andreas Fault zone. Consequently, :
the likelihood of.any differential ground movement at the site during the life of the Plant exceeding a fraction of an inch 1 is extremely remote.
To assure, however,.that no reasonable question can be raised as to the. safety of the Plant'from possible seismic i activity.in the area, the Company has modified the design of 1
the Plant to withstand several feet of differential ground motion, all as more particularly described below in answer to the Commission's questions.
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.(4) Question 1 - Part I Assuming, for two different cases, total shear displace- l ment of as much as (a) 2 feet, (b) 3 feet, including both l horizontal and vertical components, along any line and in any direction in the foundation, would the plant, as de- ,
signed or as you may r,copose to modify the design, be constructed so that:
(1) the structure and leak tightness of the containment building would not be impaired? j In addition to designing and constructing the reactor containment structure to withstand earthquake ground motion as described in Amendment No. 6, and as further detailed in the reply to question No. 4 following, the Company will design and construct the reactor containment structure so that it will withstand 3 feet of horizontal or vertical differential ground
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tainment area without damage sufficient to cause impairment of function. Basically, the design provides for a 3 foot unob- [
structed radial clearance between the outside of the reinforced concrete containment structure and the inside of the containment pit, completely around the circumference, from elevation -73 to yard elevation at t25 The walls of the reactor containment pit will be lined with reinforced concrete to prevent possible spalling of material into the pit. The annular space will be permitted to fill with water. The reactor containment structure I
will be founded on a layer of carefully selected sand of known !
characteristics which will permit horizontal movements up to 3 )
I feet without impairing the function of the containment structure, j although the structure might be shifted or rotated. Details of l I
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)b this design are as indicated on the drawing accompanying
? knendment 7 except' that the annular space will not be filled ,
with compressible back till. . Differential vertical motions up'to 3 feet may cause'the containment structure to tilt or shift, but in no case will the containment' function be im-paired.
The Plant will be designed with no rigid structural interconnection between any major component. The reactor con-tainment structure will be structurally independent of the turbine generator foundation,'the Plant control building, the -
radwaste facility, the. stack, and the Plant service buildings, which contain the office and machine shops. Piping and wiring interconnections important to safety between the reactor con-tainment structure, the control building and the turbihe ,
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generator will have sufficient flexibility to accommodate 3 feet of re:ative movement. In order to prevent overstress at points of penetration for piping connecting the dry well with the turbine, the Company will provide adequate anchors and bracing adjacent to the containment shell and beyond the double isolation valves. This anchor will be adequate to withstand all piping loads due to differential motions in any direction up to 3 feet between the reactor containment structure and the turbine generator foundation.
In summary the Plant as designed will be constructed so that the structure and leak tightness of the containment S. building would not be impaired in the highly unlikely event
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k of total shear displacement of as much as 3 feet, including both horizontal and vertical components, along any line and in any direction through the reactor containment pit.
4 Question 1(ii)
The ability to shut down the reactor and maintain it in the shut down condition would not be impaired? ,
The ability to carry out and maintain a safe reactor shutdown involves considerations of control rod operation,.
excess heat rejection, dec' ya heat removal and makeup water.
Normal insertion of control rod drives will be provided by the control rod drive pumps driven by either normal auxiliary power ,
or the emergency engine generator. Additionally, scram energy - '
will always be available from either reactor or accumulator
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pressure. Further control will be available from the liquid poison system which can be powered either from normal auxiliary power or the emergency engine generator. Components of the control rod drive system, liquid poison system and emergency power system will be designed to remain operable during and after the design earthquake. (i.e., the earthquake upon which Dr. George W. Housner's design spectrum is based. See Figure 1 and the answer to question 4 below.)
Should the turbine load be lost during an earthquake, i
or for some other reason a shutdown be initiated from high power .j l
1evels, the reactor stored heat and the immediate decay heat i I
would be rejected to the main condenser by means of the turbine
) bypass valves. Condensate storage in the condenser hotwell l
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L !' i {;) willibe sufficiently large to cbsorb this heat. following a scram even with'the circulating water system inoperable. If an isolation signal occurs during or.following'a' scram, the reactor steam would pass to the main condenser and be condensed in the hotwell for about 30 seconds while the main steam isola-
't tion. valves are closing, at which time the solenoid. actuated bleed valve, discharging to the suppression pool, and emergency >
. condenser would-provide for heat rejection. Safety valves on
,the primary system provide additional assurance of the constant ,
availability of a heat removal system. Removal of decay heat for periods longer than.the first few minutes after shutdown involves the same systems as
,s _ above,-with the added availability of other low pressure systems.
1 . . The main condenser will-be available as long as vacuum is main-tained and the main steam isolation valves do not close. The l solenoid actuated bleed valve and suppression pool coolers can E' be used for long periods if reactor makeup water is available, , but the use of the emergency condenser with its capability of . rejecting heat without losing reactor water is preferable. Makeup of water to the shell side of the emergency condenser , will be available either from the Plant's raw water storage tank (gravity flow) or pumped (a) from the Plant's condensate storage system, or (b) by mobile equipment or fire protection system via hose lines through a standpipe with a connection
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mounted on the exterior walls of the refueling building. Water
'y supply for (b) could be the storage tank itself or Bodega Harbor.
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cooling systems can be used for further cooling. Makeup water to the reactor can be supplied from the condenser hotwell or condensate storage via the motor driven start-up feed pump, the control rod drive hydraulic system pumps, or from the liquid' poison pumps.
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t Much of the auxiliary equipment described above to J shut down and maintain the shutdown reactor is totally contained I _, in the reactor containment structure. Elsewhere in this amend-ment descriptions of the reactor containment structure give assurance that the structure and its contents would not be damaged by a shear displacement of up to 3 feet. Where vital components of the emergency systems are located within the turbine generator foundation or the control building, the interconnecting piping and cable will be designed to withstand up to 3 feet of relative displacement between the reactor con-tainment structure and the turbine generator foundation, or control building. Where components of an emergency system are remote from the power plant buildings and would be endangered by shear displacements of up to 3 feet, alternate methods will be provided to restore the system effectiveness in a timely
; manner. Thus, a mobile pumper and hose system will be provided 8 i
q V v+ STMGiM&zD2922.c5a559N%bCM +*&3&sh:MbliV'Mca*12:.hsusMLL;[ n ,I ~ . p '<?- 9 , u i . ([)- on the site 1to restore water to the emergency condenser shell , should-the normal means and the line from the raw water storage tank be unavailable. Similarly, seawater can be pumped through the' fire system pipes (or hoses) to the suppression pool heat y exchanges. .' In summary the Plant as designed will be constructed so that under the displacement assumptions contained in the question the ability to shut down the reactor and maintain it in the shut down condition would not be impaired. Question 1(111)' ,. Primary system would remain intact? . The primary reactor system is located within the
~ ~I) reactor containment structure. The design is such that a 3 foot displacement would not affect the structure's function '
as a support and container for the primary system. Accelera-tions experienced by the primary system during such a dis- - placement would be less than the accelerations used in the I design of the equipment. ! The design of the main steam lines and supports be-tween the anchor at the drywell penetration and the turbine inlet and turbine bypass valves in the turbine generator foundation will accommodate relative movement of 3 feet with-out pipe failure. An analysis of the preliminary design of .
. the piping system shows that ultimate tensile stresses in bending are not exceeded with 3 foot relative movement between l
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vr~ those adjacent supports which would resist the' displacement, Q ' assuming elastic pipe response and rigid supports. In the l actual case stresses in the pipe would be much less from e 1 yie'1 ding of the pipe and supports. The isolation valves in the main steam line would remain operable under these condi-tions. Thus the Plant as designed will be constructed so
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that under the displacement assumptions contained in the question the primary system would remain intact, s. Question 1(iv) Supply of power to the facility would not be , interrupted? ,,., l The Preliminary Hazards Summary Report describes
) the various sources of power to the Plant. Briefly these are the main unit turbine generator, two 220 kv transmission circuits, a 12 kv circuit from the local subtransmission net- !
work, an emergency engine generator set, and a station battery. The overhead transmission lines are not particularly vulnerable
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I to a random 3 foot ground displacement, nor is the 12 kv pole line which enters the site from the opposite direction. Should the turbine generator be tripped or in some other way fail to supply power to the 4 kv auxiliary power bus, the standby-startup transformer fed from either of the 220 kv transmission circuits would automatically be switched to feed the station auxiliaries. Should both of these sources be lost, 1 1- the auxiliary bus would be switched to the transformer fed from i 10
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f , t ,f* , j I((] the local 12 kV circuit. This system has the capacity for a . normal shutdown. If all of these external sources are unavailable, the , engine generator would be started and loaded on the 480 volt ! 1 bus with those auxiliaries required for a safe Plant shutdown.
. The engine generator, switchgear and control will be located in the reactor containment structure and control building, with their interconnections capable of withstanding 3 feet of relative movement. Since the engine generator will be connected ;
to a normal 480 volt bus, any of the 480 volt loads can be con-nected to the engine generator. - The station battery, d.c. switchgear and control for _, vital loads will be located in the reactor containment structure y and control building, with their interconnections capable of withstanding 3 feet of relative movement. The battery supplies , control power and serves instrumentation, emergency lighting, communication, isolation valve and certain other vital loads. The battery is normally in continuous use and is charged from the 480 volt bus, so it is considered to be of the highest reliability and continuously available. The battery can pro-vide the required power to shut down the Plant safely and main-tain cooling for 8 hours if gravity makeup to the emergency L condenser is available, but in no case less than 2 hours.. Assuming loss of all other power the emergency engine . generator would be required about 2 hours after a shutdown l' begins, to pump makeup to the emergency condenser if it were 11 ! 1
bk.clis55slkNGeb5?,htEdi MbGtliMinkhdddiOR$bdLE$"$b *'-'05 Y$Akik2:5N: 1b:b!:=hQ \ : f,'..,,c o n [h. .- not available from the gravity or mobile pumper sources. Emergency.' power to provide makeup to the reactor would not be ' required for 8 hours following a shutdown. In short-the Plant as designed will be constructed so that. under the. displacement assumptions contained in the question the supply of power to the facility would not be interrupted. .
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f : [If' 1" Question'1 - Part'II Under the displacement assumptions stated;above, what . 1 measures are proposed to assure that the reactor-could be maintained safely in a shut down condition indefi-nitely if all vital connections to the reactor build- !
- ing were. severed?. Additional related questions in- 1 cludeLthe following:
'a. What are the arrangements (pumps, power-sources, connecting lines) which give confidence that the reactor could still be' shut down.by normal or.by alternative systems?-
- b. .For what direction or directions of ground slippage are vital internal components most vulnerable in case of damage to the building?
How would vital safety components inside the building be protected against damage if the building were damaged by slippage? ac . lWhat and where are the vital components of emergency electrical power sources? 'What
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equipment would they serve luud on what time
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- d. .What are the alternative sources of emergency cooling? Where'are their vital components and:what assurance is there that these would not'be inactivated by any accident which might ,
inactivate.the primary cooling system? What ! are their capacities and time schedules of effectiveness? If the Company considers that assumptions, criteria, or design objectives different from those stated in the questions contained under item 1 above could be used and provide adequate protection, please describe these and explain how they would be met. . As permitted by the concluding sentence in this part of question 1, the Company considers that a different assump-tion than the one stated (i.e., "if all vital connections to
, the reactor building were severed") can be used and will pro- -
vide adequate protection. This follows from the fact that
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[]) as. outlined in the answer to Part I of this question the Com-l pany's proposed design' for the ' Plant will assure that under, l t
.the-displacement assumptions stated the function of the reactor ,
2 containment ~ structure 'would not be impaired and all vital con-l - 'nections to the reactor building would remain intact. This in turn will assure that the reactor could be maintained safely ' in a shutdown-condition indefinitely. Accordingly, detailed answers to the additional related~ questions are unnecessary i since under the displacement assumptions of the question: ! (a) The arrangements outlined in the answer to part (ii) L of the question would still be available so that the reactor I could be shut down; (b) The reactor containment structure would not sustain " ?'g. - any damage which would impair its function. Thus, the structure and vital internal components would be equally invulnerable to ground slippage from any direction, and vital safety components inside the reactor containment structure would continue to be i protected by the structure; i (c) The vital components of emergency electrical power l 4 sources, the equipment they would serve and on what time schedules would be as set forth in the. answer to part (iv)
- i of the question; and
'(d) The alternative sources of emergency cooling, their ;
vital components, capacities and time schedules would be as set forth in the answer to parts (ii) and (iii) of the question, s 1 t 1 14
a v. w _,. : w x, . , M~' V) . .asioutlined in the answer to Part I of this question.the Com-pany's proposed design for the Plant will assure that under the displacement assumptions stated the function of the reactor :
' containment structure would not be impaired and all vital con-nections-to the reactor building would remain intact. This in turn will assure that the reactor could be maintained safely
'in a shutdown condition indefinitely. Accordingly, detailed answers to the additional related questions are unnecessary since under the displacement assumptions of the question:
(a) The arrangements outlined in the answer to part (ii) of the question would still be available so that the reactor could be shut down; , (b) The reactor containment structure would not sustain 1 g 1 any damage which w'ould impair its function. Thus, the structure and vital internal components would be equally invulnerable to ground slippage from any direction, and vital safety components inside the reactor containment structure would continue to be protected by the structure; (c) The vital components of emergency electrical power 1 sources, the equipment they would serve and on what time
' schedules would be as set forth in the answer to part (iv) of the question; and I (d) The alternative sources of emergency cooling, their vital components, capacities and time schedules would be as
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set forth in the answer to parts (ii) and (iii) of the question.
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l' (,m Question 2 If damage should exceed all expectations and actual , core meltdown appeared imminent, or if decay heat from released fission products was causing pressure , buildup which threatened the integrity of the con-tainment, what actions might be taken to prevent , the meltdown? What devices and preparations could be made in advance? What time schedules would be
' involved?
The question assumes that simultaneously with a major reactor. coolant loss accident the multiple arrangements for core cooling are ineffective. For this to happen the two sets of core spray pumps and the two motor driven startup feed pumps which can feed the vessel either through the feedwater system or core spray nozzles must all be unavailable and in-capable of restoration.
) The design of the Plant'will incorporate a drywell spray system which could be connected to the fire system or i
a mobile engine pumper through a standpipe with connections on the outside of the reactor containment structure. This system would .be available on one hour's notice, and would function to both cool the containment system in the absence of core spray, and provide for additional fission product scrubbing action,. The system would be started on indications of a primary system rupture and failure of the core spray system to operate. If the normal core spray system could not be re- ) {
, turned to service, the entire drywell could be flooded by this j system in 24 hours, using Bodega Harbor as a source of water.
3' The suppression chamber air space would be vented through the i 15 b
%2i i2% ML-' Mba 2 &%bdikdh­G&&&&%hdi~dmham stack to limit pressure in the chamber during flooding. The drywell air space would be vented during the last 8 hours of the flooding period through the gas treatment system to the y stack. The shutdown cooling system or the suppression pool .
coolers would then be restored for cooling. If all efforts to restore cooling to the containment system failed, venting 1 i steam from the flooded drywell through the gas treatment system 'I I and stack and supplying makeup would provide for indefinite j l cooling. , 1 The drywell spray system with its alternate water j supplies could be prepared during accident recovery conditions j I even with the core spray system operable as a ba'ck-up for possible failure of core spray. '~ O l Other methods of introducing water into the reactor l 1 in addition to the core spray system include the'feedwater i system, control rod drive pumps and liquid poison pump. The j l degree to which these systems would be effective in cooling j i the core would depend on the size and location of the primary { system break. Question 3 What is the degree of damage to the reactor building and the reactor to be expected from shear displacement along any line crossing the reactor building shaft? This analysis should not assume a size of displacement. What is desired is damage as a function of displacement. ; What displacement leads to fracture of the concrete
. structure? What displacement would rupture the con-tainment? What displacement would lead to rupture of the primary reactor system? It is vital that these ,3 '
judgments be based on features of the system as it is l to be built, and not be supported only in general terms. { The effects of both shear and tensile strains should be ~ considered. 16 i l I I
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1 (_s The reply to this question is based on the method i of design and construction set forth in the reply to question
- 1. This method provides 3 foot radial clearance around the circumference of the reactor containment structure from eleva- ,
tion -73 to elevation +25 at yard level. This space is to be kept free of all foreign material and debris except that it will be permitted to fill with water. The reactor containment
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structure will be founde'd on a layer of carefully selected sand of known characteristics which will permit differential move-ments in the bedrock beneath without transmitting them directly to the reactor containment structure. As stated in the reply to question 1, for differential _s vertical motions up to 3 feet in any location through the re-l actor containment area the containment structure would be tilted or shifted, but, because of the inherent strength of the rein-forced concrete containment structure, in no case would its ; containment function be impaired. For horizontal shear displacements up to 3 feet along any diameter of the reactor containment structure, the structure itself would not be damaged but might be moved or rotated. With 3 feet horizontal displacement along a diameter one portion of the containment pit wall would come in contact with the re-actor containment structure. From 3 feet to 6 feet the reactor building would be vibrated and moved additionally on its bed of sand. At 6 feet total horizontal displacement along a major l', diameter the reactor containment structure would have moved 3 l 17 E--___---_--_
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l Although minor damage might be incurred by the structure con-tainment would not be impaired by up to 6 feet of total hori-zental movement. Beyond 6 feet of total horizontal movement , along a diameter local spalling and crushing of concrete of the containment structure would occur, and progt'essive failure of the structure would result depending upon the total amount of ground movement. For horizontal shear displacements at locations other than along a diameter larger' displacements than 6 feet could be absorbed without damaging the reactor containment structure. Because the major Plant components are not structu- _ rally interconnected there would be no dama6e to the reactor containment structure due to contact with other structures in i the event of horizontal shear displacements up to 3 feet. The l 1 main steam piping and reactor feedwater line between the reactor 1 1 and the turbine generator will be rigidly anchored adjacent to the drywell shell beyond the double isolation valves so that no forces or thrusts due to differential motion up to 3 feet would be transmitted to the drywell shell or isolation valves. All piping forces due to 3 feet of differential motion in any direction would be absorbed by the anchor structure. The steam and feedwater lines themselves will be so designed as to absorb differential motions of up to 3 feet without failing. l l, T l 18 L_____---.-
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fg Question 4
';f We wish to be sure we understand the specific methods the Company proposes to use to analyze the ability of the structures to withstand earthquake oscillations.
Some new features of the analysis were introduced at the last ACRS meeting. Does the Company propose to modify the frequency spectrum previously proposed (based on El Centro, 1940) to take into account the ' rock foundation at the Bodega site? The methods which the Company will use to analyze structures for earthquake loads have been presented to the Commission at other times (see Appendix V, Preliminary Hazards Summary Report, Anondments 4 and 6). What follows makes no changes in the design principles and criteria previously proposed, but amplifies the earlier documents and for clarifi-cation brings to6cther in one place all of the data pertaining to the specific methods of seismic design. The word " structure" as used in the question is taken to mean (a) a structure in the ordinarily accepted sense, such as a building or a tower, (b) elements of a structure such , I as walls and floors, or (c) equipment or systems in which seismic j loads are an important part of the design. ' Critical and Non-Critical Structures For purposes of seismic design, structures, equip-ment and systems are classified as critical and non-critical. Critical structures (called Class I structurco in Appendix V, Preliminary Hazards Summary Report) include the reactor con-
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tainment structure, reactor refueling building, control build- !
. I ing, the ventilation stack. Critical equipment and systems j
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l / l include the nuclear steam supply system inside containment, isolation valves, liquid poison systeni, emergency cooling systems, emergency electrical power system, and the necessary associated instrumentation and controls. Non-critical stractures, equipment and systems will be designed to withstand seimic loading depending on the purpose of the structure. Each structures, equipment and systems concerned directly wiC: electric generation will be designed to withstand lateral loads of 20% of gravity on each element based on dead load plus one-half the live load. Such structures, equipment or systems not concerned directly with electric generation will be designed to conform to the Zone b.- III requirements of the latest edition of the Uniform Build-ing Code of the International Conference of Building Officials. This procedure is in accordance with the Company's usual prac- , tice for conventional steam plants. . Critical structures, equipment and systems will be designed in accordance with the recommendations of Dr. George W. Housner. Lateral forces on critical structures, equipment and systems for design purposes will be computed on the basis of the design earthquake, taking into account the natural fre-quency and damping characteristics of the particular structure, equipment or system. Stresses in critical structures, equip-ment or systems, including the stresses due to seismic load, will be limited, where applicable, to the allowable stresses permitted by the ASME Boiler and Pressure Vessel Code,
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) -Section VIII, including applicable nuclear code cases, the ASA Code for Pressure Piping, and the Uniform Building Code, .
except that the usual'one-third over-stress for earthquake permitted by the Uniform Building Code will not be applied. In addition, the design of the Plant will be checked to assure.that all critical structures, equipment and systems will be capable of withstanding earthquake ground motions cor-responding to spectrum displacement, velocity and accelerations two times as great as shown on Figure 1 without impairment of l functions necessary for containment and safe Flant shutdown. Following is a description of the methods by which lateral loads for design purposes will be obtained for critical _ structures, equipment and systems. Design Spectrum For Critical Structures Figure 1 is the design spectrum proposed by Dr. Housner to describe the horizontal oscillation characteristics of the design earthquake. The design spectrum represents the statis- l tical ensemble average of ground motions recorded in the epi-central regions of large earthquakes. The design spectrum curves are for ground motions recorded on rock or on firm soil; they are not applicable to soft soils. For purposes of illus-tration, Figure 2 shows computed velocity response spectrum curves for the earthquake of July 21, 1952, recorded on rock at Taft, California. The zero damping spectrum curve and the
~
10% damping spectrum curve are shown. The dotted lines show i !) the corresponding design spectrum curves plotted at 3/4 scale. I
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fj Figure 2 shows that the shape.o't the design spectrum curves ! are good. representations of the; ensemble average's of spectra
..n .,
'like those for the Taft ground motionF that was. recorded on rock, g)s
'Further illustrations are'given in Figures 3;and 4.
- Figure 3 shows - the zero damped tatt ~ velocity . responed ' spectrum curve plotted versus frequeney so as'to expand the.short period 6 .
region. A similar plot is shown in Figure'3 for the jelena, '
~'- (a Montana earthquake of October 31, 1935. This Magnitude 6.0 t i
shock was-recorded on rock within a few miles of the epicenter. The dotted lines in Figures 3 and 4 are the zero damping de- 1 signspectrumcurvesplottedatreducedsca'idlofacilitate
.. 4' >
comparison of shapes. The spectrum curves shown in Figures 3- l and 4 are typical of all' strong ground motions recorded on rock ) l3' ' and firm soil. - { The design spectrum (Figure 1) is a set of curves - l representative of the significant spectral characteristics of l past, recorded, destructive ground motions on rock and firm soil. The scale of the design spectrum corresponds to the maximum intensity of ground shaking expected on rock foundation in the vicinity of the San Andreas fault during an earthquake ) greater in Magnitude than the 1906 shock. The spectrum curves show a maximum acceleration of 0 33g at very small periods of vibration (maximum ground acceleration), and a maximum accelera-tion of 2.0g at period of 0.25 seconds and zero damping. Damping coefficients which the Company will use for design of critical structures, equipment and systems are those l i L 22 l l l l 1
plt?S322kbri59 N' @McE6sx=2db&6b.dNisL=$ithhE.:U :::.:] M -- k' t 4e 4, f. '~' given in the following table by Dr. Housner: , Structure, Eauipment, System, % Critical Damping l Steel fram? structures 2.5 I l Reinforced concrete frame I structures 4.5 ! I Reinforced concrete reactor ] containment structure- 7.3 j l Welded assemblies , 1.0 Boltedangemblies 2.0 Vital piping systems 0.5 a For structures and equipment supported on the ground, I such as the reacter containment structure, the natural period of vibratIcn and the shape of the fundamental mode will be determlaed and the equivalent single-degree-of-freedom system s , I will be calculated. Reading from the spectrum curves for ap-propriate period and damping will determine the maximum ac-celeration of the equivalent system and from this the maximum acceleration of each level of the structure is determined.* Equipment within a building will be excited into vibrations by the motion of the part of the building to which the equipment is fastened. This motion will be taken to be a steady-state sinusoidal motion having the period of vibration
- Chapter 5 - Nuclear Reactors and Earthquakes TID-7034 USAEC Chapter 2 - Blume, Newmark, Corning - Design of Multistory Reinforced Concrete Buildings for Earthquake Motions, PCA,1961 '
i
. 23
(($MMM?pWA WwQa.;gyg.gggagg;gygggg,{g.ggfg.g] . f o. e o, i v,. ~ f I . P's L_; of the fundamental mode and having an amplitude equal to the maximum amplitude of that part of the building as determined by the spectrum curves for the fundamental mode. In addition, to account for the effect of the higher modes of vibration, each part of the building will be assumed to have accelerations the same as those of the ground but multiplied by a factor that varies linearly over the height of the building, being unity at the foundation level and zero at the roof. The design ac-celebrations of a piece of equipment will be that produced by. both motions and will be determined by means of the resonance cur'Jes for steady-state harmonic excitation, and by means of the design spectrum curves multiplied by the appropriate factor. The resulting stresses, strains, or displacements of the equip-
- 3. .
~
ment produced by these two excitations will be combined as the square root of the sum of the squares of the individual values. For some pieces of equipment a detailed dynamic analysis will be made. This will involve representing the reactor containment structure and the equipment under consideration by a system of lumped masses, springs and dashpots, and calculating the maxi- j mum accelerations and forces produced by a base motion corre-sponding to an earthquake accelerogram the spectral values of which are equal to the corresponding values of the design spec- ) l trum. This procedure is based on ground accelerations.having i I the same design spectrum as shown in Figure 1. The ground will be assumed to move vertically as well i as horizontally, and the vertical motion will be represented
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by design spectrum' curves having ordinates two-thirds as great - as those for horizontal ground. motion. Stresses due to vertical and horizontal seismic. loads will be combined by taking the square rootLof the sum of their squares, and the result will be added algebraically to stresses due to other design' loads. I e e r O e 25 _ _ ____m._s________. _______
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What measures would be taken to protect against ' tsunamis greater in size:than the breakwater at Bodega Bay would suppress? . 1 Records of earthquakes and tsunamis indicate that l tsunamis are caused primarily by major submarine landslides or dip-slip type of faulting. Tsunamis do not appear to be -
. i associated with strike-slip type of faulting such as is ,
characteristic for the San Andreas fault. A review of avail-able historical records reveals no tsunamis striking the . California coast from earthquakes on the San Andreas fault. l Nearly all tsunamis recorded along the California coast originated'elsewhere and traveled across the ocean to California. None of these has affected Bodega Bay levels i See Appendix I, " Tsunami Information by more than 5 feet.
]
In Regard To Proposed Nuclear Power Plant Site, Pacific Gas 1 and Electric Company, At Bodega Head, California" by Robert L. Wiegel.
]
In any event the configuration of Bodega Head with the location of the proposed nuclear plant in Campbell Cove is such that this site is not subject to direct approach of tsunamis. The effect, on the site, of large tsunamis which are not suppressed by the breakwater at Bodega Bay would be similar to that of a rapidly rising tide rather than a breaking wave against the site and structures. The forces which would come to bear on the structures would be hydrostatic pressures } rather than dynamic blows from impacting waves. With mean l 26 !
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... ,,- i h lower low water at elevation zero and extreme high tide at about f7 the Plant yard level has been established at elevation t25, or about 18 feet above extreme high water. This is greater than ,
the height of any seismic sea wave of record along the California i coast. The western edge of Bodega Head, facing onto the Pacific Ocean in the direction of possible tsunami approach, ranges in , elevation from 75 to 100 feet above high water. This height I would be more than adequate to protect the site against tsunamis 1 l greater than the largest recorded along the California coast. ! l Disregarding the Pacific tsunami warning system, which would give Plant operators adequate time to prepare for a possible emergency, the arrival of a wave at the Plant site would first be indicated by rising water level. For waves higher than eleva-4'
~
tion f15 in Bodega Bay the water would flow over the breakwaters - and probably over Doran Beach. For this case the constriction . at the entrance to the harbor would essentially disappear, and the water height and duration within the harbor would be sub- ; stantially the same as in Bodega Bay. The first operational signal at the Plant would be from failure of the cooling water pumps, which would be short circuited when the water reached an elevation of about +20. This would initiate Plant shutdown. If the water continued to rise it would start to flood the yard when it exceeded elevation 125. Assuming the tsunami struck at i high tide (about elevation t7), this represents a rise in the i i l . water surface of about 18 feet, or approximately 3-1/2 times I the maximum seismic sea wave that has ever been recorded in
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tion f25, it would contact the Plant structures. Design of openings at ground level into critical structures will be such that water will not enter them at a rate that would prevent essential control equipment from completing the shutdown initiated when the circulating water pumps failed. Inasmuch as seismic sea waves .cecede within at most a few minutes, there is no more than a remote possibility of completely flooding a major component of the Plant. , All critical structures will be inherently strong enough because of seismic design requirements to withstand . water pressure up to the elevation of the operating floor
.(57' 6") without structural damage, and the Plant could be brought to a safe shutdown. When the wave recedes it might wash away portions of the excavated slopes and the yard, but -
the major structures, being founded on rock or other deep substantial foundations, would not be damaged because of pos-sible loss of support. Tsunamis do not represent a significant hazard to the proposed Plant. The possibility of a large tsunami reaching
. Bodega Harbor is very remote. Even in the unlikely event that one should somehow enter Bodega Harbor the probability is that I
the Plant would be unaffected. Even in the unlikely event of a tsunami large enough to flood the yard to a considerable depth, the Plant is capable of safe shutdown and no significant l l structural damage would result. On this basis it is considered 1 l 28
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C) u unnecessary to take further measures to protect the Plant against tsunamis greater in size than the brealciater at Bodega Bay might suppress. t. l 29 _.-m_-.__----
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F . f'.. .. l l Letter of July 8, 1964
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l l Question 1 l Assume an earthquake having a vibration spectrum similar to that of El Centro but with maximum ac- . celeration of 2/3g*; velocities up to 2-1/2 ft/see; ground displacements up to 3 ft; shear displacement (faulting) of 2 ft in the foundations of the plant, and with the occurrence of an aftershock equal to El Centro before remedial action could be taken. What plant arrangements, design criteria and pro-cedures could be developed to prevent impairment of functions of structures, equipment and systems important to safety? This should include method of analysis indicating the margins to allowable stress or deformation limits. The design spectrum, Figure 1, when multiplied by two, will correspond to earthquake ground motion having maximwn acceleration of 2/3g, a maximum ground velocity of 2.5 ft/sec,
'3 and maximum ground displacement of 3 ft. The design of the Plant will be such that when the base of the reactor containment structure has this motion, the containment will not be jeopardi-zed and the Plant will be capable of a safe shutdown. For example, the steel reinforcing of the containment structure will be designed so as not to exceed 20,000 psi stress under a 1/3 g ground motion.- The yield point stress is at least 40,000 psi and, hence, under a 2/3 g ground motion the steel will not exceed the yield point at its most highly stressed point. Also the failure stresses of the concrete are signifi-cantly greater than twice the allowable working stresses and,
*The maximum of acceleration, velocity and displacement would
) not occur for the same periods of vibrations.
30
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.. c, I hence, the concrete can withstand the stresses produced by a 2/3 g ground motion without danger of failure. The peak base shear under the 2/3 g ground motion will be o.8 times the weight 1 of the reactor containment structure. ,
l
~
To allow for the possibility of fault displacement in the rock beneath the reactor containment structure, as previously stated there will be provided a three-foot clear space between l the outside of the structure and the wall of the containment pit. . The structure will be founded on a layer of sand having a co-efficient of friction such that sliding will occur on the under-lying rock without exerting a damaging force on the base of the structure. The structure can withstand a horizontal fault dis-placement of two feet without any-damage to the reactor contain-
~
ment structure. Vital electrical and mechanical connections between the reactor containment structure and other structures, such as the turbine generator foundation, will be designed so that they can accommodate a relative displacement between the two structures of 3 feet ws'hout failure. There will be no structural interconnections bauween major components of the Plant. The reactor containment structure will be sufficiently strong to withstand without damage the postulated 2 foot vertical fault displacement under the structure by spanning either as a simple beam or a cr.ntilever beam. The reactor containment structure would not be damaged by this vertical fault displace-ment but, at worst, would be tilted out of plumb slightly. If the slip on a fault occurs under the reactor ]) 31
.s e, ;- ', containment structure simultaneously _with ground acceleration of 2/3 g intensity, the stresses in the structure and equipment will be different than if these two actions do not occur simultaneously. The prime effect would be to reduce the ,
oscillatory earthquake stresses in the structure and equipment because of the shearing through the sand layer under the struc-ture. While this horizontal shearing is taking place, the ground accelerations in the direction of shearing would not transmit any force to the base of the structure since the sand layer would already be slipping under the action of the fault displacement. For example, even if the fault trace bisects the structure, the ground acceleration would be only partly effective in applying accelerations to the base of the structure, T ' and the net effect would be to shake the base of the structure with an intensity less than that which would obtain if the faulting did not take place. The most severe condition, there-fore, would be produced'by the 2/3 g ground motion without simultaneous fault slip, and this condition is covered by the earthquake design, which will assure that the integrity of the 1 containment will be maintained and the Plant will be capable of safe shutdown. Actually, the maximum stresses will be somewhat less than twice those computed from the design spectrum since the sand layer will act as an elasto-plastic connection ! between the base of the structure and the rock, and the co-efficient of friction of the sand will be such as to limit the shaking of the base. See answer to question 2 below for 32 l
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q' y further discussion of the-behavior of the sand layer. " An aftershock having maximum ground. acceleration of 1/3 g would not jeopardize the containment or the functioning of,the' equipment since the foregoing' hypothetical 2/3 g ground ;
- l motion plus fault movement would leave a11' structures with' !
their original strength and, hence, they would be capable of : resisting the aftershocks without. exceeding ordinary' allowable working stresses.., The' design criteria discussed herein leave a'large margin of safety against failure from increases in seismic load. Consider, for example, a critical structure designed for seismic . loads obtained from the design spectrum (Figure 1) having a-maximum ground acceleration of 0.33g. The total design stress D for any element of this structure is the largest total stress obtainable by. superimposing dead and live loads, including seismic loads. Uhder these conditions stresses will not exceed allowable working stresses given by the appropriate codes. ..It is apparent, therefore, that increasing the seismic load will not represent a proportional increase in the design stresses, except for those unusual elements where the seismic load is the only one producing stress. Even in these cases, - doubling the seismic load from' O.33g to 0.66s maximum ground acceleration'would increase the computed stresses in these elements only to approximately the yield strength, inasmuch as.for most materials the yield strength is about double the ,. allowable working stress. A member stressed to its yield point y 33 a __ __ -__ _ _____-- -____- _ - - A
r
* ?
h L t'3 i/s or even somewhat higher by a vibratory load still retains a large capability t'o absorb oscillating energy in elastic and plastic deformation before failure takes place. Ductile materials are not damaged by being stressed to their yield points and upon removal of the stress retain their original strength and elastic properties. The ultimate strength of such materials is appreciably greater than their yield strength. The design of critical equipment and systems will be i analyzed with regard to stresses and deflections to assure that the equipment and systems will withstand seismic loads l double those of the spectrum in Figure 1 without impairment of functions necessary for containment and safe Plant shut-down. . For a further discussion of this matter as applied - i specifically to the reactor containment structure, refer to 'l ' attached Appendix II, " Margin of Safety Against Failure", l l prepared by Dr. Housner.-
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C - ,q Question 2 VJ Assume an earthquake having a vibration spectrum similar to that of El Centro but with accelerations generally up ' to 2/3 g plus intermittent pulses of acceleration up to 1 g*; velocities up to 2-1/2 ft/sec; ground displacements up to 3 ft; shear displacement (faulting) in the plant foundations of 3 ft; and with the occurrence of an after-shock equal to El Centro before remedial action could be taken. What plant arrangements, design criteria and pro-cedures could be developed which would assure shutdown and maintenance of the plant in safe shutdown condition? A horizontal fault displacement of 3 feet under the reactor building can be accommodated by shearing in the sand layer under the building. Inasmuch as there will be a clear space of 3 feet around the structure, there will be no contacc between the structure and the wall of the pit after the 3 foot displacement. Calculations show that for ground motions of 1/2 g , '3 : maximum acceleration the base shear on the structure would be ) approximately 0.6 of its weight. Design recommendations sub- I mitted by Dames and Moore, soils consultants, indicate that sand can be provided for the sand layer that would allow slip i to occur through the sand at approximately this ve.lue. See l Appendix III. For ground motions more intense than 2/3 g maximum acceleration, the slip in the sand layer would ef-fectively clip the peak accelerations the stresses of which
~
*The maximum of acceleration, velocity and displacement would not occur for the same periods of vibrations.
$) 35
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-m would be additive to the shear stresses produced in the sand kJ 1ayer by the fundamental mode of the building. For ground accelerations reaching 2/3 g there would be frequent slipping l
in the sand layer. The slip in the sand layer would introduce damping so that the accelerations of building and equipment
\
would be less than twice the accelerations produced by 1/3 8 ground motion. The sand layer acts effectively as an elasto-plastic connection between building and ground. Intermittent peaks of 1 g ground acceleration would be clipped by slip in the sand layer and would not increase the vibratory accelera-tions of building and equipment. If the fault slip occurred simultaneously with the - hypothetical ground motion, the effect would be to decrease
) the vibratory accelerations of building and equipment because of the shearing in the sand layer produced by the slipping.
Ground accelerations in the same direction as the slip movement would not transmit forces to the base of the building and, hence, the effective accelerations of the base of the building would be less than in the case where no faulting occurred. Since the hypothetical ground motion would produce vibratory stresses in structure and equipment that are smaller than twice the stresses produced by the design earthquake (Figure 1), the structure and equipment would naintain es- ' sentia11y their original strength. An aftershock of intensity equal to the design earthquake could, therefore, be resisted
, without overstress.
- i 36
E J122 9.?n C_I E 2 S $ d e_ S .i_ M E db 8 sib $ 1 $.d N Ess.sk U[2W 'h:1.fh i di.I j ! b : l r ., (7) Plant arrangements and design criteria as given in questions 1 and 4 will therefore assure shutdown and maintenance, of the Plant in a safe shutdown condition under the conditions assumed. 1 In the event of a conflict the information in this j 1 amendment supersedes the information previously submitted. ) l Subscribed in San Francisco, California, this 20th day of July, 1964. Respectfully submitted, PACIFIC GAS AND ELECTRIC COMPANY By ML 4 /J,: -
/ S. L. Sibley Vice President & Gener Manager '
3 l RICHARD H. PETERSON PHILIP A. CRANE, JR. Attorneys for Pacific Gas and Electric Company B 1 0 - fL8AA.9 Y Philip /A. Crane / Jr. I Subscribed and sworn to before me this 20th day of July, 1964.
/ .. .
,n
/ (SEAL)
Rita' J. Orepry, Notary Public in and for the City and County of j San Francisco, State of California 1 My Commission Expires July 16, 1967. ! j j 37 l
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