ML20234B065
Text
,
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July 16, 1970 l
ACES Members DESIGN PRESSURES FOR STgEL CONTA13 BENT VEg8ELS Section III of the ASW Better and Pressure Vessel Code
- allows the
" design pressure" of steel containment vessels to be as les as 9.9 of the " maximum pressure". (This seems to be a carryover from the codes for vessels which have safety values and which specify that these valves must have a capacity to limit tho' pressure to 110E of design pressure.)
Maximum alleeable stresses listed La the Asas Code era used in conjunction with the " design pressure" to determine such thtags as shall thicknees even though, during as laCA for esegla, the " design pressure" might he easeeded. This is acceptable for new frequency events according to section III as described in paragraphs
- N414(t) and 5-1320.
This suplaims the occasional aos of tuo numbers for a steel torus and dry well containment; that is 56 pois design pressure and 62 pois maximman pressure.- Incidentally, the apssected post-1DCA accideat pressures seen always to be lass than M pois.
R. Maccary assured the writer that the 62 'psig design pressure een-tataments designed under previous versises of section III and siso under Section VIII are the sans strength centstaments as the surrent ones.
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The following changes should be made in your Design Paransters lists to reflect these comments and to correct soms errors in the listes y
m e-L e gna - w -
n oyster Creek 1 MLas Mile Point 62 pois 0 175F (eld definition) 62 pois (eld definities)
Dresden Station Unit 2 62 pois (old definitima)
Millstone Point Unit 1 62 peig (eld definition)
Dresden Unit 3 62 pois (eld definition)
Vermont Yankee 62 pois (eld deitaitism)
Prairie Island 41.4 pois Eewounee l
61.4 pois Davis-lessa 36 peig oraragraph N-1312
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OFFl:lx nsg 9 D
EXCERPT FROM THE
- r-119th ACRS MEETING MARCH 5-7, 1970 March 17, 1970 EXECUTIVE SESSION l
I i:
3.
Isolation of BWR Instrument Lines which Penetrate the Contai Committee decided to schedule discussion on a " case" ba nment - The instrument line isolation question at the 121st ACRS meeting (M e BWR-subject to'the Subcommittee being ready.
ay 1970),
The hydrogen " case" would be given precedence if required..
prepare for this " case").
(The General Electric Subconnittee is to t
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g DEC11TEL PP.CSTRESSTD CONTAINISHTS
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The prit.aty porpose of the twe tings held with Eechtel Corporation i
reprecente tives in San Frt.ncisco, from Deceraber 16 to December 19, If;69, was to c.tteupt to resolve and cinrify with Lachtel the staff i
concerns specifically related to the Point Beach, Turkey Point and l
Ocon<. plants.
Thesc three plants represent the units now under i
revicu for FOL't: and follou the firco Bechtel prestressed contain-rie:it design tl'at u.es ured at Irlise? n.
The secondcry purpore uns i
r to examine the Ecchtel precf. rest,ed '6 containments as a generic item and at teinpt to answer concerns that are expressed at the.tino of review on all Bechtel prestressed FWlt containments.
The concerns discussed that related to the three specific plcnts under revieu for FCL vere resolve ( into four categories as follows:
g Bechtel Prestressed 2
Contain'unnts l 1.
Verbal - An explanation was given by Bechtel which adequately ann:cred the ques tion and was an item that did not directly require documentation.
2.
Voluntcry - An cnplanaticn vill be given in a revision of tbc FSAR which will bc submitted voluntarily with Ecchtel acting through and with the utility.
C 3.
Formal - An iten that requirec documentation due to a.
concultant 's quertion or beccuse it is on item of extreue it.por tance.
't 4.
Topicci - A subject that will be discussed end p:csegted in detail in the form of a Bechtel Topical Report.
These I
will then docur.unt the Dechtel corporate pocition on the subject which can be referred to on other future Bechtel projects.
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t Bechtel Prestrerced 3
Containments
!!any of tbc r.oncerns c:: pressed in the form of questions for l
l the three plcntn noted cbove have arisen from the ever-changing criterin and derj gn as presented by Lechtel. Although the two are inter-relcted, they will be separated here so as to briefly trace their evelutica.
priteris 1.
Basic structural criterie related to stress limits I
depart ed from existing codra in soma areas by necescity l
since there were no codes, strictly applicable to contain-l.
i inen t s. These criteria were initially studied in the review of the fi~ct fully post-tensioned containments (excluding base slab) (i.e., stress litaits in prestressed concrete i
under normal stresses cnd flexural stresses).
2.
Adherence to codes recommendations although there was not I
sound basis for it except that there was nothing better l
)
on which to base a design (i.e., shear design for prestressed i
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Eechtel Prestrear,cd 4
Containment concrete per ACI 318 and relience on beam tests).
3.
Departure from (2) sc new data and rationale were developed, (i.e., Mettock's revisions to ACI 318 equations for shear l
in prestressed concretc).
4.
Revision in the criteria for the araaunt of prestressing force applied to containment.
(Reduction from 1.5P to 1.2P i
for pre. stress and allowance of more concrete tension).
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1.
Change to a three buttress geometry instead of six buttresc.
l 2.
Change to large capacity tendons; in the order of 1200 to j
l 1400 hips working load.
I 3.
tiew liner plate materici (ASTM-A-285) with low yicid.
l 4.
Stiffened dome liner plate in order to eliminate dome truSSe5.
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f Dechte) Prestressed 5
Contaitonenta l'
5.
Use of 6000 psi concretc incread of 5000 psi concrete.
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s 6.
New tendon ccafiguration combining verticals into the
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douze end adding dora hoop tendons to make a two-way dome cycteu inrtecd of a three-wcy dome system, 7.
Uall and dorre thicknesses reduced by 3" or about 87..
(8.
Concept of design to allow for slip-forming and elimination of the ring beam.
9.
11aench at the bare sleb-cylinder wall junction deleted.
i 10.
Cont.ept of a double hinge in the cylinder wall near the e
base slab.
This evolutionary process hcs been shapad by the changing econo-mic balance of different approaches cs evaluated by Bechtc1, new research inforraation that has becone available, new analytical pro-cedurcs that have gained wider acceptance and neu construction materials and methods that have gained more common usage. Also some of the
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Eechtel Prcrtressed 6
Contafnt.nts chcage is ho; efully due to Ecchtc l's evoluction of a better way to solve the given enginecting problem.
Some of the op: cific iten:: of inport.nnce included in the dis-cussions ralating to the techtel centvincents were as follows:
1.
The principal stresses stated in the criteria have always 1
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been interprete.d as the criteria thct were to be used to judge the adcquacy of the concrete.
It var indicated, hm.ever, while at Pechtel that several of their engineerr do not interpret the statgenants in this manner.
It was also not possibic to verify that these principsi stres.scs i
e i
had ever been calculated and compared to the allowables.
Ecchtel vill be determining thne stresses in the future.
2.
It was learned that for the Bechtel prestressed containments, such as Point Beach, the critical shear stresses are caused i
by the post-tensioning forces and not from the accident or 1
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Echt61 Fres.trepred 7
Con t a irmen ts ear t h eptahe s.
3.
Continued concern was exprer. sed over the techniques and the artumptions. us.ed to analyze and proportion'the concrete anchortte zore of the tcadon enchoraces.
The ancelytical-techniquer, have not accounted for the three-dimensional strr as field and the effects of the thermal gradients and craching.
It was decided that varioun conservative assump-tions vill be mc'c as to crack locations in the c.nchorage 2.ca : cad au ent.1,'f.s h performed to chee!' whether the d
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I enchorage ccn ba maintnitred in static equilibrium.
This is to be carried out for the large tenden anchorages such i
I as on Arkansco 1.
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l Of the broad items to be addressed in topical reports Eechtel agreed to submit formal reports on the following items.
i 1.
Tornado Criteria - This would be a formal submittal of the i')
c bechtel h estressed 8
Con tain:. :n t s docunnnt received infort.:lly but with added information i
relr. red to the applicc.ticn of the$r criteria to design.
q It ucc not.ed that sample calculctior.s were a vcluabic k
twthod with which to illustrate the design application.
2.
Seismic Dersign - Thic document. would be a comprehensive i
discussion of the Lechniques used by Bechtel which are the o
response sp;etra tactbod and the tit:c-history v.ethod as well
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of their new thirting such as spectral power an aco duusity functicas which t.my be used in future applications.
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i Bechtel takes account of soil and struct.ure interaction i
using equivalent springn and analy::ds equipr'ent and piping 1
in one of three, methods, all based on a time-history i
i f
approach.
i Use the peak of the floor recponse spectra for the a.
particular clevation.
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i 1:.echtel Prectrent ed 9
l C o n t.i irc:.r.:n t s b.
Deraonstrate ths.t the ~ freqdency of the item is not on the pot.k of the floor responec spectra (plus or t.inus cotc error band) cud use reduced g-levels.
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I:stcraine the frequencies and determine the g-levels c.
h usina,the floor respontc crectra with an error band i
on Ihe cotr.puted Irequcncies.
Bechtui speciff ee fi>:cd valuas of horit.ontcl and vertical.
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4 acce krt. tion or t.w.f des the floor recp mse spectra in the trurchase specificatic le for 'the Clacc I items of e quipe:,an t.
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3.
Contaimtent Design - Itotes to be covered in this document would include the details of the large penetrations and the anclysea used, Lypical results of these analyscs, anchorage analysis and desi;;n details of how the shear forces are resisted and designed for, details on the i
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k Bechtcl Prestressed 10 Containnn:ntc structural calsecity and philosophy used in the tendon access gallery design, details of the analysis and design of int ernal structures in the contaitrtent and the major support systenn of r. jor equipment.
4.
Ter. ting and Surveillance - A report to present the pro-cedurcs to be used in the structural proof test, predicted e
reeults for the strainc in the instrumented containtnents as uell as deflections.
Defining the techniques and bases i
the.t will be use d tc jt.d;> tl.e adequacy of the containments i.
from the test recultc and the details of the items to be reported.
I' rom the surveillanec stc'ndpoint a program for tendon surveillance, the pressurn boundary surveillance, etc., will be discussed in detail.
With regard to the current ctatus of Arkansas I the following items are noted:
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I Bechtel Frcotrecced-11 Contain:' ats 1
1 1.
Teste on ti,e'EURV syttou bearing platen and vashers are k
being coupleted for the 1crce tendon system and in part 4
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have beca tubmitted to the staff for review.
These items have cico carried I;DT requirements.
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2.
The base olab conttcuetion ic co::picted at the site to the l
c): tent thct h chtel intended to place concrete in the first C
lif t of the w.llr. in tdiately cher 1 January 1970.
3.
Lut tree cen, uh5 ch 7.ro 1 u:1uded in thin first lift in the 1
l vall pour, have not beco jpproved due to the reinforcing
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not having been approved for the hoop tendon anchort.ca zone.
4.
Ecchtcl nr.d the applict.nt were to cdd more reinforcin;; for the louer liftn it. the unll to get DRL cpproval to proceed.
5.
The detaile of the added reinfo cing would be subaitted to DRL for the approvcl.
_ _ _ _ _ - - _ - - - - _ _ - _ =
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Lechtel Prectrorfed!
12 Con ta ir..< er.in i
6.
Ucchtel: and the cppliumt id11 later submit on analysis l'
of cachortge renes based on conservative assumptions of a crcched tr:chorcr,e zone.
This is to include hoop, dome' I
and top verlicL1 tendon'cnchor to;,ec.
1 7.
T<> date the 1u.teric! of item (4) hcVe not been received by DRL.
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1 Wit.h regt.rd to the current cratus of E,ncho Seco the following 1
i to:: are noted:
i 1.-
The plant CP ws.s incued based on the BBRV 90-wire system uith 6 buttresses and a 1.5P prectress icycl.
2.
Suppicuen t 1, post-CP, changed the contcin: cent to a
.)
3-buttrees structure,tuvived the prestress to 1.2P level and changed the tendons to c large tendon size l
1 such as 180-wires.
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t r:.chtel Prc'strasced 13 Containm ntc 3.
Supplet.ent 2, post-CP, etcted the.t the VSL syctem of 55-0.5"[ ct: ands with friction eachors with grease vould be vred.
4.
DEL ret;uec t c d added inf o);;.ation. by forral ques tions on the VSI. r:y: t em and thic uca not setir factorily supplied.
The ni'pliccat urs inforued that "on the basis of the information you have subnitted to date, we cannot conclude thet the propost d char:-
from the E.BRV tendon system to I
the V6L tcu'en cyr te m h c be en jus tified."
j 5.
The conntrection at present has cort.pleted one-hcif of the bcse slab with the VSL at;chors cnbedded in the slab for the bottoa vertical anchorn.
Thic ucs done without DRL approve:1 and at the applicant's risk.
6.
Additional tests are being completed by VSL and more infor-mation is due to be received by DHL.
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'OfflCT. USE OM.Y 0
Execrpt from the Summary of 115th ACRS Meeting Nov. 6-8, 1969 I
i Meeting with the Division of Reactor Licensing
.....a.,
3.
Inertine Review (Peach Bottom, N. S. SAVANNAH, Humboldt Bav) - The C0 Staff has discussed with Peach Bottom, N. S. SAVANNAH, and Humboldt Bay whether the requirement to inert the reactor containment resulted in any
. problems. They all reported that there were no access problems and that the frequency of inspections was the same as if there were no inerted environment.
No pocketing of nitrogen has been observed when deinerting for access.
In at least one case inerting has reduced corrosion in the containment.
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0FFIC' At USE ONLY 1
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l REPORT OF NOVEMBER 7, 1969 TO ACES CONCERNING POSTACC1 DENT MONITORING INSTRUMENTATION T. G. Robinson and C. S. Walker pgg l
ORNL Instrumentation and Controls Division TOPICAL OUTLINE-I.
Introduction A.
Monitoring with respect to the loss-of-coolant design-basis accident B.
Plants: Ginna, Oconee I, Palisades, and Browns Ferry (see Attachment 1 for pertinent data)
II.
Application of Efforts A.
Major emphasis on instrumentation that monitors the effectiveness of the engineered safety features (ESP)
B.
Lec emphasis on instrumentation that controls and monitors operation oi she individual ESF systems C.
Determination of availability of above types of instrumentation D.
Determine 'on of adequacy of above types of instrumentation III. Possible Instrumentation to Detect Loss of Core Integrity A.
Temperature measurements at bottom of pressure vessel B.
Monitoring of radiation inside containment system C.
Heat balance determination D.
Level measurement of coolant in reactor core IV.
Tabulation of Instrumentation Found, with Comments (see Attachment 2) l V.
Conclusions j
A.
Instruments.tionformonitoringUnvironmentisavailable I
B.
Instrumentation for detecting loss of core inte6rity is unavailable
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C.
Consideration of usefulness of monitoring of gross accidents
{
1.
during course of accident f
2.
postaccident examination l
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- s.
NUCLEAR POWER PLANTS INCLUDED IN THIS INVESTIGATION Nuclear Power Plants R. E. Ginna Oconee I Palisades Browns Ferry Type PWR PWR PWR BWR I
Power,Mv(e) 480 839 700 1098 i
Date of Construction h-66 11-67 3-67 5-67 Permit a B&W" CE GEb b
Designer Westind; house b
Architect Bechtel Utility Rochester Gas Duke Povera Consumers Power Co.
TVA"
& Elec.D of Michi anb 6
Site Ontario, N.Y.
Seneca, S.C.
South Haven, Mich.
Decatur, Ala.
" Organization visit'ed.
bOr6anization contacted by telephone and letter, s
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l 8-DETAILS OF ITEM IV
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'l Categories of.Postaccident Instrumentation A.
Instrumentation used to monitor effectiven'ess of engineered safety features in controlling environment of containment cystem and radio-l active release from containment system:
1.
Containment pressure-2.
Containment radiation h.
Heat balance on heat-removal mechanisms 3.
Power level of reactor 5
Hydrogen and oxygen concentrations j
6.
Area-radiation level (building enclosures) 7 Environmental radiation level (field and stream) l 8.
Stack radiation monitor
~
9 Meteorological conditions
-B.
Instrumentation used to control and monitor operation of individual engineered safety features:
2.
Containment spray systems 3
Containment isolatier nystems h.
Containment air cool og and cleaning systems 5
Secondary or equipment cooling systems 6.
Reactivity control systems (boron injectioh) 7 Auxiliary and' emergency power systems t
Specific Instrumentation and Comments -
A.
Instrumentation ilsed to Monitor Effectiveness'of Engineered Saf'ety Features in Controlling Environment of Containment System and Radio-active Release from Containment System 1.
Containment Pressure a.
Ginna.
6 transmitters; 3 have range o to 60 psig and 3 have range O to 90 psig. 'For operator information (not part of ESF), an additional transmitter with range -3 to +3 psig is provided.
b.
Oconee. h transmitters; 3 have range -5 to +15 psig and 1 has range 0 to 90 psig.
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Pa2irades.
4 transmitters; 2 have range 0.to 100 psig and 2 have range O to 5 psig.
d.
Browns Ferry.
4 transmitters with range 0 to 100 psig.
Note: All above transmitters have control room indication.
COMMENT: We believe that two vide-range channels should be displayed to the operator. The availability and adequacy of the above instru-mentation is otherwise sufficient.
2.
Containment Radiation a.
Ginna.. Employs continuous sampling by withdrawing air from the containment system through a particulate monitor and through a gaseous monitor and then returning it to contain-ment. system. Range of particulate radiation, 10 ' to 10 '
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pc/ce; range of gaseous radiation, 10-6 to 10-3 pc/cc.
b.
Oconee. Employs a coaxial ion chamber inside containment system packaged with a solid-state preamplifier and shielded from contact with steam environment. Range 10 " r/hr to 10"
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r/hr.
c.
Palisades. Employs a unit with chamber in containment system.
Range 10 " r/hr to 10" r/hr, d.
Browns Ferry. None in dryvell, but a unit is employed in the secondary containment building over the fuel storage pool.
Range 10-" r/hr to 10 r/hr.
8 Note: All units have control room indication.
COMMENT: We recommend a sealed access penetration near the entrance through which a survey monitor might be inserted for personnel protection when they reenter the containment vessel for cleanup. We believe the containment radiation monitor should have sufficient range to register a TID-lh8hh release and survive such a release for the postaccident period. A determination should be made as to what constitutes a. sufficient range. We also believe the above sampling-type instrument vill survive the accident environment, since it is external to the containment vessel.
3.
Power Level of Reactor No one is villing to claim that nuclear instr [unentation vill survive the loss-of-coolant accident environment. This is also true of the control-rod position indicators. Cables may be susceptible in the first case, and forces generated during blovdown may silence the position switches. No effort was made to evaluate this.
h.
Heat Balance on Heat-Removal Mechanisms All reactors units have u.eans for determining a heat balance on the reactor. coolant side of the low-pressure safety-injection-system heat exchangers, but three different indicators are required in addition to a calculator or slide rule.
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3, The PWR's,with the exception of Palisades, have the above in-strumentation for a heat balance 'on the service-water side of the air coolers in the containment vessel. The BWR units shut down their air coolers with the safety-injection si nal.
6 COMMENT: We believe that an analysis should be made to determine the possibilities of heat balance for diagnostic purposes.
5 Hydrogen and Oxygen Concentrations Sampling equipment is or will be provided. The' details have not been fully determined ~in all applicable cases.
COMMENT:
It would appear that as long as the radiation in the sample is contained and does not affect the analysis, this type of equip-ment should be available and adequate. We made no effort to evaluate the adequacy of this instrumentation.
6.
Area Radiation Level All reactor units have area radiation monitoring for operating l
personnel protection. The range of this equipment is not in-tended to include accident conditions in most cases.
COMMENT: We have not evaluated the adequacy of this instrumentation.
It appears that sufficient availability exists.
7 Environmental Rediation Level All reactors units have radiation. monitors within and at the plant site boundaries. Additional units are usually found at a 10-mile radius. Dosimetry is also provided on some grid basis both for ground-level detection and for purposes of ecology. We have not investigated the adequacy of this instrumentation.
8.
Stack Radiation Monitor a.
Ginna. Vent is isolated from containment vessel.
b.
Oconee. Penetration rooms are maintained at slight negative pressure following the loss-of-coolant accid,ent. Exhaust is filtered, monitored, and released through vent until radiation monitor signals that radiation level in released air is too high.
c.
Palisades.
Stack is isolated from containment vessel.
d.
Browns Ferrv.
Secondary containment system is maintained at sli ht negative pressure. Exhaust is filtered, monitored, and 6
released through stack until radiation monitor signals that radiation level is too high. No effort was made to evaluate this instrumentation.
b
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h 9
Meteorological Conditions All operating utilities report that ground-level temperature and some elevated-temperature measurements are made to determine the i
possibility of a temperature inversion. Wind-direction and I
velocity-detection equipment is provided at plant site. No effort was made to evaluate this instrumentation.
B.
Instrumentation Used to Control and Monitor Operation of Individual Engineered Safety Features 1.
Emer6ency Core Cooling Systems All reactor ' nits have injection-pump output pressure and injec-u tion-line flow monitored in the control room. All valve positions involved have their positions monitored in the control room.
In addition, various coolant levels are important in the proper opera-tion of the emergency core cooling system.
a.
Storage Tank Level t
Ginna.
2 independent level instruments inform operator to
- manually switch suction to containment sump.
Oconee. 1 level instrument informs operator to manually switch i
suction to containment sump, l
Palisades.
2-out-of-h level switches provide signal to automati-cally switch suction to containment sump.
In addition, 2 of the above h transmitters are monitored in the control room and each has high-and low-level alarms in the control room.
Browns Ferry.
1 float indicator is located on tank with high-and low-level alarms in control room. Also, 1 level transmitter is monitored in the control soom with high-and low-level alarms.
There are 3 tanks in this system and each tank has the above in-strumentation. The 3 tanks are interconnected in 2 independent header systems.
b.
Accumulator Tank Level and Pressure
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Ginna (600 psi).
2 independent level and pressure instruments monitored in control room on each of 4 tanks.
Oconee (600 psi).
1 level and pressure instrument monitored in control room on each of 2 tanks Palisades (200 psi).
1 level and pressure transmitter provides monitoring in control room for each of h tanks.
In addition, each tank has one level switch for high-level alarm and one level switch for low-level alarm. The level transmitters also provide additional high-and low-level alarms in control room.
c.
Containment Sump Level Cinna.
2 independent level instruments monitored in control room.
Oconee. I level instrument monitored in control room.
Palisades.
2 independent level instruments monitored in con-trol room.
Browns Ferry (drywell).
1 level instrument monitored in control room with high-level alarm.
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d.
Core Coolant Level Ginna. None i
Oconee. None Palisades. None
, Browns Ferry.
2 independent level transmitters with control room indication.
i COMMENT:
We believe the pump discharge pressure, injection line flow, and valve position instrumentation is available and adequate for the monitoring and control function required.
Since the storage tank level is pivotal in the realignment of the low-pressure system, we recommend that the storage tank system have 2 independent level instruments monitored in the control room. In addition we recom-l mend that the core coolant level in the reactor vessel have 2 in-dependent level instruments monitored in the control room.
2.
Containment Spray Systems
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1 The PWR's have pump outlet pressure and injection-line flow moni-i tored in the control room, except in Ginna. The spray-pump outlet pressure is indicated locally at the pump. Flow is monitored in the control room following the initiation of the recirculation phase when pump suction is transferred from storage tank to low-pressure-system i
heat exchan6cr output. This then provides the operator with the flow.
division between reactor cure and containment atmosphere.
The BWR j
unit has manual initiation of this system.
COMMENT:
We believe this instrumentation is available and adequate only when it is continuously monitor,ed in the control room.
1 3.
Containment Isolation Systems The positions of all valves involved in this system are monitored in the control room.
COMMENT: The adequacy of this system rests entirely on the ability and the l
importance of the valve position switches inside the containment surviving the loss-of-coolant accident environment. We made no effort to evaluate this.
k.
Containment Air Cooling and Cleaning Systems a.
Ginna.
h independent units where 2 have charcoal filters.
b.
Oconee.
3 independent units (no charcoal filters).
c.
Palisades.
h independent units (no charcoal filters),
d.
Browns Ferry. Shutdown following loss-of-coolant accident, j
Note: The PWR's have a radiation monitor in the service cooling vater discharge header to detect lekks from the containment-vessel.
Leakage into the containment vessel, if large, may be detected immediately by flovmetere and, if small, will have to be detected by sump boron-content analysis. In Ginna cach charcoal filter has 2h temperature sensors for detecting excessive heat rise due to decay heat of the fission products. A dousing system is provided in this case with supply taken from the containment spray system.
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6 COMMENT: We believe that an. analysis should be made to determine the poshi~
bilities of heat balance instrumentation for diagnostic purposes.
Where charcoal filters are employed within the containment, we agree with the practice in the Ginna plant, that some qualitative i
means should be provided to advise the operator of air flow in the appropriate ducts.
5 Secondary or Equipment Cooling Systems This category is intended to include secondary cooling systems such as pump-bearing coolers and pump-room atmosphere cooling where needed.
Time did not permit an evaluation of the availability and adequacy of this instrumentation.
6.
Reactivity Control Systems (Boron Injection) l Equipment is available for sampling and analyzing the boron content i
of the recirculating coolant in all PWR units. The Ginna unit has concentrated boric acid tanks used during the injection phase of the loss-of-coolant accident.
Two independent level instruments are provided with control room indication.
COMMENT: We made no effort to evaluate the adequacy of this equipment. The availability appears to be satisfactory.
7 Auxiliary and Emergency Power System No effort was made to evaluate this, i
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Purkinn PB-1 Containment entry permitted.after six hours with-an oxygen concentration' greater-than 177..
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Purning.
NSS Primary system.cooldown and effluent monitoring requires more time than purging operation.
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l Purging; 'Humboldt -
Primary system cooldown and effluent monitoring.
requires more time than purging operations.
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Makeun PB-1
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Loss is primarily attributed to. evaporation from storage tank. Tank also supplies liquid nitrogen to traps in purification system.
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Entry Provisions PB Routino:
Requires tuo men,-including one H.P., permitted after six hours using.
an air pack, protective clothing, oxygen --
concentration greater than 177. and monitor--
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ing for oxygen and radiation.
Emergency Entry:
Requires three men, in-cluding one H.P., uith one stationed at air lock and 30 second communication between all parties. Requires an air pack, protective clothing, o::ygen concentration greater than 177 and monitor for oxygen and radiation.
Entry limited to inspection only with 30 minute stay.
6/
Problems PD-1 An inert gas generator (IGG) 7as tested prior to initial startup which resulted in a change to the liquid nitrogen system.
The IGG removs 02 by reaction with propane in a combustion chamber. The exhaust gases were cooled by a-freon refrigeration unit. An analysis of ex haust gases revealed N, 00 H 0, NO, NO2, 2
2 2
CO, S0, 02 and soot.
It was difficult to 2
maintain an optimum combustion efficiency and an analysis of condensate found on cold surfaces inside containment indicated a ph of.1.9. ' Personnel also observed evidence of incipient corrosion on liare iron parts, such as flanges, piping and deck plate.
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