ML20151A256
| ML20151A256 | |
| Person / Time | |
|---|---|
| Site: | 05000214 |
| Issue date: | 09/28/1965 |
| From: | Fraley R Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20151A233 | List: |
| References | |
| FOIA-88-184 NUDOCS 8804060404 | |
| Download: ML20151A256 (154) | |
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5-69 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC gNERGY COMMISSION WASMiksetOM. D C. SeHs September 28, 1965 MEMCRANDUM To : ACKS Members from : R. T. Traley xecutive Secretary ACRS
Subject:
MALIBU NVCLI AR PLAST, UNIT NO. 1 . QUESTIONS RELATING TO SEl$MIC DESIGN RAllID BY TKI ATOMIC SATETY AND LICEN51SC ,
BOARD The Ate-ic Safety and Licensing leard has recessed the Malibu Hearing until
- October 19th with an evtline of soee of the matters that have occupied the Board's concern. This tutline, sumnartted in the attached list, was sub.
eitted to the participants "for consideration, comrent or submission of evidence if they dee* tt necessary, in the October sensten." E. C. Case has indicated that he w?vid like to discuss item a wtth the Committee during the October ACR$ Meet t*g to determine if the absence of an active fault at the Malibu site was an ACR$ requirement, as interpreted by the Nearing Board, or eerely a statement of the information presented to the Comrittee at the 56th ACR$ Meeting (July 196a).
It should be noted that the draft VSC&CS report considered at the $6th meet.
ing included the f ollowing statement "All of the known surface ground displacement on the Malibu Coast fault tone is prehistoric . that is, more than 200 years old. If the band of deformed rocks just south of the Malibu Coast fault trail is considered to be part of this fault aone, the most recent ground displacenent occurred sometime between about 200 and t.00,000 years ago,
The likelihood of ground displacement at the site due to earthquakes depends on the frequency and severity of earthquakes along the Malibu Coast and related Taults.
"The Malibu Coast f ault is considered to be part of an active system that includes the Newport Inglescod tone. Only 3 to $ magettude shocks have been associated with the Malibu Coast fault; none of these has resulted in known displacement at the ground surface in histeric tiet. However, in prehistoric tiae f aulting at eight known localities along the general trend of the Maltbu Coast fault has displaced rocks no older than 400,000 years.
It can be inferred frem these data thag similar f aulting a.ay have occurred 1
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4 within the site but the displacements have not been detected because of 4 geserally poor exposures. On the basis of this record the probabtitty
, of ground displatement at Corral Canyon in the next SO years is very low." r the Cometttee in its letter of July 15. 1964 interpreted this report and discussions at the meating as follows:
"The Com'tttee was informed that the geology of the site was suitable for the preposed construction. It was reported that no active geological faults are present at the site."
J The CSC&GS report considered by the Coenittee at the 60th meeting (January 1965) tecluded the following summary:
"The Corr 41 Canyon site is bisected by the east trending Malibu Coast f ault, which is part of a wide, east trending tone of north dtpping faults, asy9eetric folds and shears. the Malibu Coast Zone." "Evidence indicates ,
that the tone is active en a regional scale." -
1 "lased on aeatlable ge
- logical evidence the probability of pereanent
- displacement on the gre.ed surface by faulting in the Corral Canyon site j during the next 50 years as negligible. Setsmic shccks can be espected at 4 the Carral Canyan site. re than $a seismic events of P.agr.itude ; to 6.3 have been recJtded sat.t'. 62 miles of the site in the past 112 years."
The Cune.i t t e e , t r. t t , letter of January 25. 1965. "retterates its beltet that the proposed Faltb. Swelear Plant can be constructed with reasonable I assarance that it can be operated at the site without undue risk to the i health and safety Jf the pabitc." i Expletatory trenches sere dug at the site during April 1465, at the sugges.
I tion ni tne A$tt and a subsequent USC&CS report was tssaed in July 1965.
I three copies sere )t04tded to the ACR$ as a Catetsry b Report. (This s t pe r t wa s su9r.a r t t e d in a mema b.s M. C. Caske which was distributed to all ACR$ Me-bers or .'.ly 26. 1965.) Tne following comreets were inelsdsc ta this report:
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"Taults of several eagnitudes are present in bedrock cf the C:rral Canyon site. The Halibu Coast f ault. About 800 f eet nerth vi the reactor ,
] location is of regional significance and large magnitude of displacement; i i where well exposed. Its trace is marked by a mone of brecciated and sheared j rock as euch as 75 feet wide. Taults of lesser eagnitude such as f ault t A near the north boundary of the plant, separate enflerent foreational i i units and are charactertted by aenes of sheared and brecciated rock up to (
j seseral feet wide. Such f aults can be traced for only hundreds to thuu. ;
sands of feett they probably have displacements of handreds el feet. (
Intraf or-attanal f aults, su:h as f ault T. exposed in Corral Creek and l j trench 3 (the teactor. location trench), are character 1 ed by local truece. !
j t i o?. of str.cture ard are consenly earked by thin, but rec.gstrable eines 2
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6-71 of shested rock or breccia. Such features can be traced tens to hundreds of feet; their displacements are probably on the order of tens of feet.
Finally, innumerable shears locally continuous or concentrated in narrow bands, pr~ wade the oudstone of the Corral Canyon site. Minor displacement has occurt ' on these features, as indicated by disrupted sandstone beds and slickensides. Aggregate displacement across several feet of such sheared rock may amount to several feet. All demonstrable fault movement in the Corral Canyon site is pre.Recent (more than about 10,000 years) in age."
"Comparisons of degree and time of deformation in different parts of the fault system indicate that future faulting is at least as likely to occur in the Malibu Coast zone as in any other part. The available seienic record is not sufficient to establish the recurrence interval for large. magnitude f aulta in the system; this interval is greater than the approximately 200 years of historic time and it may exceed the approx 1 r.ately 10,000 years of Recent time. As this recurrence interval is large compared to 50 years, the probability that a large. magnitude shock with .
center near Corral Canyon will occur during the next 50 years is very low.
This very low probability, coupled with the lack of evidence for surface f aulting in the Malibu Ceast sone during Recent time indicates that the probattlity of permanent displacement of the ground surf ace by faulting at Corral Canyon during the next 50 years is very low (this same very low probability was described in the U. S. Geological Survey report of 1964 as negligible, which was used there in the sense of very low). This assessment implies nr- j.dgment of public risk; it is not intended as a judgment of the conseqwences of surf ace f aulting in any particular utili.
sation ef the Corral Canyon site."
It should b( n:ted that fa.it F is about 35 feet nortt. east of the center of the reactor building which means that it passes under the reactor con.
tainment structure.
E. Case will be prepared to discuss this information in more detail.
Attachment:
Summary of Matters IdentifLad by the Malibu ASLR dtd. 9/27/65 i
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Su msry of Matters identified by the Malibv Asy ,
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- 1. Since the containment building is not specifically Jesire J to withstand ground displaceeent, has it been estabitshe. aiw on what basist (a) what is the ground displacement that it can' {
withstand; (b) what would be the amount of fission producss ro. ;
leased if a displacement greater than that identtiled in (a) :
1 occurredt and (c) what would have to be provided in the design to give such resistaneet ,
) De Board suggests that experience with relative ground movetent '
f be used as the basis in the Southern California area for the t j selection of useful values in these parameters. 5
- 2. Is it acceptable to grant approval on the basta that the structural l regwirements are "within the range of accepted practice and estab. ;
11shed knowledge" even though the detailed design has not been i j presentedt '
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- 3. What is the meaning Si the phrase "without undue risk to the health i
and safety of the public" as understood by all participants, especially the word "undue"? References to appropriate authority
] are requested.
i i 4 De leerd interprets the ACR$ Report of July 13, 1964 as carrying
) "the admonition that this reactor should not be located over an i
' active fault." W e Board requests standards or suggestive standards to measure an active f ault.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION W ASHINGTON. D.C. telet July 24, 1968
$UMMARY 99th ACRS MEETING JULY 11 13. 1968 WASHINGTON. D. C.
SPECITIC JROJECT
Its repett tc the Chairvan, AEC, concluded that, with the resolution of several itens, the propossd platt could be built with reasonable assurance tr.at its op4 sti:n would rot pose an undue risk to the health and safety of the public.
Tra itees to be resolved during construction are:
- a. Design of the off-site power supply to assure that no single failure vill intatrupt power to the plant.
- b. The contrcl rod drive power supply design.
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- c. Matters previously identified as verranting careful considers-tior. for all large, vattr-cooled, power reactors.
9 Thett was an extensive discussion of the possibility that thermally-induced shock could fail the reactor vessel on initiation of the ECCS. The applicant believed that the approach taken had been a conservative one, but there still remain issaes to be resolved with the Regulatory Staff concerning the heat transfer and fracture mechanics models, and the materials data to be employed.
Ar. altert.ative approach, in the event that final results are unsatisfactory or incon:1usive, would be to heat the injection water, although the ef fect of this on the injection tank design and on core cooliat had not been investigated.
o The containment design was unusual in that no diagonal reinforcement was included, garthquake-induced shear forces would be withstood by a combination
< of rod devel and aggregate interlock action. After a lengthy discussion, the j committee concluded that the design was acceptable because of the low seismicity of the Maine Yankee site, i
In response to questions from the Committee, the applicant indicated that they vould calculate the Yanket (Rove) pressutt vessel fluence and the Connecticut
, Yankee modarstor coef ficient using the analytical techniques used for the Maine Yankee plant.
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- 2. Hanford N. Reactor The Cosunittee completed its review of the N. Reactor Ef fluent Centrol Project.
The repert to the Chairman, AEC, states that the ACRS believes the modifications involved represent a significant step toward the goal of operation under the same criterin as are applied to coemnercial power reactors and recosamends imple.
mentation of the pccposal.
In additien to reviewing the modifications to the smergency Raw Water and Confinement systems, the Coasnittee also discussed several other areas where changav might be appropriate. It was noted that decisions could not be made with respet.t to these items until a ceci11on is made regarding the extent to which the N-Reactor sheuld be made comparabia to comunercial power reactors.
Questiens were reised on the assurance of proper functioning of the confinement closures which prevent overpressurization of the building. It was noted that pressure relie f is possible through the refueling canal "water seal", but several members felt that the question had not been coupletely resolved.
It was als: suggested that DUN perform a critical review of the plant protection system in light of the IEEE standstd for nuclear plant instrumentation.
The Chairmn observed that an additional review would be necessary af ter all further studies had been completed, including that of conformance with the 70 ;
General De sign Criteria. l t /
! 3. Rancho Seco l
The Co::cittee completed its review of the Rancho Seco Nuclear Station. The I Connittee's report to the Chairw.an, AgC, states that the Coomittee believes that the proposed plant can be constructed with reasonable assurance of its l i
t operation withcut undue risk to the health and safety of the public. The report also noted several items of concern which the ACRS believes, however, can be resolved during construction. These include
- a. The design of the protection system to cope with systematie, non-random, concurrent failure of reirent devices.
- b. The prediction of integrated fast flux at the teactor vessel vall.
- c. The adequacy of the prcposed pressure vessel material surveillance program.
- d. Metters, previously identified, which warran't careful consideration by manufacturers of large, water-cooled, power reactors.
There was particular concern during the Comunittee's review over the lack of experience of the applicant with either nuclear or fossile-fueled steam plants.
It was noted that the applicant has already begun staf fing with experienced personnel and intends that key members of the operating organization be on hand !
very early.
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'e The use ef a 10' yeer visc in estabitshing the plant structural design basis was quartiened. It was t: tad that the curve of wind speed vs. frequency of occurrente is ratter fint ar.1 that the seismic loadings would govern, in the design, ever. if extremely higt wind 6 vers assumed. It was agreed that a wind speed havits a probabilth of 1/1000 cr 1/10,000 years would be used since it makes littit istfarer.:4 t- the actual plant design.
The design ci tts rad.warte systez was considered at some length, particularly with respe:t tc tra ba:1du; att d sp:sti of tritium. The applicant contended that the Rar. h; 5 .:: 0$stgr van not different in kind, in this respect from other PWR desig'.s. A: Rar:h. Se:c, a:onomic conaidarations heve led to the recycling cf demi:4t sita d wetet and the re-use of boric acid, which make for sore dif fs:en.,1. the negr e e cf r ad.wsste treatment undertaken.
An eres c f :c .:::: ir :r.*t: tion with this project was the fact that power lines t; ths Rt :ts Sa:: pier.; are P3&E lines, and, therefore, under the contret cf the :ast ::.;c:h4r. The agglicant indicated that this should not redv:4 the si,;.sti :( !f- a t t e p ower to the p'. ant.
- 4 Zitn 5'.e't:-
The Oo:rit:te, ce s:1uiet 1 5 d6*. iteration en the Zion proposal during an exten-sio . ef tr e 9's". :4 t ' .. g - July 21, 1968.
Items tr at r.; steed ;3:ti:.lar a; ten:1:. daring the Committee's discussions t W'f*:
- a. Tte lo.k cf st;ainti:- :f cor. trol and safety functions in the pist: 1. st r ar.e- i s i r . Censideration of this item is to be
. co ?te.u+d wet- DE. hs t ::mpleted the review regarding diversity as well as e. par ati:: trat wie requested durinF the 98th ACRS
' me t - ir.g .
- t. A requiram4:.t it: the fa:ility to design to cope with limited sitt leaka s:d eritts in the pressure vessel vall, particularly in !).* r4gior. tt.at re:tives the highest neutron irradiation dose.
Dr. Okrent appe:d6d saditic:.81 rensrks to the Committee's report on this project q
regardingt reselution of qs actions related to thereal shock, periodic taspection for prinary sy ster # , derign provisions to cope with a leak or split in the pressure vessel vs'.1 and adittio al conservatise in several aspects of the plant design er.d accident evaluation. The Camnittee, in its report, observed that Dr.
Okrent's comments had t een censidered during the review and the Committee believe that the status tf these matters, as they pertain to the Zion units, is satis.
factory.
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- 5. ACRS Repert on Renitor Freesure Vesseis/ Location of Reactors at Sites with Greater Populetter. Density that. Indian Point . Zion :
The Ceamittet dis;ussed several preposed reports on reactor pressure vessels that dealt primarily with the requirement for design of reactor facilities to cope with rse:ter pressure vessel f ailures. It was concluded that this mattar steuld be in:orporated in an ACRS report dealing with siting of reactors at lo:ations ne:e dansely populated than Indian Point - Zion. A draft report :
I was prepared. Additisnel discussion is planned.
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MrEi*.O WITH THE DIRECTOR OF RgGJ1ATION AND MEMBgRS OF HIS STAFF Meetina with the Director ef Regulatten
- 1. Bolsa Island geview Mr. Price raised the geestion of how best to come to grips with the Bolsa Island review at this time. It was agreed that the ACRS subcommittee on
- Boiss Island would hold itself in readiness to pursue the matter and that Mr. Price wculd determine whether the applicants were ready to proceed.
A reeting of the Bolse Island Subcommittee, including the Committee's censultants on the project, will be held at the site following the August ACRS meeting. The Committee's approach to the Bolsa Island proposal will 1 be dis:ussed ir.ternally during the August reeting. Discussions with the
- Regula: cry Stef f are also being scheduled.
- 2. Subpoera l4tter Mr. Pri:e noted that the Commissioners were aware of the Committee's desire for en agreement cot.:erning subpoena of ACRS members by Hearing Boards. He added thtt the most recent draft of the agreement had not yet been prepared
, for the Co:cissior , hcwever.
' 3. persona.41 Requirements of DRL Mr. Price reported that he had discussed personnel requirements with the ;
I Bureau of the Budget in light of the cutbacks recently ordered by Congress.
! i They recognize the Regulatory Staf f's needs and will rot force staf f reduc.
tions, although additional personnel may have to be added by way of reducing the General Manager's staf f.
I Corpliance Divisi:- Repcrt j i
- 1. Oyster Creek 1
Coepliance reported on the status of pressure vessel repairs. In connection with the procedures being used to control dimensional changes in the stub tubes during velding, Dr. Bush noted that it would save time if profile measurenents had been obtained befors the Committee again takes up the review of Oyster Creek,
! 2. Nine Mile Point I
l The Staff reported on the proposed repair procedure for the stub tube welds which have been found to have less bonded area than is acceptable. taare was some question as to whether the simple addition of more veld metal to meet the code's "fusion length" requirement was acceptable. The Regulatory Staff felt I l
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that it was and would meet the code requirements. Dr. Bush obseriod that a high percentage of similar velds suffer from the same defects, but 'Obe welds in the Nine Mile vessel have been inspected with a technique which allows their detection.
- 3. LAC 3WR i
The emergency condenser steam line isolation valves which failed under test 1 were found to be standard valves, used in an appropriate environment, and for the proper service. The cause of failure is still unknown and the situatioe ,
is being investigated by the valve manufacturer.
4 Elk River A high pressure test of the pressure vessel resulted in an accumulation of water in the vessel cavity which was apparently proportional to the pressure.
However, ner.e of the tracer (volatile and non volatile) which had been intrew duced into the vessel was found in the cavity and the applicant has concluded that the vessel is sound. A report is to be submitted.to the Regulatory Staff for evaluation.
- 5. Cinna T-1 steel, which was found unsuitable for a large reinforcing ring in the Ginna containment, has also been used in fabricating supports for several l 9
major components. This is not a code violation, although there is some feeling that it to poor practice. The Staff plans to evaluate the structurst tests carefully. l l
Caterary B Items . DRL Report
- 1. Big Rock Point - Change No. 14 1 Kr. Skovholt reported that all of the powder cospect fuel elements being tested in the center-melt program had failed. One of the "intermediate" bundles was found to have a rod with blistered cladding. Hydriding of the cladding is i suspected. The licensee is still obligated to keep the advanced-burnup bundles out of the core until investigative results are available but this is not expected to hamper the program too much. l l
- 2. Plum Brook Reactor i There was some question over high radiation levels reported at the Plum grook i site. The report of surface contamination to the level of 73,000 dps/en I actually involved a chemistry lab hood and was the maximum found anywhere on I
- site. No personnel contamination resulted and there was no airborne activity detected in the lab as a result.
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- 3. Palisades The DRL review of ECCS and reactor protection systes design has been started.
The protection system is similar to that reviewed in connection with Fort Cal-houn and is, therefore, in fairly good shape. The ECCS design will include two independent cooling systems and will be capable of tolerating any passive failure. The Staf f inquired if the Committee desires to review the proposed
- designs but time did not permit discussion. ,
(DRL plans to continue its review and will report to the Committee during the
- August meeting.)
- 4. Waste Discharze at Multi-Unit Facilities Dr. Morris noted that the Regulatory Staf f had begun working on criteria covering the ef fect on the environment of multi unit sites and of many separate plants close to the same population center, or affecting the same area.
, Criteria for In-Service Inspection - DRS Report Mr. Case reported that the N-45 Code Committee and DRS were close to an agree-ment on a draf t set of criteria for in-service inspection. There is still some disagreement about providing access to the beltline of pressure vessels and
', performance of periodic hydro tests at high pressure. Mr. Case felt that if he
' cculd provide the Code Group with some assurance that inspection requirements
! established for future plants would not be required on existing units, a set of
' :riteria could be issued quickly.
j The Committee later voted the following positiont With regard to the implementation of the proposed in-service
] { inspection standard, the ACRS believes that routine in service inspection requirements of reactor pressure vessels in use or under construction should continue to reflect recognition of built in limitations on the inspectability of such vessels, <
l except where evidence of pressure vessel deterioration dictates l othe rwise .
(Mr. Case was informed of this position of July 15,196&)
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EXICUTIVE 3E551(21$
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- 1. Appointment of L. $ quires During the session on July 11-12, 1968, it was agreed that a further effort (at the highest level) would be made to obtain the approval of the du Pont Corporation for Dr. L. Squires to serve as an ACRs member. Mr. Mangelsdorf agreed to pursue the matter with du Pont representatives. On July 21, 1968 Mr. Mangelsdorf reported that he had been informed that Mr. Squires had con-vinced himself that he was available for an appointment to the ACR5.
(The Comissien has been notified of Dr. Squires' willingness and availsbility ,
for an appoititment to the Consnittee.)
- 2. In-House Revisv of Regull ., Trogram.
It was reported that Mr. 4sdorf has been nared Chairman of the recently constituted panel to revi: .he AEC's regulatory program. Tne Coreittee't representatives on the pan.1 agreed to keep the ACRS informed of the groug's activities by an oral report at each ACRS meeting.
- 3. Article by Dr. J. McKee in "gnzineering and SciencA",
California Institute of Technology The Coccittet discussed the recent article by Dr. Mc~ in antineerint and l In particular, there was some concern over the use made of the
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' article by the press, etc. There seemed to be a consensus that no action should be taken with respect to this particular matter, though there was some feeling that this type of publicity should be taken up with the. Comis.
y sion in tennection with urging a more vigorous carpaign to gain vore wide-spread public acceptance of nuclear power.
4 Discussion by Dr. P. Paris on Pressure Vessel Failure Dr. parts presented an analysis to the Comittee which indicated that small j flaws eight trisger fatture of pressure vessels during normal operation as !
they approach the end of-life condition. The Concittee decided to ask Dr.
Paris to prepara a more couplete discussion containing the assumptions and development of his analysis. (Dr. Paris has agreed to provide this report j
by Septerber 5,1968 or sooner, if possible.)
- 5. Reactor Safety Research Workina Groups Dr. tabel announced the formation of working groups which will follow items of safety research topical in nature. This would include reactor vendors' safety research programs, consideration of topical reports, etc. The follow-ing assignments were made:
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r Safety Research Subcommittee AEC Safety Research Program Dr. Okrent, Chairman Dr. Okrent, Chairman Dr. Hanauer Dr. Isbin Dr. Hendrie Dr. Monson Dr. Isbin Dr. Stratton Mr. Mangelsdcrf Dr. Monson C. E.
Dr. O' Kelly Dr. O' Kelly, Chairman B&W Mr. Etherington Dr. Hendrie Dr. Hendrie, Chairman Dr. Isbin Mr. Etherington Dr. Hanauer Westinghouse Dr. O' Kelly Mr. Mangelsdorf, Chairman '
G. E. Dr. Isbin Dr. O' Kelly Dr. Isbin, Chairman Dr. Stratton 2
Mr. Mangelsdorf Mr. Palladino Dr. Stratton j 6. Jritium Release from_ Power Reactors f Dr. Zabel agret* to approach Dr. Morris about th. Dest way of documenting the information requ tid in Dr. Isbin's memo to Mr. Fraley of March 22, 1968.
' f j 7. Culf General Atom is 9
Dr. Zabel agre.ed to contact Dr. Morris and Mr. Price concerning a preliminary review of CGA's 1000 MWe reactor concept. The Committee will include this on I I
its August agenda.
- 8. Con. Ed. 'Iuclest Units 4 & 5 This project is to be scheduled for consideration at the August meeting. An
' extended period for internal discussion is to be included.
. 9. GE Training Simulator Dr. Zabel reported that the Committee had been invited to visit the new GE l simulator near Dresden, Illinois. Interest in such a visit was expressed by Dr. Hansuer, Mr. Etherington, Dr. O' Kelly, Dr. Isbin, Dr. Hendrie, and Dr.
Stratton.
I This visit will be scheduled at an appropriate time. i l
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- 10. AS&LB Decision on Fort Calhoun Dr. Zabel reported that the Hearing Board's decision on Fort Calhoun was somewhat unusual in that it recommended certain restrictions on the license which almost conflicted with ACRS technical judgments. The restrictions suggested by the Board were not included in the construction permit, however.
- 11. Hanford K-Reactor The Committee heard a report of a meltdown incident a: the KW-Reactor,
- apparently caused by an obstruction of unknown origir in a venturi used to measure coolant flow. The failed tube had been freed from the graphite matrix at the time of the report, although it had not yet been removed from the reactor.
- 12. Criteria for Instrumerf.ation Design The Committee voted to request the Regulatory Staf f to provide an elaboration as soon as possible of their most recent requirements for instrumentation based on the ACRS :equest for a review, made at the 98th reeting. It was also
- agreed that the ACRS report on the Zion Station include a request for a Commit-tee review of the it.strueentation before its fabrication and installation.
The Regulatory Staff was informed of this request following the meeting.
} 13. Future ACRS Meeting Dates I
The following schedule has been established, subject to confirmation at the 100th ACRS meeting, August 8-10, 1968.
l 105th ACRS meeting . January 9-11, 1969
' 106th ACRS meeting - February 6-8, 1969 107th ACRS meeting - March 6-8,1969 l 108th ACRS meeting - April 10-12,1969 ;
109th ACRS reeting - May B-10,1969 1 110th ACRS meeting - June 5-7,1969 MISCRIMNEOUS
- 1. Meeting with JCAE Representatives Dr. Zabel reported that a very worthwhile meeting with Representative Hormer
- and Mr. J. Conway, Executive Director, had taken place on Friday, July 12, 1968. Activities of the ACRS as well as the AEC Internal Review Group were discussed. Mr. Conway and Representative Hosmer noted that a valuable line l of communication has been established by this meeting which should be main-tained.
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- 2. Meeting with Commissioners The f'ollowing items were discussed:
- a. Technical Support of the Regulatory Staff The need for short-term, concentrated support of the Regulatory Staff when specific potential problem areas are identified was discussed. This support should be
'readily available and broadly based since a wide variety of topics might be involved.
The Commissioners agreed to consider such a scheme.
- b. Licensability of New Reactor Concepts Commissioner Johnson inquired if the ACRS considered itself availeble to assist in a determination of safety -
regarding new reactor concepts before a development of the concept was comp:eted.
Several members indicated that the Committee did consider itself available to assist the Commission in this area.
Itwasgotedthatsomedevelopmentwork(approximately
$5 x 10 worth) ia required to identify the safety questions,
- c. Use of Comparative Analysis in Evaluation of Reactor Safety Commissioner Johnson inquired if the Committee considered comparative analysis helpful in making safety determinations.
Several ACRS members indicated that it is a helpfel tool but noted that considerable staff support is required to eake the comparisons required.
- d. Meetina with JCAE Representatives k Dr. Zabel reported briefly on the meeting with Rep. Hosmer and Mr. Conway. He noted that all agreed that this line of cornunication should be maintained. l
- e. AEC Internal Review Group ,
1 The status of the group's activities was discussed briefly.
There was some discussion of the respective roles of the Regulatory Staff, AS&LB and the ACRS in the regulatory process.
- 3. Future Agenda The following projects are tentatively scheduled for consideration at the 100th ACRS meeting, August B-10,1968:
Consolidated Edison Nuclear Units 4 & 5 - Site review Culf-Ceneral Atomic 100016ie Gas cooled Reactor - Conceptual design
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NINETY-NINTH ACRS MEETING JULY 11-13, 1968 WASHINGTON, D. C.
Attendees
- C. W. Zabel, Chairman S. H. Bush H. Etherington W. L. Faith F. A. Gifford S. H. Hanauer J. M. Hendrie H. S. Isbin ;
H. G. Mangelsdorf '
H. O. Monson A. A. O' Kelly )
D. Okrent I f
N. J. Palladino C. P. Siess W. R. Stratton R. F. Fraley, Executive Secretary M. C. Gaske, Asst. to the Exec. Secretary J. E. Hard, Sr. Staff Assistant M. W. Libarkin, Sr. Staff Assistant i
L. Blische, Administrative Officer i Letters Written:
Rancho Seco Nuclear Generating Station Maine Yankee Atomic Power Station Hanford N-Reactor N-
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TABLE OF CONTENTS l
l I. Executive Session A. Appointment of Dr. L. Squires i B. Review of AEC 'egulatory Programs 9 C. Article by Dr. J. E. McKee "Science and Engineering" D. AS&LB Initial Decision - Fort Calhoun E. Report on KW-Reactor (See Classified Supplement)
F. Report of Dr. P. Paris G. ACRS Requirements on Instrumentation Design H. Safety Research Working Groups
- I I. Documentation of Information on Tritium Release j l
J. Report on Zion Station; Additional Remarks l
K. Report of Meeting with JCAE Representatives h
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L. Meeting with the AEC l l II. Meeting with the Director of Regulation and Members of His Staff A. Meeting with the Director B. Compliance Report C. DRL Report !
D. DRS Report III. Hanford N-Reactor IV. Rancho Seco i V. , Maine Yankee
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MWL:ci DRAFI 7/25/68 s 99TH ACRS MEETING MINUTES
- 1. Executive Session A. Appointment of Dr. L. Squires Dr. Zabel asked if Dr. Okrent and Mr. Mangelsdorf had pursued the matter of Dr. Squires' present status. Mr. Mangelsdorf reported that he had called Dr. Squires to see whether he had any objections to contacts with representatives of the du Pont Corporation. Dr.
Squires replied that he had no objections but thought it would do no good. In addition, Dr. Squires did not see how he could ave Q {44 edthe k necessary time for work on the Comittee at present. Dr. Zabel sug-might be ,
f gested that discussions with high level du Pont personnel / fruitful. !
I j Dr. O' Kelly added that one more try was worthwhile. If Dr. Squires i
is to retire in six months or so, then time will not be a problem.
i l Dr. Zabel noted that in his discussions with Chairman Seaborg , l l
l t the matter of Dr. Squires' employment with the Cocinission following his retirement had come up. Chairman Seaborg that there were many places where Dr. Squires would be valuable. The Comissioners p a s.g pc had, howeve.r, given their approval to the Conunittee ati.x pidly to seek Dr. Squires as a member. Dr. Zabel felt that a concerted effort j l
, should be made to get Dr. Squires' membership on the Cocxnittee.
- There seemed agreement with this course of action by the members and Mr. Mangelsdorf agreed that he would pursue the matter with top level du Pont personnel.
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Part 1, Page 2 B. Review of AEC Regulatory Programs Dr. Zabel reported on the status of the in-house review of AEC Regulatory Programs. A meeting was held on June 2 at which the charge to the review panel was drafted and appointment of members to the
. panel considered. A public announcement dated June 28 lists the mem-bers of the panel. It was agreed that the panel would try to meet approximately once every two weeks and would attempt to complete its work in two to three conths.
Mr. Mangelsdorf added that one of the first things which was agreed was to seek ideas from the reactor vendors. The Comission thought that a frank discussion would be possible and agreed that they would encourage vendors to express their real feelings. Dr. Morris I
noted that DRL has been having frank discussions with the vendors.
- Dr. Zabel also reported on a conversation with Mr. Price concerning i
- the review panel. There is considerable effort in the direction of strengt.hening the hearing boards' position, although this was not part of the panel's specific charge. The Comittee might want to provide input for the panel's consideration. Dr. Okrent pointed out that the public announcement lists the Comittee members as ACRS members and t
suggested that they keep the Comittee informed of progress. Mr. Man- ;
8elsdorf agreed and added that in his opinion the other groups repre-sented on the panel would be having "strategy" discussions. The Com-mittee is entitled to the same background. As f ar as formal minutes are concerned Mr. Mangelsdorf was not sure of their status.
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DRAFI - 99th ACRS Meeting Minutes r Part I, Page 3 C. Article by Dr. J. E. McKee "Science and Engineering" i I
Dr. Zabel referred to the relatively innocuous statements which l Dr. McKee had made concerning the Bolsa Island project in a recent l l
issue of "Science and Engineering" published by Cal. Tech. Dr. Zabel
. felt that Dr. McKee had been ' misrepresented by the press and considered l
. l that it was worthwhile noting how sensitive statements of even ex-ACRS ,
members can be. He asked if the Connittee wanted to pursue this mat-l ter. There seemed to be agreement on the part of many members that b e, I the matter should not pursued further. Dr. Okrent pointed out that i 6
Dr. McKee is no longer a member and was in conflict on Bolsa Island i I l in any case and so would not have taken part in the Cocnittee's deli- ;
I l i berations on that project.
Dr. Cif ford thought that he would prefer to have Dr. McKee speak I
on the issues involved than some other,perhaps biased, individuals.
Although there seemed to be some feeling that this type of pub-licity should be discussed with the Connission in connection with urging a more vigorous campaign to gain more widespread public accep-T s
tance of nuclear powerj There seemed to be a general concensus that 1
no further Cocnittee action should be taken on this particular matter. l D. Atomic Safety and Licensing Board Initial Decision - Fort Calhoun i 1
j Dr. Zabel referred the Coccittee to the AS&ls initial decision on Fort Calhoun. He personally felt the approach taken to be a serious 4
one) Since the Soard was recocuending in its decision certain n'
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DRAFI - 99th ACRS Meeting Minutes Part I, Page 4 restrictions on the construction permit. If the ACRS is to be respon-i sible for advice to the Comission on overall safety matters, then this decision represented a poor tactic. Specifically,the decision indicated that TID 14844 conditions are not necessarily applicable and refer to other general technical matters in which the %oard was r
in conflict (or al.nost) with the Comittee's decisiors. The Comission's guidestothebearingboardssaythattheyarefreetoreviewtechnical matters. Mr. Price now feels that this was a mistake and will try to have it removed from the 3oard's charge. In his opinion the real objective of the Nearing bards was to determine that the documentation, i
etc., was complete and complied with the law.
i Dr.Hanauerdidnotseeanythingimproperinthe3oard's' action.
In his opinion,if the board members believed the statements then they were,in fact,chliged to make them. He suggested that Mr. Price's* ap-1 j proach was wrong. IftechnicalmembersareappointedtotheOoards t4 and are asked to review projects they will do so.
Dr. Okrent felt somewhat as Dr. Hanauer. He thought that the Comittee should not become involved and that the matter would probably l i
come up S ::AI[t [a with the Taternal teview Fanel's discussion'. or.
Okrent did recall several times when the 3oard was critical of the ACRS c.
and;as an example)Aited the Savannah and the time-to-melt criterion.
In this case the board was not specifically critical of the Comittee, ct M t.he . 4.ipc'ints made ar; :::hnicall-j ::ak th:n h should stand or fall on their own merits. Perhaps there is some advantage to such an approach
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Part I, Page 5 bytheharingboards. The ACRS as well as the Regulatory Staff will tb be aware that they may be "reviewed". Perhaps from the point of view of the public this is the best overall situation.
In answer to a question by Dr. Siess, Dr. Zabel observed that the conditions listed in the provisional construction permit were
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core or less standard ar.d not those suggested by the Nearing Ocard.
Dr. Hanauer thought it was always the case that even when the Regulatory Staff has serious reservations these are not reflected in the permit. Perhaps the Cocraittee should ask Dr. Morris where he
- thinks he now stands with respect to Fort Calhoun.
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- E. Report on KW Reactor -- Classified Supplement i
F. Report of Dr. Paul Paris i l Dr. Paris observed that while gathering data and thinking about i
l the thermal shock problem his attention was directed to a problem I
with operating reactors. Dr. Paris referred the Conraittee to a hand-out which he had prepared which outlined the basis for his concerns. (1<c
.thnet) He referred to the extrapolations of fracture toughness values and 1
i pointed out that for irradiated material one cannot prove that a more conservative extrapolation,and <eneepently. values of fracture tough-ness ranging from 60 to 50,is not actually the case. Dr. Paris briefly
, e, discussed the calculations outlined in 4he handout.rafe--ad *a ^c ec.
He pointed out among other things that for stress intensities +f. twice the fracture toughness, then crack division and asqa.tly shattering l
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Par; I, Page 6 will result. He noted how critical flaw size as calculated in his paper will decrease with irradiation and residual stress. If one considers the wall thickness necessary to achieve flat fractures and M strain conditions then very thick walled vessels are involved;
, as much as twelve inches perhaps. In Dr. Paris' opinion it is not i 1 possible to-extrapolate experience with non-irradiated and thin (i.e.,
- L i less than eight inches) vessels,to predict that experience in the future with thick walled, irradiated vessels will be satisfactory.
Dr. Paris observed that in his experience failures have been associated with flaws which were less than about 20*/. of the vessel thickness. He was aware of no test programs on suberitical flaw i C- s . w growth under p4eh strain conditions. In addition, there are no tests 4
. at all which have been proposed on irradiated materials.
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i Dr. paris' conclusions from considerations dhe all of the above are as stated in his outline: critical flaw sizes at the end of vessel life may be extremely small*, it is not possible to predict the characteristics of thick walled, irradiated vessels from the behavior of relatively thin unirradiated f material samples' local stresses and residual stresses are not necessarily negligible *, there are no tests presently being performed or planned on subcritical flaw growth, 1
l Dr. Paris ended by noting that a reactor failure near end of life might be credible under normal operating conditions if the ac-4 tual facts are no better than his evaluation had let him to believe.
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Part I, Page 7 Dr. Zabel asked if Dr. Paris' point of view might not be the subject of some disagreement by other experts in the field. Dr. Paris felt that most of the previous evaluations were from the point of view of nil-ductility temperature transition. Such material tests as have
, been performed in the fracture mechanics field were done on small, thin ;
Charpy specimens. Dr. Okrent asked how much of a drop in flux could be expected through nine or so inches of steel, and what dif ference this might make. It was noted that the flux might drop by a factor ;
of five through a nine-inch thick steel specimen. Dr. Paris thought
, this would not make any significant difference based j
on the numbers A ti , M N w II which he had discussed with the Coccittee. N flaw will grow and '
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] its influence will grow faster than the fracture toughness will in-u crease. Of first importance is the obtaining of good fracture tough-ness values for irradiated material.
P Dr. Bush commented that he could take considerable comfort in the flux gradient through the wall. In addition,he felt that it would be prudent to wait! for the results of the AEC's two million
, dollar program which is aimed at obtaining the necessary data during the next five years. Dr. Bush was not convinced a b the validity of the numbers which Dr. Paris had discussed with the Comittee and
- noted that the program he had mentioned was supposed to provide irra-diation data and to lead to A development of a model which would pre,-
dict ductile failure. He could agreejhowever, that the critical flaw l size would be very small at high fluenN'w r.
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DRAFI - 99th ACRS Meeting Minutes Part I, Page 8 Dr. Zabel asked if this approach might lead to concern over heatup and pressurization transients. Dr. Bush replied that the worst situation which he could visualize would be a startup trans- l ient resulting in a pressure pulse at low temperatures. Such a situation would result in shattering, etc. Mr. Palladino commented i i
that if there are uncertainties of 2 to 3 in a twenty year life estimate then perhaps only seven years or so are available.
en es t Dr. Bush observe,d that,at present,the Committee must be, con-cerned about Indian Point 1 and Yankee from this point of view.
Mr. Palladino asked how much better the data must be in order e%c to e4+ee+ Dr. Paris' concerns. The answer to this question was not clear. Dr. Hanauer asked if Dr. Bush had reviewed the HSST pro- )
gram and felt that suf ficient data vill result to allow calculations f
is to be done with some degree of confidence. Dr. Bush replied that the
$ difficulty in collecting the data is that )if materials of the wrong I
geometry, sizes, etc., are irradiated,the data will probably look too good. Work now is being aimed at establishing the appropriate 9
sizes, etc. Dr. Bush felt that the program would ultimately furnish
[ a fair number of answers.
Dr. Paris asked if Dr. Bush agreed that the HSST program would
! not provide much in the area of suberitical flaw growth. Dr. Bush I
did agree and suggested that several of the individuals involved in mapping out the program are also of the same opinion so that this too i s.H W L.
will probably be cyp-c2:hed, l
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Part I, Page 9 Later in the meeting Dr. Stratton suggested that Dr. Paris be asked to prepare a careful and detailed writeup of his paper. Dr.
Isbin asked if Dr. Bush could comment on why the fracture toughness values in Dr. Paris' paper had been so low compared to those used 1
by some vendors. Dr. Bush replied that Dr. Paris used the available ;
data which was mostly taken at temperatures much less than room 1
a temperature)and approaches these highly conservatively. The diffi-culty with fracture chanics is that the majority of the fracture !
toughness values available are not valid. NdO$i strain conditions should have been present while the data was taken*, and the samples a
r . .$ . & t . . . o o used h.sve been too small. Any non-hlI55 strain datajib Dr. Bush's*
n opinion,will be tuo hight that is non-conservative.
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{ Dr. Okrent suggested that the Committee try to arrange a meeting i
e with as many knowledgeable people in this area as possible. He felt t
u that it would be as Dr. Bush had suggested, that is, that the results depend on how the extrapolation of the data is approached. Dr. Okrent f(-b l also pointed out that Dr. Paris had saidg in his opinion it would be 3 i
difficult to detect half-inch flaws. He noted that NASA had performed f some tests on teme thin plates with 20% thickness flaws. These were sent to their five best contractors and the success in detecting these i flaws was about 50%.
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With respect to fracture toughness data,Dr. Bush commented that if the AEC were to cut back the HSST program then it would be desirable to take some action. The current program, however, is simply going to take a great deal more time.
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Part I, Page 10 There seemed to be agreement that Dr. Paris should be asked to flyihouthisreport to the Comittee.
Mr. Mangelsdorf asked how much of an ef fect the heat transfer model would have on the results. Dr. Bush felt that the ef fect would be trivial. He suggested that the thermal conductivity of the steel I
' would in fact control.
C. ACRS Requirements on Instrumentation Design Mr. Mangelsdorf suggested that the Regulatory Staff be asked to proceed with an appraisal and develop their position on the separation of control and safety instrumentation. He thought that,in effect,the
{ Comittee had just such a proposal from DRL in their most recent docu-i e ment on the matter. Mr. Mangelsdorf was willing to have the paragraph I
on instrumentation which appeared in the Salem letter also rppear in the letter on the Zion Stationy f the Coenittee agreed that DRL would
, be asked to make a reasoned proposa!.
Mr. Fraley pointed out that Mr. Schroeder's opinion of the para-graph in question was that it meant that the Comittee had accepted
- the DRL approach which was used on the Prairie Island review. In addi-tion, diversity would be provided. Mr. Mangelsdorf comented that his
! acceptance of the paragraph was based on the fact that it was opene+
to several interpretations. If,in hi: : pi 9 ; it meant that the Com-mittee had fixed on the position into which the staf f had been forced on the Prairie Island review, then he'could not accept it.
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Dr. Stratton thought that some evolution was taking place along the lines which Dr. Hanauer wanted but did not feel that an abrupt change in instrument design was appropriate. Dr. Hanauer recalled l
that the Comittee's concerns over instrumentation had started with uhM
, b Indian Point 2 which was,more than two years ago. He agreed that designs could not be changed in a day, but could not understand n :- a : a ,.. .. ,. <
why the Cocnittee was unable to-dccidc cn _ d'. m *4aa *^*h Dr. Hendrie added that a vendor could not be "evolved" if he would not evolve himself. If Westinghouse will not move from its position then the design changes must be abrupt. Dr. Hendrie also felt that Mr. Mangelsdorf's suggestion would not give the Regulatory Staff any guidance. Whether or not the Comittee's position on Diablo Canyon, for example, was right or wrong the Comittee should try to decide
, its own opinion at this time.
- Mr. Mangelsdorf did not feel that DRL was in need of specific Comittee guidance. He thought that the staff now realized that the problem was extremely comple that they are now ready to make an absolutely clear statement of their position,and Mr. Mangelsdorf thought that he could go along with the Comittee's opinion following such a statementj whatever it might be.
Dr. Hendrie could agree if he thought that the Regulatory Staf f
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was in that position. He thought that before it would be possible to have the Regulatory Staff produce a clear-cut statement of its own position on separation and on diversity, The Comittee cost first i ,-- . ..-
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Part I, Page 12
' tell them clearly t. hat all past Committee cocrnents in this area are to be ignored. Mr. Fraley observed that that was what the Regulatory Staff had been told following the last Comittee meeting. Dr. Zabel agreed.
There was additional discussion over what position the Comittee should take. Mr. Mangelsdorf stated, as his own position, that be could accept the staff's approach to the Prairie Island instrument design if they reached that conclusion on their own. If, however, it was reached because ci tne Comittee's statements on Diablo Canyon, then he could not accept it. He had no objection to the proposed paragraph for the Zion letter if too many limitations are not put on it. He suggested again having the Regulatory Staff proceed as they I
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- heve- proposed and bring a clear cut position to the Co::raittee as soon l s
as possible.
i Dr. Hanauer recalled that the staff had first accepted Westinghouse's
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proposals on Diablo Canyon. The Committee took exception to this in its letter and the staff then modified its approach in connection with the Prairie Island review. In the second case,the Committee took no
[ position, but simply asked what the staff would do if they had had a 1 1
free hand. The Regulatory Staff's reply was to retreat to the Diablo Canyon position and to accept the IEEE standard on instrumentation. At this point the Comittee wrote a memorandum t o DRL asking that systematic failures te be considered. Dr. Hanauer felt that@ it was*being pro-posed that DRL be asked to do the Cocxsittee's thinking for it. He
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suggested that what would actually happen would be that the Staff would again try to "read the Comittee's mind."
Mr. Mangelsdorf thought that was not the case. He proposed to have the Staff do Staff work. He suggested that the Comittee had been making arbitrary decisions concerning instrument system designs without sufficient information.
Dr. Okrent observed that since the Diablo Canyon letter requires that that system's instrumentation be brought back to the ACRS for review that would provide an opportunity for any member who is not satisfied to dissent. He suggested that the Staff be told that the
, Comittee's present position, as outlined in the Salem letter, should f apply to the Diablo Canyon reactor. In addition, the R.'gulatory Staff could be asked to propose an approach for the Comittee's review.
Dr. Hanauer did not think there would be a chance to dissent on N
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i 6 Diablo Canyon instrument design. He did not feel the Comittee would i write a letter on the Diablo Canyon instrumentation and cited as an example the case of the Indian Point 2 core cooling systems. Mr.
. Mangelsdorf observed that, in the last analysis, Indian Point 2 had h
ended up with a better core cooling system than was thought possible at the time of the Comittee's review. Mr. Mangelsdorf moved that the Cocnittee ask DRL to provide an elaboration of their requirernents on instrument systems expeditiously. Dr. Stratton seconded the motion.
Af ter some discussion of the meaning of this motion, Dr. Okrent moved the following amendment
- that the motion include a statement I l in the Committee's letter on the Zion Station to the effect that the instrumentation should ed b t ,e omittee before its fabrication
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Part I, Page 14 and installation. Mr. Etherington seconded the motion to amend, and wa s b l it was passed,eight in favor and two opposed. The motion was thenr.
passed , thirteen in favor, none opposed.
H. Safety Research Working Groups Dr. Zabel noted a suggestion made during a Procedures Subcommittee meeting that working groups be formed to discuss with individual vendors the general problems not connected with specific cases. 'Ihese groups would follow vendors' safety research programs and, perhaps, give t.
meaningful reviews of topical reports. The following working groups have been established:
Safety Research Subcommitteer D. Okrent, Chairman, H. S. Isbin,
- S. H. Hanauer , H. O. Monson, A. A. O' Kelly, J. M. Hendrie j H. G. Mangelsdorf; I
Babcock 6 Wilcox Working GrouprJ. M. Hendrie, Chairman,
. S. H. Hanauer, A. A. O' Kelly, H. Etherington;
_ Combustion Engineering Working Group,. A. A. O' Kelly, Chairman, H. Etherington, H. S. Isbin, H. M. Hendrie; General Electric Working Group _rH. S. Isbin, Chairman, H. G. ;
Mangelsdorf, N. J. Palladino, W. R. Stratton; Westinghouse Working Groupr-H. G. Mangelsdorf, Chairman, A. A. )
O' Kelly, W. R. Stratton, H. S. Isbin;
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AEC Working Group,.D. Okrent , H. S. Isbin, W. R. Stratton, H. O.
Monson.
Dr. Zabel observed that the Comittee's list of unresolved questions could be used in this connection, but that the Working Groups' consideration of vendors' research programs need not be limited to this list. _ _ _ , , _ __ .____,n
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Part I, Page 15 I. Documentation of Information on Tritium Release There was some discussion of the approach which should be taken to the questions raised in Dr. Is. bin's memo on tritium release.
Dr. Isbin suggested that the Regulatory Staff ask the applicants to
,- furnish the information requested in the documentation provided with each case.
Dr. Zabel agreed to approach Dr. Ms.cis and attempt to decide s-on some way of handling documentation E this information.
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7/17/68 Oidf,/ E5' E' 99TH ACRS MEETING MINUTES I. Executive Session J. Report on Zion Station; Additional Remarks During the Comittee's discussion) cancerning their report on the Zion Station, Dr. Okrent presented a draf t of additional he,c,.4kc.(.he.da remarks which he proposed 2 ppa"A4 3 te the C m itt::' r: pert te
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"G . While not dissenting from the Committee's opinion concerning issuance of a construction permit, Dr. Okrent suggested certain additional matters for consideration and resolution prior to the completion of construction. In addition, Dr. Okrent's remarks concerned future reactors of the Zion type which might be located t at sites having simils.r or somewhat higher population densities.
l Dr. Okrent's remarks took note of recent improvements in quality control, fabrication techniques, and provisions for in-service bu.t I
inspection and leak detection of primary systems,end indicated that other questions have arisen in connection with design and f abrication errecte- of pressure vessels which are as yet unresolved. He suggested that for the Zion reactors studies should be made,within the framework of the proposed pressure vessel design,of methods of protecting against limited size leaks and splits in the pressure vessel wall. For future
. reactors to be located at sites having similar or greater population ,
' l densities than that of the Zion site, Dr. Okrent suggested addftional protective measures wit.h regard to pressure vessel failure. Six specific suggestions were included.
I DRAFT - 99th ACRS Meeting Pgr t I.J.b/F2ge h oI, s' 94 There was considerable discussion over the effect which such an addendum would have on the Committee's letter and of whether the Comittee wished to address itself to Dr. Okrent's remarks in the body of the letter. Several members suggested that additional remarks by ACRS members should somehcu be restricted to the specific case at hand; others felt that members with concerns about aspects of l l
reactor designs affecting the health and safety of the public should i be given free rein to voice them. Several members suggested that .
the coments might be more appropriate in a general Comittee letter on nuclear reactor pressure vessels since they could support many of
- the coments suggested by Dr. Okrent in such a letter. The discussion ,
e i then turned to the feasibility of some combination of the four proposed t
pressure vessel letters prepared by Dr. Okrent, Dr. Bush and Mr.
Etherington. Dr. Bush indicated that he was willing to attempt to i unify three of thepressure vessel draf t letters and noted further that his own approach had been to assume that vessel failure will occur in those cases where reactors are proposed for siting close to population centers. The penalty paid to get into a metropolitan-type site would be design for protection against vessel failure. Mr. Etherington I
4 objected to that approach. His draft had taken the view that neither the PWR nor BWR vessels, as presently designed, ri '. k kL but are mot
-yas acceptable for close-in sites. The PWR vessels must be redesigned to limit fluence at the vessel wall to such values that the steel will not be embrittled. The BWR vessels must be redesigned to make the internals removable on a prictic al basis and make in-service inspection truly feasible.
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Dr. Bush added that, in his view, a Committee letter on pressure vessels is not appropriate for sites similar to Zion ce ed those having lower population densities. He pointed out the good progress which had been made in such areas as quality control,
! in-service inspection, and the formulation of primary system piping design criteria. These improvements have led him to take a more relaxed view with respect to systems that will be proposed after issuance of criteria in the above areas, sometime late in l
1969. In fact, Dr. Bush continued, he would probably dissent from ;
such a letter. In his view, a strong Comittee letter on pressure vessels at this time,which applied to plants sited at Zion or better 1 sites,would probably undermine Mr. Case's work on standards. l
, The discussion of Dr. Okrent's proposed additional remarks l e :
j continued. Several members suggested that Dr. Okrent be asked to rewrite his remarks so as to make them more periinent to Zion N '**
although Dr. Okrent indicated that, in his view, this was already the case.
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Several members suggested that the Comittee should somehow restrict the content of additional remarks attached to Comittee i
reports. Others thought that the Comittee should not attempt to
. regulate the content of such additional remarks although in some memberd views the rest of relevance was an important one.
Dr. Okrent repeated that, in his opinion, almost all of his added coments applied directly to Zion or Zion-like reactors. He recalled that the Comittee had voted,before its review of the Zion proposalj that they would no a Diablo Canyon type reactor at
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DRAFT - 99th Meeting F-: I>J. , rag ^- ,, _. ' f fy,.l*[, /k SG /V the Zion site. In his opinion, there had not been very many changes made from the Diablo Canyon design. Dr. Okrect felt that by approaching things as he had in his additional remarks it would allow him to approve reactors at even worse sites than Zion. If the Committee did not permit his coments to be attached to the letter it would be,in a sense, forcing him to take a strictly negative position. '
i The discussion continued with reference to the relevance er 1.sel-ewmee o f s ev er al s p e c i f ic i t em s in Dr . Okr en t ' sde smr.44
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- c. e . g e . e u e Dr. Hanauer fcit that s4/frring to future Zion-like reactors 4*-
1 thir -- n : would severely limit his freedom of action with respect ,
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' to future proposals similar to Zion. He did not think the Comittee i
should indicate any position about future proposals and felt that this was central to his objection. Other members agreed that such r
an approach would weaken the Comittee's position ji future.
Mr. Mangelsdorf suggested that the central difficulty with Dr. Okrent's suggestion was that the bulk of the ideas included were subjects already discussed by the Consnittee and that Dr. Okrent was proposing to make these coments at this time as an individual member.
Mr. Mangelsdorf thought that, if Dr. Okrent's remarks were attached to the Comittee's letter, this should be pointed out. Dr. Hanauer agreed that the additional coments by Dr. Okrent would representj in large part 3publication of incomplete Comittee deliberations which the Comittee as a whole was not yet ready to publish.
A motion was made to the effect that the Committee would issue its letter without including the additional remarks of Dr. Okrent.
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DRAFT - 99th Meeting ,, , ,
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Puy 1.J., Fa g 5
/ aai1, /h.Sc >U Several objections were raised on the basis that the time available for discussion was not sufficient ;'.. . J. . . . u u, ..; .rr;;;i i o 2:
= and that such censorship of individual members might lead to ineffectiveness of the Comittee as a whole.
There seemed to be agreement that wf thout suf ficient discussion on such a proposal the Committee could not proceed and an cdditional 1 meeting on this matter seemed appropriate. It was agrued that a special meeting would be held on Sunday, July 21, 1968, at which the Zion letter and the proposed additional comments,as well as a general ACRS letter on pressure vessel design and fabricationjvould be discussed.
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F DRAFT - 99th Meeting - -
Part I., Page21 ,*)
.o z . -
K. Report of Meetinst with JCAE Representatives Dr. Zabel reported on the meeting with Representative Hosmer and Mr. J. Conway, Executive Director of the JCAE Staf f, which was held on Friday, July 12, 1968. ACRS and the Internal Review Panel activities were discussed. In Dr. Zabel's opinion, both Mr. Conway and Representative Hosmer thought that this meeting represented a valuable 'N of comunication which should be maintained.
l Dr. Zabel also observed that the Joint Committee representatives ,
i were of the opinion that the Comittee should resolve technical issues
- in camera' rather than having them aired in public hearings.
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99th Meeting - DRAFT RFF:mg 7/23/68 -
. ,I Part 1, Page 22 L. Meeting with Comissioners This meeting was limited to the following participants:
ACRS Regulatory Staff
- Members H. L. Price R. F. Fraley, Exec. Secy. C. K. Beck R. L. Doan Commissioners M. M. Mann G. T. Seaborg, Chairman C. L. Henderson J. T. Ramey *^*"##
- G. F. Tape * * #8' W. E. Johnson Office of the Secretary AS&LB F. T. Hobbs, Asst. Secy. A. Wells (during portion on AEC Internal Review Group Only)
Technical Support for the Regulatory Staff
, Dr. Zabel described the need for diverse, expert support of the Regulatory Staff to assist in the expeditious evaluation of h
problems which are identified during the review process. He noted that prompt, short-term, concentrated effort is frequently needed in specific areas related to safety, but the present system does a
not seem to provide this support.
The Corsnissioners appeared to agree that an arrangement could be made to provide this type support for the Regulatory Staff.
Comissioner Johnson suggested that the national laboratories might be able to provide this type of support if the laboratory directors were made directly resp'nsible to the Director of Regulation for a certain portion of the laboratories' activity.
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t 9th Meeting A+I
- 9 Meeting with Comissioner s, Page 23 ACRS Participation in Safety Evaluation of New Reactor Concepts Comissioner Johnson inquired if the ACRS considered itself available to assist the Comission in the safety evaluation of new
. reactor concepts before the concept has reached the licensing stage.
Several members of the Comittee indicated that the Committee does consider itself available to assist in such activities. Dr. Okrent i recalled the suggestion that a PSAR for a typical fast breeder reactor be prepared by DRD6T and processed through the Reguletory review. It was noted that some development work (approximately five million dollars) does have to be completed to identify safety related
. questions before a meaningful review could be made.
t
, Use of Con.parative Analysis
)
, Corrnis s ionet Johnson inquired if the ACRS considered the use a
of comparative analysis useful.
Several members indicated that it is useful and is being used in many instances in safety evaluations. It was noted, however, that a considerable amount of staff work is needed to make the comparisons required.
Meeting with JCAE Representatives l
Dr. Zabel reported that a very worthwhile meeting of Mr. !
Mangelsdorf, Dr. Zabel, Representative Craig Hosmer and John Conway, l
Executive Director, had taken place. He noted that all had agreed l l
that this method of communication was useful and should be maintained. l
. .o .
99th Meeting /.ov - ,, , . _f )
Meeting with Commissioners,g Page JEpl Activities of the AEC In-House Review Group Mr. Mangelsdorf noted that the first meeting of the group had been held and that additional meetings were scheduled at approximately two week intervals.
There was some discussion of the roles that the Regulatory Staf f, ACRS and the AS and LB fill in the regulatory process.
Commissioner Ramey suggested that the AS6LB review should be comparable to the review by a higher court of a decision made by a lower court or review at the General Manager level of staff work that has been referred to him for a final decision. He questioned the usefulness of public hearings in uncontested cases, however.
f Dr. Okrent inquired regarding the role of the ASLLB when the ACRS no longer reviews reactors of "standard" designs. Commissioner Ramey
, suggested that this would probably be an evolutionary process but '
I anticipated that the Board would have to prebe deeper into the work ,
l of the Regulatory Staff as the ACRS withdraws from this activity.
l He noted that the AS6LB could:
1
- 1) Review the proposal and reach a conclusion regarding the issuance ok a license with any added requirements f
considered necessary or
- 2) Evaluate the review of the Staf f and return it to the Staff if an adequate job has not been done.
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99th Meeting 8 - - - - - 1 Meeting with Comissioners q Page ,Rf '. .
.3 He noted that he favored the second alternate. Mr. Price indicated, however, it would be an impossible task to adequately document the l l
extent, scope, etc. of the Staff review on the public record. j l
Dr. Zabel noted that the ACRS evaluates the work of DRL in its l I
review of a case and suggested that another review of this nature I by the AS6LB was a duplication of effort.
Commis-toner Ramey indicated that he considered the following j i
assignments appropriate; j
- 1) Regulatory Staff does the basic, in-depth review, ;
1
- 2) ACRS makes in-depth review of key safety itees, l 1
- 3) AS6LB does an overall review and probes occasionally ]
t f in specific areas.
I Dr. Okrent noted that an occasional critical review of the Regulatory Staf f and the ACRS work would help the regulatory program if done judiciously.
Reply Regarding Subpoena of ACRS Meebers Chairman Seaborg noted that the Cornission has no basic ,
1 disagreement with the concept that ACRS members should not be l l
subject to subpoena for AS&LB Hearings. A reply to this effect l
- 1 is being prepared.
S
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-1 MWL:mg d n . .. I DRAFT - 99th Meeting II. Meeting with the Director of Regulation and Members of His Staff A. Meeting with the Director
- 1. Bolsa Island Review
,- Mr. Price raised the question of how best to approach the Bolsa i
Island project at this time. He recalled for the Comittee the submittal by the applicants containing information on the site as well as the report of the sp :iaF panel on geology and seismology, etc. Mr. Prise observed that it would be possible to do nothing until the applicants have formally submitted a complete application. In his opinion, however, I
the Staf f and the ACRS could profitably review the site at this time i before more of the application is on file. The Staff has sent a preliminary ,
I report to the Comittee sunnarizing the available information.
I Dr. Okrent asked if the seismic criteria would be out for comment
^
rn h e.
t in the near future. He pointed out that these criteria connected with Bolsa Island, although the review could probably be carried out 4e U
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Mr. Price felt that it would not be wise to judge this, or any other f acility jon the basis of criteria which have not yet been issued. He stated frankly that the delay in issuing the C c$iktri emente was of wok partly due to the fact that he wanted to know whether they would rule out
{ the Bolsa Island site.e m t, M rc umir in a w Dr. Zabel asked if it was 2ppropriate for the Cottnittee to start action on the Bolsa Island project at this time. Mr. Price replied that it was probably not appropriate now. He noted.however, that the project had been with the Staff and the Comittee for some time. He would not 4
like to be in the position of deciding that the site was not acceptable t
af ter all other fact ily resolved. Mr. Price [
[ a . -- ---.-
. p ._ - . y DRAFT - 99th Meeting Part II, Page 2 did state that the information in the Staf f's hands would not be complete until input was received from the Coast and Geodetic Survey Ste Nt, t.
and the U. S. Geological eervey. He had wanted to know eely if the o.p u r kle Comittee was r_ceptim to an early review. It was not clear whether i* 4.r.
the others involved, th:: is, the applicants, are ready to talk at this time.
Mr. Mangelsdorf closed the discussion by noting that the best position would be for the Comittee arid Regulatory Staff to be available and for the at.plicants to be unavailabic.
i
- 2. Subpoena letter i
Mr. Price comented that the Comittee's suggested draf t of June 27 i
l had not yet been sent to the Comission. The Comiestoners do know, '
i however, that the ACRS wants some sort of agreement concerning subpoena of members by hearing boards,and they have a copy of the Comittee's draft. Mr. Price has not yet discussed it with them or worked up an alternative suggestion, etc. Neither the Comittee nor the Regulatory l 1
i Staff are yet ready to discuss the subpoena letter. l
- 3. Regulatory Staff Personnel Requirement,s,
, Mr. Price reported to th,e Comittee that he had been '. cing 21.- l
- f. s e. w s s i n
>--->t .s. p,,,...,, nr es. ,,g g e C the Regulatory Staff's I
fs.R We b.u *uk N blyk need for personnel in connection wit.h the legislative requirement to cut back to personnel levels of 1965. The Bureau of the Budget has indicated that it recognises the Regulatory Staf f's personnel needs and will not
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DRAFT - 99th Meeting Part II, Page 3 force such a cutback. However, any additional personnel required to
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meet staff needs may have to come from the General Manager's :!d: :I
-t h e -C : m i e * ^ n -
4 Mr. Price reported to the Comittee that the Regulatory Staf f had been approached by the Gulf-General Atomics Company concerning a 2
discussion of their product line reactor with the ACRS. The Regulatory Staff had told the that they are not in a position te undertake any J.% ) c,. Q t m reviews in -th e c.bm.cc o f a design.
', Dr. Zabel added that he had received a phone call from Mr. De Hoffman indicating that GGA would like to come in with a 1000 MWe design study I
for the Comittee's consideration. Mr. de Hoffman had added that it
}
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would cost tens of millions of dollare for a detailed design,and the company does not want to spend the money if the door is going to be
{ closed in the last analysis. Dr. Zabel had replied that he would CV discuss the matter with the Comittee and recomended "^- ' $ wr comunication with the Regulatory Staf f.
Later in the meeting it was decided that Dr. Zabel would talk with Dr. Morris and Mr. Price and determine exactly what the Wlf- On:: 1 S t y('a 2.t~-iz: Company wanted and what the dif ficulties were. The Comittee agreed that it would be possible to hold discussions concerning a conceptual design even though there was not suf ficient detail to warrant a DRL review.
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DRAFT - 99th Meeting O -. M Part II, Page 4 B. Compliance Report Oyster Creek Pressure Vessel Repairs Mr. Reinmuth reported on the status of the stub tube repairs in the Oyster Creek pressure vessel. With the exception of the two Li-la tubes which hc.ve been held,at the request of the AEC, cladding has to been completed on all tubes. The bottom Inconel-1-shop weld tie-ins have been completed. Several of the upper field weld tie-ins have also been completed, adhere are about 30 days of work lef t tdat cia.1 at the rate of approximately 3 per day.
The control rod drive housings have not been replaced as yet.
I The 69 in-core monitor tubes must all be repaired. The hydrotest i is presently scheduled for August 18, 1968 which seems optimistic.
Mr. Reinmuth felt that the quality of the welds was considerably j m o
, better than those which had been y p^~d originally. Ultrasonic at uks inspection and dye checks are revealing no significant du < rmiss, although two of the first four tubes to be repaired needed replacenient because of errors in setting the welding machine. After reset of the machine)all velds perfomed have been satisfactory. Mr. Reinmuth pointed out, in addition, that the standard which is being used to 3
check the dye test results is tighter than that being used in the shopf
". t' fr and[than the code requirements. l h \
The operators are still awaiting receipt of a special micrometer i l
which will allow the outer and inner diameter measurements which have !
been requested on the two tubes being held for the AEC. Following these !
measurements, reference dye checks will be perfomed and repairs will be started.
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DRAFT - 99th Meeting '
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Part II, Page 5 Dr. Bush asked if there was any special effort being made to control fitup on the thimbles. He also asked if profile checks would be performed following the welds to reveal any dimensional changes.
Mr. Reirenuth noted no special ef fort at alighnent and added that as no profile measurements were being performed. ndrel wrs being used to prevent shrinkage.
Mr. Ltherington commented that he would have liked to see some justification for a brute force' approach to the prevention of weld
, shrinkage.
, Mr. Reinmuth reported on the status of the fix for the cracking of the weld between the upper and lower shroud *. The tur(buckle arrangement previously suggested has been eliminated. Insteadj an all-welded fitting will be pieced between the upper shroud and the vessel i
[ and an adjustment will be made by boring a hole in a temporarily located i pin, af ter the fitting has been welded in place. A permanent pin will be inserted and reweldedgia -l'.1 eliminat[Sany stresses on the fitting.
In conclusion, Dr. Bush suggested telling the Oyster Creek group that some difficulty would be avoided if profile measurements been made before they 4.*p c mzy i jwe W ort N
- 6,the Cournittee,4e nir.
I Nine Mile Point Mr. Reinmuth reported that the pressure vessel is now undergoing cleaning and a hydrotest will be performed in September. A complete dye test will be done following this. If no additional cracks show up, there will be no more repairs. p-k M . The same ultrasonic inspections have been performed on the stub tubes for the Niagara Mohawk vessel as have been f 129 welds inspected, 73
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a . . - .I DRAFT - 99th Meeting Part II, Page 6 had at least 3/4 inch fusion lengths (the total icngth of the fusion line was 1.1 inches). Fifty-six of the welds tested had fusion lengths less than 3/4 inch, and 4 were less than 3/10ths of an inch and would not meet the strength requirements. n e criterion, based on code samples, is 3/4 of an inch fusion length, ne applicant now proposes only to add weld metal so as to achieve the 3/4 inch fusion length required in those velds which do not quite meet this standard, his with will leave some lack of fusion 'in the welds, but it will meet the code 4
requirements.
6 Dr. Okrent asked if anyone had examined this procedure in connection g with the basis for the code requirements. Dr. Gluckmann noted that the
, ita a
I Regulatory Staff feels this approach is acceptable if one considers A
possible extension of the voids because of thermal cycling. The code was written on the basis that the voids would not connect due to such extensions. The shear area available is the important parameter.
Dr. Okrent asked if it was clear that the shear area would act in the same manner regardless of the fact that there were internal voids. He
?
pointed out that the stress distribution along the stub tubes is non-linear. Mr. DeYoung replied that the tests to be performed would show in (, th da. A i any leaks which ve"hthee occuru4 toisoon M y=after operation.5 If leaks occur during operation, they should be slow. n e procedure is within the code requirements and it isMr.Maccaryhpinionthattheprocedure WO is an appropriate one although the code comittee hak not been asked fomally. De Regulatory Staff will have an opportunity to decide on the aptness of this procedure when the final repair report is submitted.
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DRAFT - 99th Meeting Part II, Page 7 Dr. Bush observed that a very high percentage of similar velds suffer from the same partial-penetration type lefects unlike others, however, these particular welds have been insreeted with a technique which allows the detection of such defects.
It was noted in closing that the code requirement of a fusion length of 1.25 times the component thickne.s referred only to the geometry of the veld and not to actual fusion between the veld metal and the base met,al.
Lacrosse Boiling Water Reactor (Mr. Etherington did not take part in this
! discussion)
. Mr. O'Reilly reported on the emergency condenser solation valves j which failed to open following ascram from 257. power. With respect to the I
e steam line valves, an inspection has determined that the valves were a
[ standard type, used in about 10 power reactors. They were used in an
- appropriate environment and for the proper service. The cause of their failure to close is not now known. The manufacturer is investigating the incident.
, Elk River
?
Mr. O'Reilly noted that the reactor has been shut down for about six months because of a poesible pressure vessel leak. Pressure tests had l been completed at 250 psi and 3000F. Last week a pressure test was com-4 l pleted at 1000 psi and 3000F. An accumulation of water, apparently a func-tion of the internal pressure, was noted in the cavity. At 1000 pounds in-ternal pressure, water collects in the cavity at the rate of about 500 mil-liliter per hour. However, while both a volatile M % pg
.- [4 p -
5
. 4 t- -q DRAFT - 99th Meeting Part II, Page 8 and nonvolatile tracer we weesA used in the water, no tracer could be woc detected ir. ;h; nie- in the cavity, and it is not clear where this A
water originated.
. Dr. Bush asked how hot (radioactively)the area under the vessel
.d was. Mr.O'Reillynotedthat the area had not been inspected visually due to limited access.
Dr. Isbin asked if there was some indication as to how effective the internal safety comittee was in this situation.
Mr. O'Reilly r '6t i mwed that the comittee had reviewed the results of the test. The Compliance inspector for the plant indicated that there was apparently h some disagreement on the comittee concerning the meaning of the test t
l results. The licensee feels that the lack of tracer shows that the vessel is sound. The safety comittee is not convinced, but has not disapproved l ku.t wes (; \b d
- of this conclusion. % hperationi h e e"- U tappad, and the licensee will submit a complete report.
Ginna Plant i Dr. Zabel asked if the Staff had any questions concerning the use of T-1 steel for the major component support strt.ctures. Mr. O'Reilly 2
replied that, while this is not a code violation, some of the Staff
, r~ 1;c; feel it to be poor practice. ne structural tests will be gone into in some detail. Dere is at this time no more information available.
The fabricator of the support structures is the same vendor who made the sn W. de.d.
reinforcing ring which f ailed previcusly,44-+he. situ :icr. i; 5:ing 12:hd dee+. Mr. O'Reilly also pointed out that the support structures are relatively simple whe com ared with tha. reinforcing ring.
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DRAFT - 99th Meeting fh"&f" LfA U
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Part II, Page 9
.. , , f C. DRL Report - Category B Items Big Rock Point Reactor The question of Change No.14, which would allow a partial
' loading of pellet fuel to be run at high power levels, was raised.
in &dm Mr. Skovholt noted first that a report had been received,th b all three of the powder. compact fuelbundles(ofsixtotal)'"-!!ee)in the center-melt program had f ailed. In addition, of the intermediate bundles,( , those run with incipient center melting) one was found to have blistered cladding on one rod. Hydriding of the zirconium
,n v + r..s:.
is suspected y It was noted that the licensee is still obligated to keep h
^
L the two advanced burnup"bundles out of the reactor untti investigative
- information resulting from the first set of destructive tests is available.
I 1 Dr. Monson asked if it was clear that hydriding was the cause of f ailure of the powdergnpact elements. The report received had indicated that it was not cleae. Mr. Skovholt replied that the Regulatory Staff does not knew but that the licensee thought it was hydriding on the s basis of visual observationg of the defects.
Dr. Monson noted that the report had indicated that the activity release rate increased and subsequently the power level was reduced.
l He asked if it was in fact the case that a 107. reduction in power was undertaken because MCHFR was being approached too closely. Mr. Shea !
replied that this was irrfe+t the case and that the pcver level never reached the 240 MWt predicted. The power profile was soms. hat peculiar ,
i because of the nature of the core. The profile was not that which had l l
W A*g ;
DRATT - 99th Meeting -
Part 11. Page 10 been predicted when the Staf f reviewed the proposal. Mr. Shea also pointed out that it was after the drop in power that an eWvde increase in radioactivity was % A rather complicated radio-chemical analysis of the stack gases was undertaken which indicated that the activity might have been released from the center. melt fuel. This was done prior to reactor shutdown.
Dr. Monson asked if this set of circumstances would invalidate the results of the center-ceit experiment. Mr. Skovholt replied that the licensee was now more concerned about the need to keep the two advanced burnup bundles out of the core, but unless they should be
- damaged, etc., the experiment should proceed satisfactorily. Mr. Shea
) added that it had been anticipated that a flatter flux distribution
' would be available over the bundles. A sharp radial and axial peak
! cewl69 in was actually present, however, 4; An u fewer rods which av design f
i conditions, but it is thought these will be sufficient . At a maximum, only about f4J 4000 to 5000 megawatt days per ton burnup was achieved j l
in the hottest rods. l There was some discussion of the '\ reload ( fuel. This will be representative of product line fuel from now on. Dr. Okrent asked f
what would be the change in peak temperature between the current driver i ,
l fuel and the reload E fuel following a loss-of-coolant accident.
Mr. Shta thought the peak in the reload E fuel would be about 300 F l
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w-DRAFT - 99th Meeting .. ',
Part II, Page 11 higher than that for the current driver fuel. It was not clear p iles,i% d eu * % C* A w*
how many additional rods would f ail because of the use of reload E h
fuel,i+; - * : = c f :::Iri
, Dr. Okrent thought it was not clear that the peak temperatures could be kept so low with only one core spray system. Mr. Shea pointed out that the elements are only five feet long, and that the rates of water input with the emergency core cooling systems are not T\ep.
too much different from thirt for more recent designs. That is, they are about two spm per fuel bundle.
Contamination 1.evels at the Flum Brook Reactor
}
i Mr. O'Reilly noted that the Stif f is in frequent contact with i
the Plum Brook operating group. With respect to the 73,000 dps per
! square centimeter reported, this represented the maximum value found
)
anywhere on site. It occurred inside a chemistry lab hood and folicwed A spillL*g of a fuel capsule o a lead brick. No personnel contamination resulted,and no airborne activity was detected in the laboratory. In addition, the 350 rads of beta-ganna contamination reported in the anr.ual report were the result of contact readings inside hot cells. i t I
- The Staff feels there is no problem in connection with this level of
{
contaminations (A 156b b '*
- It has been suggested that Wk.Lamual reports be written more g
1 carefully,Li i M m w 1 I
)
I
DRAFT - 99th Meeting PART II, Page 12 bQh w; s . . *)
MH-1A Mr. Skovholt recalled for the Coernittee that the Mi-1A is a floating power plant located on the karge Sturgis. The Army intends to deploy it to the canal Zone from its present location at Fort Belvoir. The Staff has been reviewing this and will shortly have completed its review. The Army would like to have permission to 0
move the barge-:: %21 :: :: July 15.cc p : sib h. The Regulatory Staff now thinks that the move will be satisf actory but is not yet ecepletely satisfied. Mr. Skovholt asked if the Comittee wanted to review the I
deployment to the Canal Zone.
Dr. Hanauer asked if e< ARCHAS had reviewed the move. Hr.
?
l Skovholt replied that they had and that their review was transmitted i
to the Regulatory Staff. A complete safety analysis report had not i
I been published, however.
Later in the meeting, Dr. Zabel proposed leaving the review of the MH-1A to the Regulatory Staf f and having the Staff report on the results to the Comittee. There seemed agreement on the part of the i Cournittee to this approach.
, palisades Reactor Review Mr. Boyd reported that the emergency core cooling and reactor protection system reviews are now both underway. It appears as though i
the codes and multe that Combustion Engineering is tsing in analyzing
, l G-s, its energency core cooling systan are similar to those e4 B&W and a c,.w e 4 < % 1 h. %
Westinghouse,de++gns. The final ECCS design identifies two independent )
l cooling systems,and these are such that any single passive f ailure can i
be tolerated. In addi that check valves in l
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DRAFT - 99th Meeting k "-
Part II, Page 13 accumulators will be tested, etc. Mr. Boyd noted that the Staff's review has not yet been completed.
Mr. Moore reported on the Staff's review of the protection system.
, In the Palisades plant, this is the same system as that which has been reviewed in connection with the Fort Calhoun review and is therefore in fairly good shape. The ECCS actuation circuits have not yet been reviewed, however. Mr. Boyd added that the Staff is planning on getting pteliminary information concerning moderator coefficients
! ~4 :In..m i.us 41G uv .vai M k from CE. He hoped to have i
sufficient informationbo allow a briefing to the Comittee by the
! l g September meeting.
p In answer to Dr. Zabel, it was noted that the dome is presently
- being placed on the containment building. The applicant plans to file for a provisional operating license in November, 1968 and '.t is usual 3 i s 64 ue.l bd .s e e a u t. A ,t that one yearn Na the t.ime of filing to the +&me of operation,6s
.:.1 l e ve d .-
Con Ed Nuclear Units 4 and 5 i
Dr. Okrent asked what the status was of the Staff's comaus dun 4 d.i. ia review. Mr. Boyd noted that the Staff is now waiting for the Comittee to indicate its pleasure. Dr. Morris added that Con Ed
} had asked im 3 u :=::ir a whether the sites would be turned down categorically on the basis of population density alone. 1.ater in the meeting, the Comittee decided to consider the Consolidated Edison site proposals at its 100th meeting in August, 1968. It was
, further agreed that there would be an extended period of time allowed
/
for internal Comi a meeting with Consolidated
DRAFT - 99th Meeting i h# Wh Part II, Page 14 , , ,- ,.
Edison personnel.
D. DRS Report Mr. Case noted that the ASME Code Comittee has been working elosely with DRS in formulating criteria for in-service inspection.
Three meetings have been held b i*cun Lii t s , evy a since March.
Agreement has been reached by negotiation and compromise onrany aspects of the criteria although two major areas of disagreement are left. The most dif ficult of these is pressure vessel accessibility, i
I eithcugh The Comit tee's 44e546 requhmcw4 e f to the complete pressure access A
, vessel volume from either the inside or the outside are still in the 3 e,w s L d.
d c ; r = ty, the industrial representatives had suggested inspection
, of welds only at the top and bottom and not in the beltline region f
since they feel that the top and bottom velds are these most highly stressed, etc.
Mr. Case pointed out that the in& strial representatives cannot get backfitting problems out of their minds. It r,eems that they feel that,1f they agree to access for inspection on future vessels,the ACRS l
or the Regulatory Staff will require similar inspection on present vessels. In this connection, Mr. Case pointed out that the industrial l representatives had had rather more f aith in ultrasonic inspections i l
that the DR3 personnel. They feel that such inspections could be done satisfactorily either fron the inside or the outside of pressure vessels. The other area of disagreement concerned periodic hydrostatic l
tests. The DRS representatives had started by asking for pressure testing at 125% of design pressure. and the industrial representatives #
1
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!"~4)I[.F#'WHU1-Q DRAFT - 99th Meeting Part II, Page 15 initial position had been foi operating pressure tests only. The DRS personnel have since backed off to 1107. of design pressure *, end t re.n the industrial representatives are reluctet to come this far. For
% , c b a n w 6 4. gh '
the rest, th:y hec; .ce;pd about 907. of the requirements k the m lini Mr. Case noted that he would like to go back to the m::t hg wich the Code Comittee with some expression frem the ACRS that backfitting of inspection requirements is not the Comittee's intention.
Later in the meeting, the Comittee voted that the following woulo be the ACRS position regarding periodic reactor pressure vessel I
inspection requirements:
"With regard to the implementation of the proposed in-t l mervice inspection standard, the ACRS believes that routine
) in-service inspection requirements of reactor pressure vessels in use or under construction should continue to reflect recognition of built in limitations on the inspectability of such vessels wW t-except, evidence of pressure vessel detsrioration dictates other-vise."
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DRAFT - 99th Meeting Part III III. Hanford_N- Reactor *
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A. Subcomittee Report Dr. Gifford recalled that the present evaluation of the offluent control project is the latest in a series of such reviews concerning ef fluent con-trol, all of which stem from an ACRS letter of March 26, 1965. In thi/
that letter, the Coenittee discussed making the production reactors equi-valent in safety to commercial power reactors.
The effluent control project involves primarily the emergency core
, cooling and ef fluent vaste disposal systems. The changes will upgrade these i
systems and make them cozparable with similar systems on cornercial power
, reacters.
t
} TheX reactor itself is a large, graphite moderated, water-cooled reactor.
I The fuel is circaloy clad and is designed for Pu production, i
Douglas United Nuclear (DUN) contends that they will achieve equivalence in safety to cocnercial power plants by the changes proposed, although details concerning meeting the 70 General Design Criteria, etc.,
have not yet been settled with the Regulatory Staff. The reason for this l
\
t early review is the standard problem with AEC owned reactors. That is, ;
some sort of indication is necessary that the proposal is satisfactory f before money can be spent and contracts ler. Dr. Gif ford thought that ;
! the minimum need at the present time was similar to what the Comittee l had said in connection Y Q
2FrTr% -
& ft L,.4;m W +n --liCC_Q DRAIT - 99th Meeting ,, ,
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Part III, Page 2 with the tall stack proposal at Savannah River. That is, that the ACRS feels that the proposal is necessary as a first step on the road to licenh<,.%1.Q'
.i L.. ny. Dr. O' Kelly added that the Cocnittee should note that it is not intended that N-Reactor be Itcensed.
Dr. Gif ford thought that the possibility of licensing or some legally equivalent step was not being ruled out but was not being requested at the present time. In essencej DUN would like a letter saying that N-Reactor is equivalent to the coceerch 1 power reactors cs in safety or yas an acceptable, although not as desirable alternative, i N the codifications proposed are necessary as a f f,rst step toward caking N-Reactor the equivalent in safety of the cornercial power 1
y reactors, e
Dr. Gifford continued with the Subcoccittee report. The tech-h nical proposals compri the ef fluent control projer:t are listed ir-[able 1 of the btaffs report to the Coccittee. S- F cc pen:n:c w L e u. . g e .,y n ? m r eye *a- th- qua*=4, :::,, =: i-frdt . D'JN has been asked for a presentation on the emergency core cooling system and on itmna as yet unresolved with the Regulatory Staff. In addition, they have been asked to state clearly what it is they want the Cocnittee to do.
Dr. Isbin observed that the applicac. eproposes spending 1.8 million dollars to improve the safety of the plant and felt that there was no real question as to whether safety would in fact be improved. He asked if the Coenittee was trying to rake some judge-ment as to the advisability of the expenditure.
Musummum. 4
.- W _i c T4T-,41 mm2 .
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DRArt - 99th Meeting Part III, Page 3 Dr. Hendrio observed that DUN had embarked on the endeavor because they were requested to do so in a past Coconittee letter.
This program is being proposed in response to that letter and the
. Committee is now being asked if the steps suggested are necessary in the Coenittee's review. An affirmative answer is required in order to allow the money to be spent.
There was additional discussion of what the C;;.-.ih was being requested M and of the approach which the Coccittee should take ,
,tc tH e prep:::1. There seemed to be agreement that the Coenittee should proceed on the basis of determining whether the proposal l represented a necessary step toward the ultimate achievement of cc g equivaleh. in safety between N-Reactor and the cocnercial power I
reactors.
B. Meeting with the Regulatory Staf f Mr. Ziemann opened by stating that there was not much to add to the staff report submitted--te tba C^-it::
y last Mayj on the effluent control program. More recent discussions with EUN have indicated that the required tests will be performed. Although not much of the work done has been reportedj Mr. Ziemann felt that DUN had been extremely responsive to requests of the Regulatory Staf f and had developed a good R&D program related t.o resolution of tha uncertaintie:J ' 'fM. He had noi changes tc report in the conclusions of the last staf f report to the Coenittee on the effluent control program. -
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DRAIT - 99th Meeting h_:r C
Part III, Page 4 Dr. Gifford asked Mr. Ziemann to state what his impression was of DUN's request. Mr. Ziemann thought that, ideally,they would like to be told that the facility would be licensable when the proposed In [eu of that th, .vn} Ls 44 me d changes are made. dla. v ... a that the j
proposed modifications are desirable and are required for licensa-s o*\k s* M i '**
bility at some future date n The Regulatory Staff feels that the proposed changes will enhance the safety of the plant and should be made. It is not possible to say at this time that the plant 's i
, licensable or comparable in all respects to licensed plants. The entire facility has not been reviewed in detail. Mr. Ziemann ob-f servedj in answer to a question by Dr. Hanauer j that the staff was I t u % .4, . .,
t not precluding,that as proposed j the plant might be licensable.
t
, Dr. Okrent asked if DRL was satisfied that the earthquake i
which has been considered in the design which would not lead to a t'% +
loss of emergency cooling. Mr. Ziemann replied that was the case.
a
$ ( 6e, k m.f C.$
Such ::17: w.ae as the instrumentation, however, have not been evaluated with respect to their seismic response.
C. Ibugl.is Unit ed Nuclear Tae objectives of the N-Reactor ef fluent control program were identified as follows:
s
- 1. To meet the limits established in 10 CTR Part 20 for routine releases.
- 2. To meet the limits established in 10 CFR Part 100 for accident releases.
.. . _- _ m p -
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DRAFT - 99th Meeting Part III, Page 5 .
- 3. To comply with the intent of the 70 General Design Criteria.
t Coals which have been set include:
, 1. Routine operation within 10 CTR Part 20; on the average this will mean releases of the order of about 1 percent of the limits established in Part 20.
C v (.t.
- 2. Comply with 10 CFR Part 100 guidelines, including an acceptable u rgin.
{ 3. ComphNith the intent of the 70 General Design
} Criteria.
E f fluent Control Pro ject l
Mr. Robinson described the project as approximately 907. com-a plete with respect to the detailed design. One major goal of the project is to increase the reliability of the emergency systems to assure the availability of cooling water. Improved effluent control and additional wide-range nonitoring of stack ef fluents are also i
major goals of the project. ,
lu .b d The existing emergency cooling system consists of two, diesel-t k a-driven pumps supplying A emergency raw water tank and three high.
h lift diesel'1,.v.n pumps supplying the reactor. One of the problems with tkt e*udd t*4+ system is that there is always raw water in the system which leads to the continuous potential for its injection into the building, i
~~
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DRAFT - 99th Meeting .-
Part III, Page 6 etc Tht.re is also no data concerning the ,'terials properties of the steel in the ERW tank.
A flow deflector will be added to A syetem which will allow filling the ERW tank with clean water which would be used initially if emergency cooling water were required. The !!:u def!:::cr ;;uld em nt u:ter f rem the !!eu 8 !f t-di ::1: te the m 'i^a ih af the iiid. lif t Ai m ie. Mr. Robinson pointed out that the pumps all operate on every scram so that a tank bypass would be necessary to enintain clean water in the tank. The deflector will be a
- s. g we s
.cypsme type, designed so that whenever the tank is full there will be
, an air cushion between the river water and the tank. When the tank
- l. scale level drops river water would be admitted to the tank. AhalfYmock-i
?
up of the deflector has been built and tested. Failure of the ERW tank will not cause failure of *.he deflector. The elevations are such that failure will not cause praferential flow to the tank and back out to the river. The necessary piping would also be protected.
Interties with other existing clean water tanks will be provided. l These include the 1 million gallori demineralized water storage tank, and the 850,000 filtered water storage tank. This will prr vide about ,
eS a w. l
, one hour emergency cooling before it is necessary to, raw water. Re- l note operats d valves will be provided en the interties. After the l system has been completed these valves may be put on automatic or )
l 1 eft continuously open but this is not the plan for the initial shake- !
down period.
i
-Ji.'
DRAFT - 99th Meeting Part III, Page 7 Mr. Robinson pointed out that it is not intended to replace the common line between the interties and the pumps. The maximum th e I'm 4 pressure possible in this line is 2fsig and 44 has been tested to 128psig. The maximum stress to which this line would be sub-jected is less than 1500 psi. This is so low that sufficient re-liability is assured. The22fsigiscontinuouslymaintainedby the water head in the ERW tank. The common line could be retested at higher pressures periodically although it would be a difficult procedure. In the event of an t.ccident the line is required as j
I long as cleah water is supplied Q1though other alternatives are i bV
, availab1p eks.e would require the manipulation of valves. It t is felt that the common line would be required for several hours l
l following an accident.
4 Earthquake effects on this line have been analyzed. The addi-i tional stress is minimal. The penetration of this line into the reactor building as well as the new pipe junctions in the lines have also been evaluated in the same way. The line itself is located s
below ground.
Another part of the project will be the addition of a second i ERW line from the high lif t dump discharge to the front face of the i reactor. Operation will normally be carried out with the new line valved in. There wi'.1 be instrumentation provided which would allow detection of a failure of this lineand heveral minutes would be I available following a loss of coolant to detect such a failure and
M_ TC ____ m DRAPI - 99th Meeting .(
Part III, Page 8 '
switch to the old ERW line. The reason for the installation of the new line is that the brittle fracture characteristics of the existing pipe are not known.
Mr. Robinson pointed out that a program is being initiated to l o od check the kw distribution on the pipe supports in order to assure de that the earthquake analysis applys to the as-built system.
Mr. Robinson also noted that DUN does not intend to replace the manifolds on the front face of the reactory dithough the material which is removed when the new ERW line is connected to the manifold will be used to determine the brittle fracture characteristics of the
. manifold.
Mr. Robinson described the stresses which will exist in the piping.
DUN is confident that the cor:xnon manifold sees sufficiently low stresses
,f to assure its continued integrity. In addition, it is hoped that the i data which will be developed using the samples noted above will give added confidence in the reliability of the manifold.
The ef fects of a pipe b1eak on other piping located on the front 1
i face of the reactor have been evaluated. DUN is confident that such j S i a pipe break will not cause progressive failures. Sensor jand other ;
, equipment j vulnerability to pipe whipping has also been investigated.
Other system characteristics which give confidence in its con-tinued reliability are the fact that the pumps will be located in individual cella with five foot s thick walls and two fto three-foot I thick roofs. pensors and instrument leads will be widely separated.
j l
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DRAFT - 99th Meeting -
Part III, Page 9 hc The thermal shock potential of cold water injection ha ve 1 been 1
~
analyzed. The analysis has shown that the transients which occur as a result are less severe than the design transients. With 1 Ke s ben respect to the fuel, th: E. 1 ha sLcca & designed for normal startup ,
. A ;
and shutdown. Approximacely 90 scrams have been undergone without '
fuel damage. In addition e adiated fuel has been heated to I
approximately 19000 F. and quenched in 1500 F. water without damage. ,
Dr. Okrent asked if there are any effects on the Bonneville
, Power Administration Grid as a result of a trip from 800 megawatts. l s
! Mr. Robinson thought that the reactor transient was milder when the station was on line than otherwise. The WSS system continues
, to take steam as the power drops so that the transient on the i
e i reactor is not so severe. The plants 800 megawatt electrical h
" O rk4 output is fed into the BPA grid remotely from wh m -..I tc.h ; power t c. < e % .
hca the grrd. Tests have been run including an intentional trip l
from 800 megawatts electrical and operation of the plant can continue. l Dr. Okrent asked if the present discussion concerning the i
reliability and the confidence in the fuel design could be related 4
to future fuel types, thatis,whatwouldbetheeffect4t(ene Ica.K e 5 F:=leM driving the fuel harder on that separation of the steam o
1 l
?
from the fission products, 31 hi. ::t ba ^^ c2:c, etc. Mr. Robinson replied that the normal procedure is to provide analyses, tests, and presentations for approval whenever fuel design or power level are changed significantly. A specifick type dhis therefere under consideration at present.
b
~
DRAFT - 99th Meeting -
Part III, Page 10 Dr. Okrent pr =it6 m.iim .1;h ou acau.wu.1 coment4d 3at % this would be an open question if the design er power level were changed. Mr. Nechodom agreed, but pointed out that in any case each fuel design would meet the same raquirements of no fission product release under the accidents hypothesized.
- fr. Robinson continued with his sumary. It was noted that pipe breaks in some locations will put forces on the fuel which might cause it to move. Therefore, impact tests on fuel and f
spacer assemblies have been performed to assure that the nozzles, I
etc. remain intact and that cooling can continue. Photographa k of spacers which were collapsed by various impacts were distributed.
I L
1 It was noted that the actual effects would really be less than I thoseobserved$nthephotographssinceamultiplespacer i
arrangement is used in the reactor,and tests on this arrangement i ac.
have shown a distribution of the impact energy among the spaers. j
^ .
oh The photographs show the results of tests of single spacers which ed I
involve a greater degree of collapse but still maintain,a coolant path.
- o. 1 The analyses indicate that the system will allow continued (
s supply of water to the reactor. In addition, the reliability f
I of the diesels, etc. will be improved as part of the effluent control project. At present the fuel supply line runs from fL the storage tank to a. low lif t diese11'then to a day tank, to
&r '%
y high lif t diesels and to the fog spray pumps. This will be modified to provide three day tanks arranged so that each tank
.s.
is associated with' M ligh lift pumps and one fog spray pump. Jailure of any one of these systems will
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DRAFT - 99th Meeting ..
Part II, Page 11 higher than that for the current driver fuel. It was not clear g s ilos m a.ien * \ Cs A w4 how many additional rods would fail because of the use of reload E A
fuel,k --': = cf :: 1 r t Dr. Okrent thought it was not clear that the peak temperatures could be kept so low with only one core spray system. Mr. Shea pointed out that the elements are only five feet long, ard that the rates of water input with the emergency core cooling systems are not d en.
too much different from thtre for more recent designs. That is, they 4
j are about two gpm per fuel bundle.
Contamination Levels at the Plum Brook Reactor i
{
6 Mr. O!Reilly noted that the Staff is in frequent contact with
)
I the Plum Brook operating group. With respect to the 73,000 dps per I '
)
1 square centimeter reported, this represented the maximum value found I anywhere on site. It occurred inside a chemistry lab hood and followed fa,spillksofafuelcapsuleoIaleadbrick. No personnel contamination resulted,and no airborne activity was detected in the laboratory. In
{
addition, the 350 rods of beta-ganrna contamination reported in the 1
annual report were the result of contact readings inside hot cells. )
The Staff feels there is no problem in connection with this level of i contaminations $6 Sud A '***
It has been suggested that $ w.ki L g amual reports be written more carefullyJu uw ' tuco 0
, .g e *
- , - , A DRAFT - 99th Meeting - d :-E R E PART II, Page 12 ; -
e . . .'>
MH-1A Mr. Skovholt recalled for the Comittee that the MH-1A is a floatingpowerplantlocatedonthekargeSturgis. The Army intends to deploy it to the canal Zone from its present location at Fort
,! Belvoir. The Staff has been reviewing this and will shortly have completed its review. The Army would like to have permission to 0
move the berge m: %21 rec :: July 15,ee peribh. H e Regulatory Staff now thinks that the move will be satisfactory but is not yet completely satisfied. Mr. Skovhvit asked if the comittee wanted to review the I
deployment to the Canal Zone.
i Dr. Hanauer asked if t4+e ARCHAS had reviewed the move. Mr.
Skovholt replied that they had and that their review was transmitted l to the Regulatory Staff. A complete safety analysis report had not I
been published, however.
Later in the meeting, Dr. Zabel proposed leaving the review of the MH-1A to the Regulatory Staff and having the Staff report on the results to the Comittee. Here seemed agreement on the part of the r
i i Comittee to this approach. '
Palisades Reactor Review j i
Mr. Boyd reported that the emergency core cooling and reactor protection system reviews are now both underway. It appears as though the codes and-rc=1:4 that Combustion Engineering is tsing in analyzing wh s 1 its mergency core cooling systs are similar to those os B6W and o e,.* c 4 < % \ M. *k Westinghouse,deci p s. L e final ECCS design identifies two independent cooling systems,and these are such that any single passive failure can be tolerated. In additienj-@eav s that check valves in
~r ,u L A '- A T__~_
DRAFT - 99th Meeting - - - - E ~ ~ ~' -'-
O Part II, Page 13
- i accumulators will be tested, etc. Mr. Boyd noted that the Staff's review has not yet been completed.
Mr. Moore reported on the Staff's review of the protection system.
, In the Palisades plant, this is the same system as that which has been reviewed in connection with the Fort Calhoun review and is therefore in fairly good shape. The ECCS actuation circuits have not yet been reviewed,l.awever. Mr. Boyd added that the Staff is planning on getting preliminary information concerning moderator coefficients
- #
- lt:...d ive. *isivii mu ovalidae from CE. He hoped to have sufficient information to allow a briefing to the Committee by the t l j September meeting.
f In answer to Dr . Zabel, it was noted that the dome is presently being placed on the containment building. The applicant plans to file for a provisional operating license in November,1968 and it is usual
( i s c b ucJ. bk , e e a u t. ., k, b that one year f*om the time of filing to the 44me of operation,ta
. 11ered.
Con Ed Nuclear Units 4 and 5 I
Dr. Okrent asked what the status was of the Staff's cc .:ideaaluu
+f thi eits review. Mr. Boyd noted that the Staff is now waiting for f
the Comittee to indicate its pleasure. Dr. Morris added that Con Ed had asked fuc ac;..: re::ti r n whether the sites would be turned down categorically on the basis of population density alone. Later in the meeting, the Comittee decided to consider the Consolidated Edison site proposals at its 100th meeting in August, 1968. It was further agreed that there would be an extended period of time allowed f or internal Comitt e = es a s a meeting with Consolidated
-~~
DRAFT - 99th Meeting 7 Part II, Page 14 -
Edison personnel.
D. DRS Report Mr. Case noted that the ASME Code Comittee has been working closely with DRS in formulating criteria for in-service inspection.
Three meetings have been held ht cun umoc e,w ups since March.
Agreement has been reached by negotiation and compromise onnany aspects of the criteria although two major areas of disagreement I
are left. The most dif ficult of these is pressure vessel accessibility.
j eithaghTheCommittee'swes4srequbcn4et +g access to the complete pressure A
- vessel volume from either the inside or the outside are still in the O m n bel
- 1 d e c r:nty, The industrial representatives had suggested inspection i
of welds only at the top and bottom and not in the beltline region f
since they feel that the top and bottom welds are these most highly ]
l stressed, etc.
Mr. Case pointed out that the indu strial representatives cannot get backfitting problems out of their minds. It seems that they feel that,1f they agree to access for inspection on future vessels,the ACRS or the Regulatory Staff will require similar inspection on present l
. vessels. In this connection, Mr. Case pointed out that the industrial
{ representatives had had rather more faith in ultrasonic inspections that the DRS personnel. They feel that such inspections could be done satisfactorily either from the inside or the outside of pressure vessels. The other area of disagreement concerned periodic hydrostatic tests. The DRS representatives had started by asking for pressure testing at 1257. of design pres _sure, _and the-industrial representative /
l 1
- l g n .. - 2 i
_ _1 __
DRAFT - 99th Meeting Part II, Page 15 l initial position had been for operating pressure tests only. The DRS personnel have since backed off to 1107, of design pressure', and
- e. r t. n 1 the industrial representatives are reluctant to come this far. For b . e_ be w c < 6 4 gI4
. the rest, they hcc acccp::d about 907. of the requirements k the e.-ia.;
Mr. Case noted that he would like to go back to t% r::th; with the Code Comittee with some expression from the ACRS that backfitting of inspection requirements is not the Comittee's intention. I Later in the meeting, the Comittee voted that the following
- would be the ACRS position regarding periodic reactor pressure vessel i
inspection requirements:
f "With regard to the implementation of the proposed in-service inspection standard, the ACRS believes that routine g in-service inspection requirements of reactor pressure vessels e
, in use or under construction should continue to reflect recognition of built in limitations on the inspectability of such vessels e > % + L-except, evidence of pressure vessel deterioration dictates other-wise."
1 O
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DRAFT - 99th Meeting Part III III. Hanford N- Reactor '""N}
A. (Subcomittee Reportn , s,n s.sh J.W mtf a b'P" Ye
Dr. Gifford recalled that the present evaluation of the effluent control project is the latest in a series of such reviews concerning effluent con-trol, all of which stem from an ACRS letter of March 26, 1965. In f4/(
that letter, the Comittee discussed making the production reactors equi-valent in safety to commercial power reactors.
The effluent control project involves primarily the emergency core cooling and ef fluent waste disposal systems. The changes will upgrade these systems and make them comparable with similar systems on comercial power
~
, reactors.
] Thex reactor itself is a large, graphite-moderated, water-cooled reactor.
( The fuel is zircaloy clad and is designed for Pu production.
Douglas United Nuclear (DUN) contends that they will achieve f
4 equivalence in safety to comercial power plants by the changes proposed, although details concerning meeting the 70 General Design Criteria, etc.,
have not yet been settled with the Regulatory Staff. The reason for this early review is the standard problem with AEC owned reactors. That is, some sort of indication is necessary that the proposal is satisfactory
, before money can be spent and contracts let. Dr. Gif ford thought that the minimum need at the present time was similar to what the Comittee had said in connection Y
~
_ :Prewi: i ir. -
n w& -
2:" _____ Z _~ . , x l DRAPI - 99th Heeting ,, , ; -)
Part III, Page 2 with the tall stack proposal at Savannah River. That is, that the ACRS feels that the proposal is necessary as a first step on the roadtolicenbeb.1.Q.
nau lty . Dr. O' Kelly added that the Committee should note that it is not intended that N-Reactor be licensed.
Dr. Gifford thought that the possibility of licensing or some legally equivalent step was not being ruled out but was not being requested at the present time. In essencej DUN would like a letter saying that N-Reactor is equivalent to the comerch 1 power reactors a
in safety or yas an acceptable, although not as desirable alternative, t N the modifications proposed are nece'ssary as a first step toward
, making N-Reactor the equivalent in safety of the cocnercial power i
t reactors.
Dr. Gif ford continued with the Subcocnittee report. The tech-h nical proposals compri the effluent control project are listed in able 1 of the ktaffs report to the Comittee. kh ec en r.;S l
-as the u.:rgem, r: . - -
'n m eye * = *ha qu =~k *M, :::,, cr: i-fautt. DUN has been asked for a presentation on the emergency core cooling system and on items as yet unresolved with the Regulatory Staff. In addition, they have been asked to state clearly what it is they want the Comittee to do. l l
l
} Dr. Isbin observed that the applicant proposes spending 1.8 I million dollars to imarove the safety of the plant and felt that I there was no real question as to whether safety would in fact be improved. He asked if the Comittee was trying to make some judge-ment as to the advisability of the expenditure.
4 '
l l
l 1
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DRAFI - 99th Meeting Part III, Page 3 Dr. Hendrie observed that DUN had embarked on the endeavor because they were requested to do so in a past Committee letter.
This program is being proposed in response to that letter and the l
. Committee is now being acked if the steps suggested are necessary I in the Committee's review. An affirmative answer is required in order to allow the money to be spent. ;
1 There was additional discussion of what the C:;;.i;isc was being requested 4*rmhy and of the approach which the Committee should take ,
j Ac thi: pr:p ::1. There seemed to be agreement that the Committee i
i should proceed on the basis of determining whether the proposal I
( represented a necessary step toward the ultimate achievement of n en equivalehts in safety between N-Reactor and the commercial power i
reactors.
l 4 B. Meeting with the Regulatory Staff Mr. Ziemann opened by stating that there was not much to add to the staff reporty submitted te tha carr!:::: last May on the j
f effluent control program. More recent discussions with DUN have t
indicated that the required tests will be performed. Although not l much of the work done has been reportedj Mr. Ziemann felt that DUN had been extremely responsive to requests of the Regulatory Staff f
and had developed a good R&D program related to resolution of the uncertainties.4'- ^^ 4.-
He had noi changes to report in the conclusions of the last staff report to the Committee on the effluent control program.
[
y
- - z:-san _
DRAFT - 99th Meeting Part III, Page 4 Dr. Gifford asked Mr. Ziemann to state what his impression was of DUN's request. Mr. Ziemann thought that, ideally,they would like to i.e told that the facility would be licensaule when the proposed I.
In leu c. s of that the, .vu{sAr me d changes are made. d liku 6v u w that the proposed modifications are desirable and are required for licensa-v a iis* %i
5111ty at some future date s The Regulatory Staff feels that the proposed changes will enhance the safety of the plant and should be made. It is not possible to say at this time that the plant is licensable or comparable in all respects to licensed plants. The entire facility has not been reviewed in detail. Mr. Ziemann ob-I servedj in answer to a question by Dr. Hanauerythat the staff was
% w\d,n not precluding,that as proposedj the plant might be licensable.
- Dr. Okrent asked if DRL was satisfied that the earthquake i
i which has been considered in the design which would not lead to a I tkd i
loss of emergency cooling. Mr. Ziemann replied that was the case.
4 twkWfCS Such (tempeaeun as the instrumentation, however, have not been evaluated with respect to their seismic response.
i C. Douglas United Nuclear The objectives of the N-Reactor effluent control program were identified as follows:
E
- 1. To meet the limits established in 10 CFR Part 20 for routine releases.
- 2. To meet the limits established in 10 CFR Part 100 for accident releases.
J J_~ _ _ . ~~;
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^
- & R-h DRAFT - 99th Meeting Part III, Page 5
- 3. To comply with the intent of the 70 General Design Criteria.
Goals which have been set include:
.- 1. Routine operation within 10 CFR Part 20; on the average r
this will mean releases of the order of about 1 percent of the limits established in Part 20.
Ev tx
- 2. Comply with 10 CFR Part 100 guidei nesj including an acceptable margin,
- 3. Comp pwit
(,*h the intent of the 70 General Design 1
- Criteria.
1 U
Ef fluent Control Pro ject 6
l Mr. Robinson described the project as approximately 907. com-I plete with respect to the detailed design. One major goal of the project is to increase the reliability of the emergency systems to assure the availability of cooling water. Improved effluent control l
and additional wide-range monitoring of stack effluents are also i,
i major goals of the project. .
how h d The existing emergency cooling system consists of two, diesel- )
the-
! driven pumps supplying dre emergency raw water tank and three high.
Alam
'.' lift diese1{ pumps supplying the reactor. One of the problems with i
j tke9med 4Me system is that there is always raw water in the system which leads to the continuous potential for its injection into the building,
_,arr m . 2 = = ,7 _ _:
_% 5 -:_.
- QJ DRAFI - 99th Meeting .
Part III, Page 6 etc There is also no data concernin8 the materiala properties of the steel in the ERW tank.
A flow deflector will be added to the system which will allow filling the ERW tank with clean water which would be used initially if emergency cooling water were required. The f1= deflects: ;=1d Gu gt arter f re the *1^o-lif t- di:::10 to the r - ina r i&--o f th idgh lif t A4aeale. Mr. Robinson pointed out that the pumps all operate on every scram so that a tank bypass would be necessary to maintain clean water in the tank. The deflector will be a
' Schu s
.cypmm type, designed so that whenever the tank is full there will be an air cushion between the river water and the tank. When the tank
?
scale l
i level drops river water would be admitted to the tank. Ahalffmock-up of the deflector has been built and tested. Failure of the ERW i
j tank will not cause failure of the deflector. The elevations are i
such that failure will not cause preferential flow to the tank and back out to the river. The necessary piping would also be protected.
Interties with other existing clean water tanks will be provided.
i These include the 1 million gallon demineralized water storage tank, and the 850,000 filtered water storage tank. This will provide about j 6 a u. '
one hour emergency cooling before it is necessary to raw water. Re-l
$ cote operated valves will be provided Gn the interties. After the system has been completed these valves may be put on automatic or lef t continuously open but this is not the plan for the initial shake-down period.
l I
- k. . . , . - _ 7 "*'&. g DRAFI - 99th Meeting Part III, Page 7 Mr. Robinson pointed out that it is not intended to replace the cocoon line between the interties and the pumps. The maximum t lu l'.n
- pressurepossibleinthislineis2fsigand44hasbeentested to 128psig. The maximum stress to which this line would be sub-jected is less than 1500 psi. This is so low that sufficient re-liability is assured. The22fsigiscontinuouslymaintainedby the water head in the ERW tank. The comon line could be retested at higher pressures periodically although it would be a difficult procedure. In the event of an accident, the line is required as long as cleah water is supplied Q1though other alternatives are
- bd availabicy Ase would require the manipulation of valves. It I is felt that the cocrnon line would be required for several hours l
, following an accident.
l
[ Earthquake effects on this line have been analyzed. The addi- l tional stress is minimal. The penetration of this line into the reactor building as well as the new pipe junctions in the lines have
, also been evaluated in the same way. The line itself is located below ground.
Another part of the project will be the addition of a second P
ERW line from the high lif t fump discharge to the front face of the reactor. Operation will normally be carried out with the new line valved in. There will be instrumentation provided which would allow detection of a failure of this linegrrd heveral minutes would be available following a loss of coolant to detect such a failure and
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- [.1 DRAFI - 99th Meeting ...
Part III, Page 8 switch to the old ERW line. The reason for the installation of the new line is that the brittle fracture characteristics of the existing pipe are not known.
Mr. Robinson pointed out that a program is being initiated to
. lo od check the hw distribution on the pipe supports in order to assure Je that the earthquake analysis apply,s to the as-built system.
Mr. Robinson also noted that DUN does not intend to replace the manifolds on the front face of the reactor y 41though the material which is removed when the new ERW line is connected to the manifold
!. will be used to determine the brittle fracture characteristics of the manifold.
j Mr. Robinson described the stresses which will exist in the piping, i
DUN is confident that the cocmon manifold sees sufficiently low streses to assure its continued integrity. In addition, it is hoped that the f data which will be developed using the samples noted above will give added confidence in the reliability of the manifold.
The effects of a pipe bresk on other piping located on the front face of the reactor have been evaluated. DUN is confident that such o 5 a pipe break will not cause progressive failures. Sensor,and other equipmentj vulnerability to pipe whipping has also been investigated.
. Other system characteristics which give confidence in its con-tinued reliability are the fact that the pumps will be located in individual cells with five foot thick walls and two-to three-foot S
thick roofs. pensors and instrument leads will be widely separated. l I
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n my i
" g""- k - , ..-A DRAFI - 99th Meeting
- Part III, Page 9 he The thermal shock potential of cold water injection ha $ been ue analyzed. The analysis has shown that the transients which occur as a result are less severe than the design transients. With hu bta respect to the fuel, th: L:1 hasL::a & designed hfor normal startup and shutdown. Approximately 90 scrams have been undergone without fuel damage. In addition a diated fuel has been heated to approximately 19000 F. and quenched in 1500 F. water without damage.
Dr. Okrent asked if there are any ef fects on the Bonneville' Power Administration Grid as a result of a trip from 800 megawatts.
8 Mr. Robinson thought that the reactor transient was milder when i
the station was on line than otherwise. The WSS system continues i
! to take steam as the power drops so tha.t the transient on the
. reactor is not so severe. The plants 800 megawatt electrical i
+* rk i output is fed into the BPA grid remotely from whui -..I tah:: power t w < . e gg.
1 4rca the e J. Tests have been run including an intentional trip from 800 megawatts electrical and operation of the plant can continue.
Dr. Okrent asked if the present discussion concerning the i reliability and the confidence in the fuel design could be related 1 *ir to future fuel types, that is, what would be the effect4 % ese ha.K e,5 I p:=1ced driving the fuel harder on that separation of the steam I
! from the fission products, ci6ht :t ba *" :: , etc. Mr. Robinson re, lied that the normal procedure is to provide analyses, tests, and presentations for approval whenever fuel design or power level are changed significantly. A specific k type Mhis therafme under consideration at present.
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- 1: L : U .T. w t he 3 DRAFT - 99th Meeting Part III, Page 10 Dr. Okrent pr:xd L6 m Livi 11 au acau.ivu.1 comentd hat % this would be an open question if the design or power level were changed. Mr. Nechodom agreed, but pointed out that in any case each fuel design would meet the same requirements of no fission product release under the accidents hypothesized.
Mr. Robinson continued with his sumary. It was noted that pipe breaks in some locations will put forces on the fuel which might cause it to move. Therefore, impact tests on fuel and spacer assemblies have been performed to assure that the nozzles, i
l etc. remain intact and that cooling can continue. Photographs of spacers which were collapsed by various impacts were distributed.
! It was noted that the actual effects would really be less than those observed $n the photographs since a multiple spacer i
! arrangement is used in the reactor,and tests on this arrangement l ac-have shown a distribution of the impact energy among the spasers.
^
og Ihe photographs show the results of tests of single spacers which ea involve a greater degree of collapse but still maintain,a coolant 5 path.
o.
The analyses indicate that the system will allow continued supply of water to the reactor. In addition, the reliability i of the diesels, etc. will be improved as part of the effluent control project. At present the fuel supply line runs from 66 the storage tank to a, low lif t diesell*then to a day tank, to 6t jr high lift diesels and to the fog spray pumps. This will be modified to provide three day tanks arranged so that each tank is associated with' %y'OhVo(lthe pWftli ligh lif t pumps and one fog spray pump. Failure of any one of these systems will
j
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2 M.. _ _ ._- 1 DP).FT - 99th Meeting Pait III, Fage 11 1
1 leave sufficient pumps operable to meet the functional criteria.
In addition, gravity feed will be used for the fuel, and the ;
diesels will be separated with firewalls. Fog sprays will be 1 provided above the diesels as well as curbing and drains to
. sp d M prevent, leakage near any one taak from affecting the others.
NK A similar philosophy will be used with respect to the air starting equipment. Air receivers will be arranged so that any one receiver is associated with only one diesel in each of the systems. Failure of any single accumulator system will
! -ther-fca not fail the emergency cooling system.
Another major part of the effluent control project involves the addition of a new dump tank and gravity drain lines. The f
a new covered dump tank will be vented back to Zone 1, that is, 5 will be provided with controlled effluent release. In addition, a gravity drain will be installed in the reactor building itself so that, even on failure of the lif t pumps, the building water level will not be so high as to damage the building. The dump i
tank will be filled with sufficient water to absorb the primary system's energy. The building drain to be installed will be provided with a 5 psi water seal.
{
Mr. Robinson pointed out that the gravity drain will lead to the disposal basin. There is a great deal of voice within the building and tank for settling and gaseous evolution of fission products before they reach the basin. The basin will be l l
designed to allow total ERW flow to percolate through the soil. )
l
2 DRAFT - 99th Meeting Part III, Page 12 A f acility will also be added to process rupture waste offluents. This will include an ion exchange system and a 300,000 gallon waste tank. The tank will be sized to handle one rupture in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. That is, 300,000 gallons are sufficient to handle one rupture, and 300,000 gallons will be processable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by the system. It was noted that it takes 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to reach power agate following a shutdown, which was the basis for sizing the waste handling facility. The fog spray system nozzles have been replaced with plug resistant nozzles, and the testability of the system will be improved, i The graphite cooling system will also be improved.
l A stack gas monitor is being added. This instrument will i have a higher range allowing measurements over the total range
?
l which might be experienced in an incident. Both total activity and iodine 131 measurements will be made.
Dr. Hanauer asked what was the basis used by DUN for deciding
, when redundancy of equipuent was appropriate.
Generally, equipment is made redundant if it is in an active system. In the case of passive systems, DUN has gone to the extent only of assuring their integrity under accident modes. Examples of the latter approach are the comon manifold on the front face of the reactor, the comon line from the low to the high lift pumps, etc. Where it is not possible to assure the integrity of these components, they are being replaced with piping for which the brittle fracture characteristics _w L be determined. The systems
DRAFT - 99th Meeting Part III, Page 13 tuve been analyzed to the extent that DUN feels their probability of failure is low enough to be an acceptable risk.
Dr. Zabel asked if there was additional information concerning the effect of coolant pipe breaks.
The evaluatiors so far have indicated that the system is capab's of handling the worst pipe break without a fission 6.5 product release, provided that reactor operation is en anticipated.
The accident has been analyzed conservatively, and, if operating
- variables were slightly diff erent than those expected, only a slight release would result.
Ccnfinement versus containment 8
Mr. Miller listed the advantages and disadvantages of i confinement and containment systems.
I 1 t '
t
- 1. A confinement vents steam to the atmosphere. DUN i considers this better than e. containment for N-Reactor since pressure tubes are involved rather than a large 4
esw g ru ( co o mV.d'gs4.o t m<qvet-pressure vessel y uh4rk Mir: ==p c esem1;atica Oc be
>>{.g"%e.
a"^ Otherwise, it is approximately the same as a
. contaireent with respect to the ab'.lity to analyze pressure l I I transients.
l
- 2. In both cases, one must ace'rately calculate blowdown l and fuel transients. DUN considers the problem definable.
- 3. In the confinement system, DUN must assure no fission product release before the pipe break.Yhat is, fuel damage is not pe ure surge fails the pipe.
9
i DRAFT - 99th Meeting i Part III, Page 14 I In this connection the dual scram systems, both of which are adequate,were noted.
- 4. Confinement closures were felt to be as reliable as the isolation valves on containments.
- 5. A confinement requires reasonable leak tightness. For fission product removal, plate cut, filtration, etc. are depended on. Containments depend on almost absolute leak tightness.
ytt"n
- 6. The N-reactor confinement permits an elevated, slow i
noble gas release. This is thought to be an advantage j since the noble gases and other fission products are not t
l kept under pressure. It is thought adequate for the N-
,! Reactor because of the isolated site.
I l 7. In a confinement concept, one must assure controlled release rates. Large metal-water reactions are not acceptable.
This problem is being studied. Among other things, metal-wder reaction rates with overheated irradiated fuel elements will be measured. The projected improvements will decrease the probability of a total loss of core cooling. This problem is also of concern in connection with containment systems.
=4- ^ The following conclusions have been drawn with respect to a confinement system: 1 l
- 1. The systems are probably not suitable for densely populated I sites because they allow the release of noble gas fission produc l A confinement system represents a tradeoff of noble gas release for reliance on which characterizes ,
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,l DRAFT - 99th Meeting Part III, Page 15 containment systems.
- 3. Both containment and cor.finement systems depend on emergency core cooling.
- 4. The confinement system at the N-Reactor on that site will represent an acceptable safeguard following implementation of the ef fluent control project.
Mr. PaEadino asked how DUN assured that the closures would function so as to allow pressure dissipation. Mr. Miller
- described the 15 vents available, 14 of which operate together.
, These remain open until primary system pressure reaches very I
low levels. The butterfly valves are normally open,and,even if closed, there would be some release depending on the rate I of steam release through the fuel discharge canal, which represents a water seal. In addition, one of the 15 vents is held open until the release to the atmosphere has allowed .
1 filters and the fog sprays to control the accident situation.
[ If a valve closes for some reason, the operators will be aware of it. The primary system is maintained in such condition g that releases can be accomodated with the remaining valves or power will be reduced. More than half of the available vents are required only on the largest double ended rupture. If all of the valves were closed and 2 rupture took place, the water in the seal 2ed pit would first be blown out. Den it was thought sM the rDmight yieldhr the door sh
. es -
DRAFT - 99th Meeting 'q Part III, Page 16 Unresolved Problems Mr. Nechodom listed the following as still requiring resolution. He noted first that evaluation of the design against the 70 General Design Criteria would probably incorporate all of these.
1.,h r _ d_rsign plant was designed t criteria g,---
which met pr e s er
- 1 y be met, but some checking is needed, owt
- 2. Tests will be carried ofa the diesel air solenoids and valves.
- 3. The design adequacy of the total system to resist pipe whipping must still be done for parts of the primary loop
} other than the reactor front face.
I 4 The fog spray system drop size and efficiency are not 6,s s s I eher following the nozzle change.
I
- 5. The new ERW system test procedures have not yet been formulated.
W ie w Mr. Nechodom felt that no further : r tim was required on analysis and testing of the seismic resistance of control systems, sophon supports, pipe penetraticus and battery racks.
G' '
r~ o e e
- In addition, no deiti^^e_t work is necessary on the single high lift suction pipe, the Zone 1 water accumulation jpower supply to the stack monitor, the ERW manifoldj or the low lift diesel fire Protection. DUN has ne additional work planned on fuel damage during blowdown or break size analyses. Site meteorology, plume
c _ ,
DRAFT - 99th Meeting 3 Part III, Page 17 p e.
contours, iodine removal are still important questions, but not as they were in the past.
The applicant was informed that the Comittee felt a letter
,. could be written in support of the effluent control project and recommending its implementation.
Apar t from that, Dr. Zabel added, it is felt that the program outlined including review against the 70 General Design Criteria, appears a reasonable approach. The Comittee
, also believed that a review against the recent IEEE Standard on nuclear plant instrumentation would be a useful and instructive t
exercise, i
i Additional review will be necessary after completion of to.\L
- all of the studies outlined,and the Comittee hd noTcoment on those aspects at this time.
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99th Meeting - DRAFT Section IV IV. Rancho Seco A. Executive Session Dr. Stratton reported for the Subcomittee. D e Sacramento Municipal Utility District (SMUD), the applicant, has no experience in the nuclear field. In addition, they have no experience in operating steam plants. Allgide"ration this utility so far A
has been by hydroelectric power. They have asserted that they are hiring and will train anployees, etc.
The plant site will be about 25 miles southeast of Sacramento.
i The surrounding environs are largely devoted to cattle raising i M g and have a low population density. The location is *4+e unique t
in that there is no water supply at the site. The applicant has t proposed using a canal with a capacity of about 100 times what I If et
{ will be needed for the plant. /the canal is constructed 3
in time, they will lay a pipe e r : 20 mile b e .ce from a nearby reservoir.
There will be a large storage pond located c1 the site which will
, contain enough water to cool the reactor for about 35 days.
The site is one of the least seismically active in California.
- et u>g t o ne history of earthquakes in the vicinity showsevents 1::: ther.
! ~3I.
intensity ')K on the Modified Mercalli scale, corresponding with about a 0.058 horizontal ground acceleration. If the San Francisco earthquake were arbitrarily rooved along the San Andreas Fault to the point on the fault cloaest to the site, then it would result 7 Ottr&
in an intensity hw*e Mcdif t:d Mesent t-4c414 at the site.
The applicant's se,isrnic consultiuit' Mal estimated that the maximuro potential earthquake would have a peak horizontal ground
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DRAFT - 99th Meeting Section IV, Page 2 acceleration of .05g. The Regulatory Staff and its consultants have asked for a design to 0.25g peak horizontal ground accelera-tion 4; and this 5.1:Niko
- p. 591yt been accepted by now.
D e applicant proposes concentrating liquid waste $and trans-porting them off site. There is some problem with tritium disposal and with its concentration in the reactor. The e(app'i r o ps~ s.l t i 92nte skoe Wh. L o
-to-heep recirculati,n,3 th(e demineralized water which might result in -tm44%p-to hazar dous levels of tritium. Continuous evaporation is one way to get around such a problem. .
i One point of difference between the applicant and the Regulatory Staff involves separation of control and safety instrumentation and I l the in-core power monitors. There is also a question of the adequacy of the liner anchorages. The analysis which has been performed I
apparently assumes that everything is perfectgb B&W is still studying the question of a failed fuel element detector. They hope to have a concrete proposal late this year.
Although there seemed some question concerning the possibility of flooding the reactor cavity, it now appears that this capability will exist.
I here was some discussion concerning the liner and anchorage Rig d4or$ tt 4(s design, but it was not clear what the nature of the question was.
u1 A h had b m . rai :d by *he-Reguhtery et f f, Dr. Hanauer ^^' ~ " that the fission product detector program which has been proposed by B&W will not lead to any positive hav e result. He felt they were proposing approaches which $44/fh been demonstrated to be unwork
~
? h 5$h DRAFT - 99th Meeting i Section IV, Page 3 Dr. Hanauer also had questions concerning the use of a 100 year wind as a facility design basis.
Dr. Bush pointed out to the Comittee that this was probably the first plant for which the added nuclear capability of the utility would represent more than 1007. of the existing generating capability. Dr. Hanauer noted that this has been discussed in Subconraittee, and an extensive intertie with the PG&E grid was described. Dr. Hendrie thought that, even so, with 757, of the installed capacity in one plant there would be strong pressure to keep it running.
, Dr. Hanauer thought that,since large plants were economic andsmallutilitiesexisted,therewouldalwaysbepools.a*4 hen
, a particular utility constructed a plant for the use of the pool, I
g it would be built to meet the pool's requirements rather than s
that of the individual utility.
There was some discussion about the proposed radwaste system. Dr. Isbin noted that this would be the first proposal which did not include any liquid waste effluent. He was not sure that all of the problems involved in recycling the entire f
primary system inventory had been recognized. Dr. Hendrie suggested that, by using resin beds, etc. , the primary system water could probably be cleaned sufficiently.
Dr. Hendrie asked if xenon oscillations were anticipated in this reactor. Dr. Okrent pointed out that a fully instrumented 1
core would be included which would allow their detection. Dr. l Stracon added that ppability to part-length rods was
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DRAFT - 99th Meeting Section IV, Page 4 included and that larger margins and lean boron would be involved than in Westinghouse designs, for exampla.
Several consultants were on hand. It was agreed that Mr."elmhkandMr.EplershouldhearB&W'spresentation concerning separation of control and safety instrumentation and then report on it to the Comittee. Dr. Paris was asked to present his report concerning end-of-life pressure vessel ,
failure (sce Section I.F.)
l B. Meeting with the Regulatory Staff I
g Following some discussion concerning thermal shock and j the possibility of pressure vessel failure at end of life 9
(see Section I.F. above), Dr. Siess asked what the Regulatory h Staf f's concern was about the containment liner. ': s .. :cd i -the+7his concern applies to all similar contairanent designs.
p Dr. Gluckmann noted that the problem is the use of angle fasteners.
Intermittent, alternating welds are used at the sides of the angles s.wd wh4 h are not thought to be adequate. The Turkey Point and Palisades designs, on the other hand, use studs as fasteners.
These are more flexible and can easily relieve differential I
stresses on buckling.
There was some discussion about the reinforcement design for the tendon anchors. It was not clear to the Staff exactly who would be doing this design, but DRL has sked to see the final
. design before the containment is built.
N DRAFT - 99th Meeting .
Section IV, Page 5 Dr. Hanauer stated that the basis for the design wind loading was not clear. He asked why a wind speed having a probability of once per hundred years had been chosen. Mr. Spickler replied that in areas where there are not severe hurricane and tornado problems the relatively flat wind speed probability curve is normal. In addition, even a 300 mile per hour wind loading would not be significant in view of the seismic load requirements.
Dr. Hanauer asked what criteria had been applied to the design of the refueling water storage tank. Mr. Levine replied that DRL j has taken the position that it is not necessary to design for 1
- simultaneous tornado and accident loadings if it can be shown that a tornado will not cause an accident. The problem of a tornado hitting the containment while the containment is pressurized has also been neglected, which means that it is not necessary to worry about tornadoes during the short term post-accident cooling period.
Follwing an accident, structures which protect the containment must have been designed to continue operation. Mr. Spickler added I that the criterion used by DRL is,1f tornado probability is less than once in 4,000 years, it is not necessary to include it in the design basis. The tornado probability for the Rancho Seco site is once in 23,000 years.
Dr. Hanauer still expressed some concern over hhe neglecting incidents having probabilities of one per cent over a 40 year life.
Mr. Levine noted that it had been suggested that, since an earthquake
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W=_
g divW DRAFT - 99th Meeting Section IV, Page 6 close to a site could affect the entire plant, one should be greatly concerned about this event. With respect to tornadoes, their probability of occurrence is based on occurrence within a large area. If the area of a standard tornado is considered, then f actors of 1,000 less than once in 4,000 years are obtained.
Mr. Levine suggested that there was no need for concern in this case.
u; nd Still in connection with the design wi:F the;d--ign 'fith l 4h4 loading, it was noted that the borated water storage tank f, may contain as much as 60,000 curies of tritium and will be located outside of containment. Should it rupture, however, dispersion in 100 mile per hour winds is likely to be quite good.
i o C. Sacramento Municipal Utility District L
! Dif ferences Between Rancho Seco and Other Recent B&W Designs Mr. Mattimoe noted that almost all of the design parameters for the nuclear steam supply system at Rancho Seco will be the 1 i
i same as those proposed for the Crystal River plant. The instrumenta-tion design will be different and will be discussed 4etes by B6W.
Organization .
!* S 12. m de Ennepeet Municipal Utility District will manage design and 1
construction. Bechtel will be responsible for supervision of the I
derig 2nd u!!1 ~'a ge construction. SMUD will invite bids on construction and will assure that the design and construction are I in accordance with applicable building codes and state and federal evovilt regulations. Babcock and Wilcox will supply the nuclear steam supply system.
- .A AiE _...r n L.} .' '
DRAFT - 99th Meeting Section IV, Page 7 Radwaste System Mr. Gideon described the radwaste system.prepc M . After letdown from the nuclear steam supply system, primary coolant gases are vented to the waste gas processing system. Liquids are sent through an ion exchanger with a minimum efficiency of 99.997.. Clean water from the ion exchanger goes to receiver tanks for holdup and then through a second ion exchange column as a backup. The second resin bed will again have a minimum ef ficiency of 99.997.. Liquids emerging from the second resin bed are sent to a boric acid concentrator which is essentially an evaporator. The concentrated boric acid resulting is sent to storage. Thedistilledwategand"volatile"boricacidcoming
} off th tvaporator are sent through a boric acid ion exchanger which produces clean j distilled water which is sent to storage.
The resins will be 407. Codion resin and will not be nitrated.
Mr. Palladino asked what disposal would be made of the fission products in the ion exchanger! Mr. Gideon noted ese :
~
are kept on the resins.and en they have built up sufficiently l 1
to cause the beginnings of a drop-off in reviw efficiency the beds can he taken of f the line. The resin beds can be flushed I
to the spent resin disposal tank and the wastes from the spent resin disposal tank packaged and taken to a disposal site. The system will be designed for 17. nominal failed fuel elements, although j it is realized that there is a great deal of disagreement over what that means in terms of activity release.
1
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DRAFT - 99th Meeting Section IV, Page 8 Mr. Gideon noted that at the San Onofre plant an ion exchange ( clean up system is used which can discharge directly t o th e o c ean pi t h = t ="-- *
- r ' * - - ' '"' ", = 5 = i, .c id cera =: J ia. In contrast to San Onofre, the distilled water produced at the Rancho Seco plant by the cleanup system will be recirculated which might result in seconcentration of tritium.
There was some discussion of the change in efficiency with useoftheionexchange\resinbeds. two system volumes have passed through ion exchangers in the cleanup system, the effluent J,\\ %. 4 concentration g h about 10 of the influent concentration (forboricacidat about1,000 ppm)andcontinuestorisefrom I then on. Contaminants such as cesium and boric acid can be j removed quite well. It is only anticipated that 14 to 16 system volumes per year will be put through the radwaste system.
n ere will be a separate "dirty" liquid waste system designed nL 1,G.h to collect evaporate and collect residues in the bottoms which A A,
, will be drumed and shipped. h e distillate will be treated i
similarly to the primary coolant wastes, but with different ion b41s.
This will ed avoid highly contaminated ion exchange usa 44 '
exchange resins in the primary coolant chanup system.
Dr. Isbin asked if it was realistic to depend on a single evaporator for cleanup of all of the dirty wastes. He asked 1
what evidence there was in support of this approach. Mr. Wang j 1
noted, in answer to Dr. Isbin, the only difference between the proposed design for Rancho Seco and the Turkey Point radwaste l system was thatpatpTyrkey Poin dated boric acid and l
n l
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r
, 1 DRAFT - 99th Meeting i Section IV, Page 9 distilled water could be discharged to the ocean which was not possible at Rancho Seco. Only conventional equipment will be used. The point was repeated that equipnent essentially the same as that at San Onofre and Turkey Point will be used about R f15CC the same dege+e and at a relatively insignificant increase in cost. One of :he Regulatory Staf f representatives recalled that at San Onofre the decision was made on the basis t;iat the cost to recover boric acid and distilled water was higher than the worth of these items. It is not thought that any R6D program is necessary to establish the claims for this system. Mr. Gideon
, in provided references which supported /his opinion, the system design, t ch4aeq 6 He get:(4 WCAP 7613,for the original work on the operation of an j
6
+hg ion exchanger in the presence of boric acid and E controlling 1
chemicals. He also quoted work done at the Hanford laboratories i
using aninonium ion as the geces ca(ton which also validated the analyses performed for Rancho Seco.
Dr. Isbin noted that the system description in the PSAR is quite similar to that in the Oyster Creek 2 PSARjwith the exception of tritium control. He did not believe it was the system which 8
re had been described +b the Coernittee. Dr. Zabel asked if the system
, described by Mr. Gideon was documented. Mr. Gideon replied that it wasexceptforcertainnumbers(suchastheefficiencyfiguresgiven above)which were points on the efficiency curve representative of fresh resin.
{?&_ _[
DRAFT - 99th Meeting Se crion IV, Page 10 In answer to Mr. Palladino, it was noted that there is not information available on each isotope which will pass through the resin beds, but that cesium 137 has been used as representative of all since it was considered the most difficult cation to remove.
Mr. Palladino asked how the high tritium levels would be approached during fueling operations. Calculations have been performed based on the 40th fuel &eg cycle, which assumes a l primary coolant volume of 299,000 gallons. About 78 microcuries per milliliter of tritium is present in the primary coolant. This j is also based on the assumption that there is no loss of tritium I
from the system throughout the life of the facility. Various l
, combinations of accidents and refueling problems have been investigated, and even for 1007. diffusion of tritium through the fuel cladding 1/3,000th of the MPC in air results for unrestricted areas. In addition, other ways of getting tritium out of the system l l
not connected with refueling were considered. An example was a '
25 gallon per day coolant loss3 sed failure of the letdown coolers, etc. The results in each case were less than MPC in air for 1
unrestricted areas. A plot of the tritium concentration in the !
I t containment atmosphere as a function of the degree of diffusion l
1 through the cladding was displayed with the number of operating cycles plotted as a parameter. During refueling it was assumed that the air velocity over the refueling pool would be 5 feet per second which tended to increase the evaporation and tritium concentration in the conta a er e. The curves indicated )
)
+-_ u l
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DRAFT - 99th Meeting Section IV, Page 11 that below 17. diffusion through the cladding activation tritium controls. Above 17.,the tritium ciffusing through the cladding is the major contributor.
Dr. Isbin suggested that the time to be concerned is when the tritium concentration in the water reaches the point where there would be MPC levels in a: c:F= cf air in equilibrium with the water. He suggested that that tritium concentration would be less than many which appear for the various cases in the PSAR.
i Dr. Isbin asked what that air concentration would be. Dr.
Goldman replied that, at 72 F and 50% relative humidity, water containing 2 microcuries per milliliter of tritium would result
-6 in about 5 X 10 microcuries per cc in the air in contact with
?
. it. This is not significantly lower than the numbert indicated by the analyses in the PSAR. l Tritium concentration in the wate ill be measured with a liquid scintillation counter. These measurements will probably i be made during each refueling , and accuracy better than 10%
I is anticipated.
l I (Dr. Isbin noted following the meeting that his calculations I
t j indicated that the above numbers were off by an order of magnitude.
Water containing 0.2 microcuries per milliliter of tritium in equilibrium with air will produ i X 10'0 microcuries per i
milliliter in the atmosphere, i.e., the MPC in air for tritium.) )
1 s
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DRAFT - 99th Meeting Section IV, Page 12 It was noted that storage capacity is available if it becomes necessary to store water containing tritium. The demineralized water storage tank will be available for this purpose.
. Dr. Hanauer asked what the effect would be of a spill of the refueling water storage tank containing about 70,000 curies of tritium.
The tank contains 300,000 gallons and might drain off site although the applicant was not clear on this point. There are aquifers at about 140 feet below ground and wells which draw se on themy but they are under pressure. It is not thought that j any spillage could get below the pressure retaining' cap".
t
- At this point Mr. Gideon wished to add something to his l statement concerning the radwaste system design. If there are
? l tsotopes for which decontamination f actors across the resin beds j
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L4L6 l are not available, they occur in such small quantities that, if A 1 1
they were to pass through the ion exchangers, they would not i approach Part 20 limits.
, Instrumentation and Control 1
Ihe applicant referred to the answer to question.7. A.5 in Amendment 2. The presentation was aimed at discussing how separation of control and safety instrumentation will be achieved for Rancho Seco with emphasis en the differences from previous B&W designs. Intermixing of safety and control vill occur in reactor coolant pump controls, in reactor coolant pressure measure-ments and in reactor coolant flow eme nt s . These instrument
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DRAFT - 99th Meeting Section IV, Page 13 systems will be identical with those proposed for the Oconee, nree Mile Island, and Crystal River plants. In all of the above, separate control channels are used.In at least two of
- mg the above systems, any channel esa be used for control)but only one at a time. Ocm ;t :f Eur f th: 4annale - : 9 -d are and Ivu svu n v'r. In addition, the control system output is isolated from the protection part of the channel with an isolation amplifier.
l e
In the reactor power level measurement system proposed for Rancho Seco, a different approach will be used. B&W feels l
that safety will best be served by a control system design I I
which minimizes shutdowns due to single control channel failures.
l 1
l An averaging and auctior.eering type of system has therefore been '
designed. As before, isolation amplifiers will be used to isolate signals. The system will meet IEEE criteria, AEC criteria and the requirements of the United States position j taken in July 1963 21:F ::r.pI:1 te Section 5.4.1.1 of the laternational Electrotechnical Connission jDocument 45-A. N e eye * = ui11 ':: :: the AEC's g nars1 heetgn f*v 4 t ev 4 m 4
{, Brieflyjthe output of a linear amplifier fed by a neutron detector goes to ten individual isolation amplifiers in each power range channel. Two will be discussed below. h e other eight are used to provide signals to the computer, to control room lights, and alarms, etc. One is used to provide signals to the protection system bistable and_one goes teameraging acplifier. Each
--E
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DRAFT - 99th Meeting Section IV, Page 14 averaging amplifier averages two of the four power range signals. An auctioneering unit will select the highest of the two averaged signals and use this to provide signals
,- for the control system.
B&d feels that this averaging technique provides more reliability and better safety characteristics. Downscale failure of 2 and in some cases 3j channels are required before j
a rod withdrawal will take place. Failure mode analysis indicates that no short circuit, open circuit, ground fault or combination of the above between the isolation amplifier
? and the control system will affect operation of the protection i
I system. In addition, cross connecting the averaging amplifiers'
- inputs will not affect the protection system and connecting the I d j av erage, out put to the input of another averaging unit will not affect the protection system.
- p = * ~df=::ien 111 af f ui she prote m c. sy;; &.
- The only failure in the control system which has any potential of being reflected back into the protection system is direct connec-i tion vf a power source across one of the inputs in the averaging j system. An analysis has been performed which assumes that I
L' (.s amplifier failure results in a short circuit
- hat ie, across the o.h k i amplifiep{_1 : m that voltages meet have been applied within j l
the system cabinets which are not located near voltage l sources. I Voltages of 400 volts DC,or peak AC j across an isolation amplifier output (the input of the averaging amp 1 f 1
the protection system by actual test 4
1.
,y-DRAFT - 99th Heeting Section IV, Page 15 5,000 volts across the output of one averaging amplifier will not affect normal operation of the protection system.
Application of 6,000 volta across the output of the
, auctioneering unit will not affect the capabilities of any of the four protection system bistables.
Voltages above 400 volts applied across an isolation amplifier output will cause the associated protection channel to trip or fail. Up to 10,000 volts,only one of the four channels will be so affectedy C 11,700 volts applied at the same point will fail a second t
- channel feeding the same averaging amplifier, but will still i
- leave the two remaining channels to protect the plant.
, 16,800 volts applied at the same point will trip or fail i
g all four protection system channels. This seems difficult to achieve within an instrumentation system cabinet.
5,500 volts at the output of an averaging unit will trip 1
or fail the two associated channels; 7,500 volts at that point t {
will fail or trip all four protection system channels. It was i
noted that 150 watts of power at 7,500 volts are necessary to
{
bring about this result.
7,000 volts applied at the output of the auctioneering unit will cause trip or failure of all four protection system channels, but at least 140 watts of power are needed at this voltage.
In terms of tests, B&W has applied 1,300 volts at the output l l
of one isolation amplifier without_seeing any measurable effects !
=E _i c _ ;17 :. ^.
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99th Meeting Section IV, Page 16 in the input to the bistables of the other three channels.
Tests and analyses have indicated to B&W that the proposed system cannot be failed by the highest DC voltage source in the plant when applied within the instrument system cabinet. The I highest voltage in the instrumentation system (1200 volts DC) can fail only one channel, and this effect is doubtful since the source is current-limited. Voltages required to fail more than one channel are not available in such a way as to make failures credible. The averaging-auctioneering system is being used because it is superior to a single-measurement control system. In addition, it mer.ts all single failure and separation i )
} criteria referenced by the AEC. Its ability to perform under 4
3 fault conditions has been tested at voltages up to and exceeding I
those available.
Dr. Hanauer raised the question of protecting against systematic of failures,such as design maintenance errors.
o The applicant replied that their only exposure to this problem j so far had been by way of a coment in the Comnittee's letter on the Salem reactor. They asked if some clarification as to the 1
- intent of that statement was possible.
! Dr. Hanauer observed that the vendor had clearly gone to considerable lengths to protect against single failures but o r-that,for faults in sensors ef *he linear amplifiers jreliance was
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99th Meeting '
Section IV, Page 17 6
place / on redundancy in the system to make these tolerable.
Dr. Hanauer felt that in the system as designed the only component that keeps a dead chamber, for example, from simultaneously causing an excursion and ruining part of the protection system was the auctioneer. This component is,therefore jnecessary to keep those things from happening. Dr. Hanauer felt that it should be part of the protection system which makes the protection system work rather than part of the control system. He felt that as long as
. consideration was restricted to various single failures the systan i
i If some systematic difficulty would probably be satisfactory.
should crop up, however, then not only is protection degraded, it bec'mes necessary. There was additional discussion of what constitutes random yconcurrent f ailures and the point was made
{
that there is a history of such concurrent f ailures which were not found and fixed. Dr. Hanauer concluded by noting that his private opinion was that control and safety systems should not be connected. As compared to the Crystal River design, this proposal seems to be in the wrong directiog in his opinion.
In connection with concurrent systematic errors, there was some question about the type of cooling to be provided for the l i
y instrumentation.
The instrumentation will be designed to 110 control room ambient temperature or 150 in the cabinet and will be tested under these conditions. ~ z ;
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Mr. Wascher observed at, in his opinion, MW had not come l to the Committee with less separation. At the time the previous
%W designs were reviewed, the data and analytical work presented 1 1
1
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DRAFT - 99th Meeting Section IV, Page 18 above was not available. Also, with respect to features in the safety system which allow the ir.terconnection, in Mr. Wascher's opinion, the existence of four channels was such a feature. Even assuming a failure in one channel, there would still be an acceptable level of protection.
With respect to nonraadom failures, b. Wascher agreed that he could see Dr. Hanauer's concern. The basis of the present design is to establish a control system which is reliable i enough to work. In Mr. Wascher's opinion, in order to get around the concern over concurrent failures, it is necessary to establish
, the design conditions adequately.
?
j Dr. O' Kelly raised the question of staffing. Mr. Stinson,
, the plant superintendent, replied that SMUD's goal is to bring 6
key people in early. In fact, offers are presently out to people i
with conriderable operating experience. Mr. Stinson felt there l
was some advantage to not having existing steam plants in that tW no established union was involved allowing hiring from outside j
the SMUD organization. Were has been a target date established for full staffing.
1 i
Dr. Okrent asked if seismic loadings on equipment would be investigated.
Such an investigation will be performed. The Regulatory Staff will supply limits on acceleration and tilt which ace acceptable. A MW analysis is now being awaited. The analyses have not yet been completed, but they will be done and testing will be done when nec .
r discussion of the 1_._ ,_ _ . -
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DRAFT - 99th Meeting -
Section IV, Page 19 anticipated maximum earthquake. Dr. Byerly took exception to the high accelerations which have been established for this plant.
a In connection with the statement that there were no plans for installing a strong motion accelerometer, Dr. Okrent questioned the approach that would be taken at some future date feHr In;; e
~_
- h. wv ing w relatively large earthquake) hen information might be necessary on the accelerations experienced. Mr. Mattimoe stated that there was no objection to installing an accelerometer at the site.
Dr. Okrent asked what fluence had been predicted at the t pressure vessel wall, i
Over the 40 year life,3 X 10 is the predicted value, LO
[i including uncertainty, n e expected value is 1.9 X 10 19 .
I p A few- group nS Codewasusedtocalculatethisjlfourgroups were used}. 'Ihis was corapared against three sets of experiments and compare ( well if a scaling factor is included.
nere was some question conerning the number of surveillance specimens planned for the Rancho Seco plant. Mr. Wascher pointed out that preliminary analyses had led to the selection of eight l h ut.
specimen holders. Rese would3bec'hpaced so as to correspond with 4,3,* 1n 90 = A
'O yc m-ef- epe r ienp * *k *= rpren % l weeld i m ;;:nt 1, 9,15, 30, 60 and 90 irradiation years on the 0
v A 2, yfo,ur vessel .
g specimen holders are now proposed which will correspond with 10, 20, 30 and 45 years of vessel irradiation and therefore is
DRAFT - 99th Heeting . "
Section IV, Page 20 sufficient te tch: *b e er e through vessel life. The two spares, the early samples, and the twice-lifetime, technology specimen have been eliminated.
Dr.Hanauerquestionedthewinfondingdesignbasisefa j 100 mph wind.
It was noted that 100 mph is the once per-hundred year wind.
Full tornadoes occur once every 22,000 years. There is nothing on the site which is not governed by the seismic load considera-tions even if extreme high winds were evaluated.
Dr. Hanauer observed that all of the existing SMUD generation is remote from the Rancho Seco site and no SMUD power lines will
}
i connect Rancho Seco with the rest of SMUD's system, It was agreed that Rancho Seco will represent more than 507.
t 2 of SMUD's total installed capacity. They are, however, completely l integrated into the pool, and spinning reserve is also shared by the pool.
Dr. Hanauer comented the.t this is the first company the Connittee has come across for which thefe-nuclear capability will represent so large a fraction of + hair installed capacity.
In addition, there will not be a dispatcher working for SMUD who will be able to provide power to the nuclear plant over company owned lines.
SMUD will be able to control the systan power levels and will have control over safety-type situations. It was felt there was not significar t tween an SMUD and a PG6E
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DRAFT - 99th Meeting I Section IV, Page 21 dispatcher. Dr. Hanauer thought that the matter would be given additional attention at the operating license stage. Dr. Hendrie added that the Comittee will need substantially better assurance thac the multiple ownership and dispatching scheme will work.
1
- k ; t n i a. h: beer d m net ::cd oi it.; g:: -* **~e_
Dr. Zabel asked what plans or experimental program was being considered to handle the thermal shock problem. Mr. Kreps replied that B&W has considered three different analytical methods, and all have agreed to within 57.. Thermal shock is not now seen as a problem, and MW has no additional plans i !
in this area.
1 Dr. Okrent asked if the analytical results referred to would be affected if the flux level calculations were off by a factor of f two or three. In addition, he pointed out that the characteristics I
of the steel may be untert ain and asked,if in the final analysis the entire situation is uncertain, what would be done. Mr. Kreps replied that, if uncertainties become troublesome, then considera-I tion may be given to annealing the vessel in place, although this is a costly and awkward procedure. He thought maintaining the
- vessel at 600 to 800 for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would suffice but l stressed that this would be done only as a last resort. Analytical I
work and comunications with the Regulatory Staff are continuing. l l
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- ) 8 u DRAFT - 99th Meeting [I l Section IV, Page 22 4
The applicant was told that the comittee felt a letter 1
could be written. The letter might contain mention of some l of the items discussed at this meeting. l i
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__...2 DRAF'a' - 99th Meeti C *\
Part V V. Maine Yankee A. Subcommittee Report Dr. Hendrie briefly described the site. n e plant will be located on a tidal river on the coast of Maine. The site itself is large with a 2000 foot exclusion radius. The population density is relatively low, h ere are about 3500 residents within five miles, including 400 to 500 summer visitors.(based on 1960 census data). There had been some seismicity problems which have since been resolved. The major structures will be founded on bedrock.
The storm-connected water level problems have also been resolved.
i nere are no site dif ficulties remaining.
I n e plant itself will be a PVR designed by Combustion Engineering.
It will be a three-loop plant with a power density similar to that f of Fort Calhoun and a power level similar to those of Surry and Palisades. We core cooling systems will be katandard)Nincluding J
high and low pressure pumps and accumulators. There will be contain-ment spray systems with an iodine removal additive. The containment
', itself will be a reinforced concrete structure. I l
The plant design has been controlled by the Yankee engineering j l
1 group who were responsible for the Yankee Rowe and Connecticut Yankee plo '
I
- and the Vermont Yankee reactor,now under construction. Stone and Webster will be the architect-engineer and constructEr of the plant.
l Combustion Engineering is responsible for the nuclear steam supply system, and Westinghouse will supply the turbines.
. .. nJFC --
o DRAFT - 99th Meeting -
Part V, Page 2 There is only onc item lef t unresolved so far. nie involves off-sitepowersupply,sadThefroposeddesigndoesnotmeettise keneral kesign briterion No. 39 in that there will be only a single
, overhead line supplying the two reserve transformers.
Mr. Mangelsdorf pointed out that the applicantAM indicated there are unsatisfactory aspects N the "easy fixes" which have been suggested. In addition, the applicant contends that their proposal does meet the criterton which requires redundancy of active components only. In addition, they have pointed out that the industry has j in generalj been critical of the criterion in the first j place.
(
l Dr. Hanauer noted that the drawings which have been submitted L
do not show separate breakers for the reserve transformers,and these may also not be redundant.
s Dr. Hendrie continued with the Subcomittee report. He thought that the Comittee might want to include some cautionary statement in its letter concerning the containment design. ne containment i
will be reinforced and not prestressed, and it is proposed that shear loads will be carried by a combination of rebar devel action and 1
aggregate interlock along cracks in the concrete resulting from the initial pressure tests. Dr. Hendrie thought that the design had not even been preliminarily completed in the areas of discontinuities.
Dr. Hendrie pointed outj as possible additional discussion points, the thermal shock question on which CE has recently submitted a report and xenon oscillations which CE now says are likely in the axial direction. '
,#j m . .-m..< <
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DRAFT - 99th Meeting Part V, Page 3 Dr. Isbin noted that the design includes many more control n
rods that other similarly-sized PWR's. He asked if this did not accentuate flux peaking. Dr. Hendrie agreed that it would be in cle smd b 3
that direction but the hot channel f actors wh4eh the applicant 6
ci ' L are better than most.
Dr. Okrent asked if a rod ejection accident had been considered.
Dr. Hendrie noted that a rod ejection accident resulting in an ' energy insertion of 246 calories per gram had been considered. Dr. Hanauer added that the rod clusters are no more reactive than are those in r designs including f ewer control rods. :
. 1 f With respect to in-core instrumentation, Dr. Hendrie observed 1
that the plant will include permanently installed in-core monitors f
at the insistence of the Yankee engineering group. The operating organization believes this will allow more efficient operation by providing better flux distribution information.
Dr. Hanauer questioned the rod scram bus power supply design.
tkwt i He noted6 questions concerning testing and use of a single AC bus I
were raised by the drawings presented. Dr. Hendrie observed that the
, scram bus power supply had not yet been designed and that the applicant !
i has agreed to criteria concerning its design which are acceptable to I
the Regulatory Staf f.
The Committee's discussion closed with a question by Dr. Okrent concerning the documented statersents on criteria for emergency core cooling system ( design. It was agreed that this question would be
. -y -
pursued with the applicant
DRAFT - 99th Heeting Part V, Page 4 B. Meeting with the Regulatory Staff Mr. Boyd noted that the Staff's final positions have been detailed in its supplementary report to the Committee and that
' the final information from the applicant is contained in the recently submitted Amendment 12. The only outstanding problen area concerns off-site power supply, and the Maine Yankee group wants to present its story to the ACRS. The Regulatory Staff feels that the proposed design does not meet the intent of the f criterion as interpreted by DRL. Mr. Moore added that it is not t, really possible to make any distinction between active and passive
} components in connection with electrical equipment. Later in the n
- meeting this point was raised again by Mr. Mangelsdorf. He asked n
how much, in the opinion of DRL, the public safety would be improved L by paralleling 300 fcet of now-single line. Mr. Moore observed that, if the Committee does not agree with the DRL recommendation concerning off-site power supply, that some consideration should be given to i
changing Criterion 39. There would be considerable difficulty in applying it if the Maine Yankee application was approved as proposed.
Mr. Moore did feel that meeting Criterion 39 overall would improve the public health and safety.
Dr. Hanauer asked if in the staff's view the situation should be improved to the point where one could feel more confident about the lines the proximity of the reserve transformers, etc. Mr.
Schroeder replied that that was the DRL position although he would defer to those who hd reviewed th ton detail. He felt 7
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DRAFT - 99th Meeting
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Part V, Page 5 he could back up Mr. Moore's comments but observed that it was difficult to quantify exactly what was being achieved in terms of public health and safety.
. Dr. Isbin observed that the discussion on rad waste treat-ment was different from the usual. He noted that the PSAR distinguishes between hydrogenated wastesand those not hydrogenated.
Mr. Clark pointed out that in some areas a hydrogen overpressure is used to shift the equilibrium oxygen concentration resulting from raalolosis. He thought that a better terminology might be hydrogen containing'.'The practice of segregating hydrogea-containing i from nonbydrogen-containing radioactive waste reflects the Yankee
?
engineering group's experience with their other plants. Dr. Isbin also suggested that there was almost no information concerning which g nuclides would be involved. Mr. Waterfield thought, on the other hand, that information was given in similar detail to that in other applications.
Dr. O' Kelly raised the question of using polyvinyl chloride in the concrete to provide runs for the instrument cables. He asked if there was suf ficient history with the use of PVC in concrete to make this a good procedure. Mr. Clark thought it was the standard way of treating cable runs,and Dr. Hanauer noted that one et4y depends on the PVC to be smooth to allow the cables to be pulled through at the beginning of life only.
Dr. Siess pointed out that reference is made to carrying earth-quake shears in the reinforced concrete containment by something other ti.'n tension in the reinf -
ast time such an approach
E DRAFT - 99th Meeting Part V, Page 6 was used was in connection with the H. B. Robinson containment, and this is a worse situation since no prestressing is involved and cracks will occur both horizontally and vertically under the pre-operational test load. Dr. Siess was willing to presume that the approach was not inappropriate in this case since the earthquake loading would be small but suggested that this might open the door toothersituationswhereitmightnotbesomna11,enednadequate analysis should have been presented. He thought that what hafeen presented was not adequate. No calculations have been made of bending stresses in the dowel bars ,and the reference to a test of l bearing stress was not applicablej according to Dr. Siess. Dr. Siess
, also pointed out that there is no design information available o
concerning the region near the bottom of the cylinder and suggested that a more sophisticated calculation of the interaction between the rebar and the liner was needed.
, , Mr. DeYoung agreed that it was not quite clear where one i would draw the line in connection with seismic loadings. It was, however, not clear that the Staff could balk in this particular i
I case. It was suggested that the Regulatory Staf f wo/1d probably j i O.Io L . t.. \
cut off this approach at about W M +h g y y ; %' the seismic loads appropriate to the Maine Yankee site, j Dr. Okrent notedj in connection with the refueling accident 1
analysis,that DRI,had asked Maine Yankee to provide a fix. He I asked why these suba.nnblies acted differently from those of other 4
_ _ _ - _ .-- yh __
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DRAFT - 99th Meeting Part V. Page 7
- reactor vendors. Mr. Boyd cornented that it was not clear that these would in f act act differently but the applicant had put a particular analys!s on the record and had taken the position that
(
there would be significantly more rods f ailed than has been the case with other vendors.
Dr. Okrent suggested that it is problematic to ask one group to do something and not raise similar questions with the others. He then asked if there was something significant to report d
t in connection with the asterisked items.
Mr. Clark replied that an outline had been submitted of what I
Combustien Engineering will report to the Subcoernittee during the j forthcoming visit to their Windsor facilities. They are aware of I
g what the Co rnittee's interests are and a great deal of their prograa
)
e is still in the planning stages, although they were responstre to many of the Comittee's suggestions in connection with the Fort Calhoun review, the N 4xperiments on flow blockage effects, etc. will be carried out.
Dr. Okrent asked what integrated flux was estimated at the t
reactor vessefall jin the Maine Yankee case. Mr. Clark noted that the point had not yet been reviewed. Dr. Okrent observed further that in the DRL public safety evaluation on the Point Beach 2 plant it was noted that the applicant would be asked to reduce thermal stresses. He asked if the approach was the same in connection with {
this case. Mr. DeYoung replied that it was not .
DRAFT - 99th Meeting Part V. Page 8 Dr. Bush recalled a statement made by Dr. McKee to the t
effect that Monsweas Bay was a nonflushing body of water and, in addition, Dr. McKee had recalled that there was a fair amount wot of lobster fishingg wJ to the position taken in the PSAR.
He asked what the actual situation was. Mr. Clark noted that consultant's report indicates that there would be considerable flushing in the bay.
I With respect co f ast flux estimates at the vessel vall, i
a A Dr. Okrent asked if it would be practical for DRL to
! prepare several sample calculations for the vendor's use resembling t
i typical,large WR's and also resembling the Yankee situation on which measurements have been made. In Dr. Okrant's opinion, this l might anow a better feel for the validity of the various calculations.
Dr. Okrent noted also Mi: : enc"h the need to account for the effects of core heterogeneities on the flux and for differences between the flux level at the location of the surveillance sample and that at the wall. He thought BNL might be asked to work on this area. Dr. Morris agreed that DRL would consider these suggestions.
It was noted in anr9er to a question by Dr.11anauer, that there is no more information than was provided in Amendment 12 concerning the control rod scram bus design. Dr. Hanauer observed that the a,tt\.wab Staff did, however, sign off in its report to the Cornittee; &as. le, there is at present apparently no good way of testing the scram bus y amid a single bus will be used, etc. Mr. Ippolito replied that, since this is CE angement, a design approach
DRAFT - 99th Meeting Part V, Page 9 had not yet been worked out. He agreed that less was known about this design than other recent systems. With respect to
, separation of control and safety, the CE design approach is to separate char.c.els fully.
C. Maine Yankee Organization Quality Control -
Mr. Minnick comented that the Yankee organization recognizes l its responsibility for all aspects of design, construction and i
operatica of the Maine Yankee plant. The Yankee engineering group's l quality contrc,1 program will include the engineering design fequip-I ment -and specifications, vendor selection, component testing, quality i
h control and inspection during erection, systems checkout,and startup g testing. Yankee representatives will participate in each of these areas, and none can be completed without the group's approval.
Wo quality control representatives are already at the site.
Their responsibility includes seeing that work is carried out in accordance with established procedures,and they can stop work if required. About half of the final station complement will be on l I
, site later this year. All equipment will be checked in accordance with written procedures before its acceptance, and all work performed by any organization will be checked by at least one other group. l As an example, Mr. Minnick observed that Stone and Webster 4 11 install the concrete throughout the plant. Itse-U. S. Testing oratory will check the mixes and the final concrete quality, !
and the Maine Yankee site representative will also perform spot
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DRAFT - 99th Meeting i ,
Part V, Page 10 checks in this area.
Vessel design specifications will be provided by one Combustion Engineering division
- a second Conbustion Engineering
, group will perform the vessel design and stress analysis. Southwest Research Institute will check the vessel stress report.
All inspection sheets filled out during destructive and nondestructive tests will be kept on file for the life of the facility.
- Dr. Bush asked if more detail would be provided with respect to the in-service inspection program. Mr. Minnick replied that
{ &as44 ally the necessity for such a program has been recognized, i s
- 3 and'u h
Aprovisions will be made for thd.u,are reasonable. The manager
- of operations for the Yankee engineering group is also Chairman of the USASI Comittee which is developing a code on in-service inspection se , c_..
- gw =uy solis w e l l ...i c v i -th!" _;; i- this cer .c: 12 ._ l Dr. Bush asked if it is the groupl s intent to start out with l 1 l a base line inspection as outlined in the draft standard. Mr. Minnick ,
replied that it was and that they intended to follow as closely as I possible the present draf t of the code document. Base line inspec-tior6v111 be performed and will be followed by periodic vessel inspections. In this connection the primary system and critical i
auxiliaries were also included.
l
DRAFT - 99th Meeting Part V, Page 11 Maine Yankee Design Characteristics Mt. Tribble provided the Comittee members with a brief
.. comparison of several of the thermal-hydraulic characteristics
! for the Maine Yankee plant as well as for the recently reviewed Surry Station. He noted that iny general,the Maine Yankee design is the mre conservative.
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Dr. Isbin comented that one could also interpret the
)
y situation to mean that the Maine Yankee reactor core is less i
[ efficient and therefore less economica1 as j presently conceived.
)* O.c.56u.it,it M
3 is the intent to increase the efficiency of the Maine ;
4 ;
j Yankee core as time goes on. Mr. Tribble replied that fuel ]
- i costs were less based on CE's bid than on the Westinghouse bid which was received at the same time. While there egy be upgrading
$ \
of the core in future, it would he only after a great deal of l
l experience has been gained in operation. Dr. Hanauer asked what !
the ultimate turbine capability was and was told it is about 107. )
i q higher than the rating. That is, there is not an unusually high 1 1
stretch capability in this case. 7 Dr
. Okrent observed that the criterion by which the performance of emergency core cooling I A systems will be gaged is not quite clear. He recalled that, in connection with the Surry Station the applicant had stated that ,
using conservative assumptions and allowing appropriately for
- fuel element distortion from the original core geometry, the emergency core cooling systems will be designed to keep fuel clad temperatures below the point at which the el ate on subsequent
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l DRAFT - 99th Meeting l Part V, Page 12 i
cooling. He asked if this statement applied to the Maine Yankee plant as well. Mr. Tribble answered that,while he was not sure !
that the statement was made so explicitly in the PSAR, it is in fact the intention of the Yankee engineering group and does represenfabasicECCSdesigncriterion.
Dr. Okrent then raised the question of predicted fast flux at the vessel wall versus experimental results. He noted l
that the Yankee engineering group has had some experience with checking predictions against experimental results in connection
, with the Yankee core and suggested that CE be given the appropriate j information and that a calculation be done using the CE methods a
which could then be checked against the experimen8. Dr. Okrent i
{ personally expressed some doubt about the accuracy of the calcula-i i
tional techniques. Mr. Minnick thought this was a point well taken and would in f act be done. Later in the meeting Dr. Ckrent also M
suggested a similar check calculation be performed on moderator coefficients fin this case using information from the Connecticut Yankee reactor.
9 Thermal Shock 3 Mr. Gibbons gave a brief history of the CE effort in the area of thermal shock calculations. In April,1967 CE was requested to study the effects of emergency core cooling on the Maine Yankee vessel. In August, 1967 a report was issued which showed that s.\ 0t temperatures ,inside surface were well above the NDT transition
[h temperature and no brittle fracture was possible. kracture mechanics
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DRAFT - 99th Meeting Part V, Page 13 approach used in the same report indicated that ;Se maximum D .10 flaw size for which no growth could be expected was about 7-40ttnr O.CO o& an inch. Flaws down to about 2 #1^i c' ^" inch are detectable
, so that this was not of concern either. Following this, in connection with the Fort Calhoun project, the ACRS and the Regulatory Staff raised additional questions. These were answered in Amendments 5, 8 and 11 of the Fort Calhoun application. Among other things, these new analyses included the effects of axial temperature differene on the vessel. The CE approach has not raaterially changed from that
- time in connection with calculating residual stresses, pressure and seismic loadings, stresses and deformations on core internals, f atigue damage, the ef f ects of axial and circumferential cracks, and l the toughness properties of grained-coarsened A-533,which is thought t
f to be similar to the heat affected zone of welds in the vessel. An additional thermal shock study was submitted as part of h Maine ,
Yankee Amendment 5. This confirmed the conclusioni, of the earlier l
CE reports, but included a sensitivity analysis as well. A meeting 1 I
was held with the Regulatory Staff in May to review this report. j i 1 At that time a major point of contention was the value of the film heat transfer coefficient which was appropriate. CE presented w.n quench dat h support
- of its approhich
)ach Jds*taken from one of the HSST program plates. This plate was 10 feet by 10 feet by 12 inches thick and was quenched from 1600 F. The data taken were cornpared with plots of the CE analytical approach in the 600*F t
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o 200 F range. The Regulatory Staff remained unconvinced, however, l l
'and an analysis was performed which showed that even with twice the 1
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DRAFT - 99th Meeting Part V, Page 14 heat transfer coefficient calculated, one would not get cracking more than 627. of the way through the wall. The studies are continuing.
- Mr. Gibbons pointed out the conservatisms which have been inherent in CE's approach, particularly with respect to the Maine Yankee vessel
The water in the accumulators will be at about 1U50 F whereas 70 F was used in the analyses.
No credit is taken for heating the incoming water when i, it contacts the safety injection piping, the coolant 1
i pipes, the nozzles, the vessel wall and thermal shield I
or for mixing with hot water already present in the vessel.
i Thermal stresses were calculated for an 11 inch section, however, the vessel in the core region will only be 8 @
inches thick. The neutron flux used, H.% was representative of the beltline, i
In obtaining the stress intensity factor at the crack tip, CE's approach assumes a fixed opening load. Thermal stresses, l
however, are actually self. relieving.
It is anticipated that the fracture toughness values will be higher at the temperatures of interest than the values used in the analyses, t
The Irwin model was used in making the calculations of stress Tui intensity f actor.wMeh is actua11 re conservative than
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DRAFT . 99th Meeting Part V, Page 15 other models since it does not permit rotation of the crack center line. That is, no strees relief is assumed.
C dditicc.,lhe peak stresses calculated were used. The peak stresses, however, actually occur in the stainless steel cladding which remains ductile.
The cladding thermal conductivity was also not accounted, for in the calculations.
Mr. Gibbons felt that,1f one actually took account of the cladding effects in the calculations, one could probably y change the film coef ficient by about 1000 without affecting i
- the results. CE still concludes there is no way for the energency I
core cooling system to jeopardize the vessel integrity.
l Dr. Paris noted that the point of the discussion was actually to establish the validity and conservatisms inherent in the analysis.
He questioned the material properties values used. He pointed out there is only very limited data available with respect to-fracture t
toughness and that in fact data is only available up to about 200 F.
I This has been extrapolated to 550 F. The analysis also takes
$ credit for a temperature difference through the wall and a lower i integrated neutron flux through the vessel wall d59:5Y which lead to co puting numbers which are not necessarily conservative in those areas. With respect to the analysis, that is, the assumption that
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stresses are, relieved, Dr. Paris pointed out that the method does
% A 41 in fact allow for stress rwin =~1y in presumes that
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DRAFT - 99th Meeting
- Part V, Page 16 the wall edges are not displaced laterally and agemruy introducing l a crack will cause the edges to bend inward and induce higher stresses at the crack tip. In short, Dr. Paris had some question about how accurate the analysis really is when one is finished assessing all of these factors.
Mr. Gibbons replied that, with respect to the conservatisms !
which he had outlined, he had tried to point out real physical The situations as opposed to, analytical approach. As an example, the 11 inch section for which the stresses were evaluated is well removed from the core,and the simultaneous use of beltline flux levels is
- a real conservatism. In his opinion, this takes care of uncertainties e in the incident flux. As further examples, the cladding is in fact
- there, the water is heated as it enters the vessel, the material i
} properties have been predicted on the basis of overpredictions of incident flux, the stresses are overpredicted. While there is h . t.k I some controversy over the, transfer coefficient, in CE's opinion l even the heat transfer coefficient was overconservative.
l Mr. Gibbons noted that CE is working toward a finite element analysis which will allow an elastic-plastic analysis to be done.
They do not want to attempt this, however, before the heat transfer l model has been resolved since it would be so much wasted effort. l As far as the uncertainties in fracture toughness and the lack of material data, Mr. Gibbons pointed out that CE had gone to
, Dr. Irwin since they wanted the best advice available,N h"tAdu.t.f
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reported by CE represented Dr. Irwin's advite.
Mr. Palladino a fracture toughness j
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_-w e e 7 _7 DRAFT - 99th Meeting Part V, Page 17 would be required before CE might predict extensive cracking of the vessel.,
It was noted that the additional dosage which could be with-
- stood was evaluated. Assuming the correct heat transfer model and taking advantage of improved temperature properties as one moves through the wall,as well as the reduction in dosage across the wall thickness, then one could reduce the fracture toughness at the midpoint from 112 to between 98 and 95 and still limit crack
- penetration to 507. of the wall thickness.
Dr. Hanauer observed that it seems as though additional work 4
! is necessary due to uncertainties in the material properties and t
i the analytic procedures. He asked what the applicant's approach would be if in the final analysis it develops that there is a serious f
problem. Mr. Minnick thought that one could heat the water in the accumulators and benefit thereby. Af ter some discussion, however, it developed that, while the dif f erence in temperature between 70
, and 170 has been evaluated with respect to heat transfer at the i e vessel wall, its effects on core cooling have not been valuated.
There was an impression on the part of the applicant that the sensible heat is much less importan Ntent heat in this respect, however. As far as the accumulators are concerned, these are high pressure tanks anyway and would, therefore, suppress boiling although problems which might arise in their design have a&oe not been looked at in detail.
Dr. Hanauer asked if the reactor cavity would flood should the vessel fail. Mr. Tribbi ti de cavity will fill af ter a
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DRAFT - 99th Meeting Part V, Page 18 vessel break. Water level would be about 5 feet above the core bottom, and steam cooling would probably suffice to protect the core.
Dr. Okrent asked if the CE representatives could repeat i their statements concerning tolerable flaw sizes. Mr.Tuppby
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replied that CE had first concluded that a .7 inch flaw was j tolerable during a 100 per hour cool down at end of life. Analyses have since been run using the Irwin model in which internal pressure was maintained at 2200 psi and cooldown was at 100 per hour. In l this situation a one inch flaw could be tolerated.
i Dr. Okrent asked what size flaw could be tolerated at end
! of vessel life at operating pressure without any cooldown. Mr.
t e @ o I a
Tupp&ny replied that,with the vessel at 550 F under full pressure, b
a flav 3.8 inches thit.k could be tolerated if the fracture toughness is about 140. It w"s noted that this value has been extrapolated l
because of a lack of high temperature data. l Arrangement of Off-Site and Emergency Buses Mr. Beckley pointed out that Maine Yankee will be part of an integrated plan designed to pring additional power to the New f England area. There will be two 11$ KV lines supplying the plant, one connected with Mason Station and the other with the 9ow M Station. On site, the lines will be brought in through motor operator disconnects and oil circuit breakers to a single bus. The bus will extend 300 feet to the two reserve transformers, each equipped with t
a motor operated disconnect.
l DRAFT - 99th Meeting Part V, Page 19 Mr. Beckley noted that the normal station supply will come from the 345 KV line through two tranformers.
The followirg additional arrangements have been considered by the applicant.
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DRAFT - 99th Meeting Part V, Page 20 With respect to B above, Mr. Beckley noted it would have been less costly but more subject to interruption. Arrangement C T
was again les expensive but much more subject to inkrruptions A
.- since the transformers can be used with only one line each.
Arrangement D was more costly than the others and is not thought to be worth the expense. Dr. Hanauer asked if Mr. Beckley would outline Maine Yankee's criterion for an outside system used to supply engineered safeguards. Mr. Beckley could not specify a i probability for outages, etc.
. Dr. Hanauer observed that the two 4160 volt buses would be i located in the same room.
f He asked if a fire would therefore not I completely take out the engineered safeguards. Mr. Beckley replied i that fire protection will be provided. The switch gear will be metal-i g
i enclosed with little combustible material included and are 300 feet from the diesels.
l Dr. Hanauer suhgested that, if the applicant feels that criterion l 39 is inappropriate,they are in a sense obligated to propose a different criterion. Mr. Tribble commented that normally the Yankee engineering
, group considers engineered safeguards to be operable without of f-site power,and he pointed out that the Atomic Industrial Forum and the Yankee group had made the same comment with respect to Criterion 39.
Dr. Hanauer stated that it was his belief that a 115 KV transmission grid was more reliable than diesel generators. He asked if it was f air to state that the applicant did not require off-site power to the engineered safeguards to be immune to single failures. Mr. Beck . sas not the case. The
-:e DRAFT - 99th Meeting Page 21, Part V imunity to single f ailures, however, is extended only to active components,and as examples he cited the motor operated disconnects and oil circuit breakers.
.- Mr. Palladino thought it was not clear why the applicant should balk at doubling up on the last 300 feet of line when two separate line4s ha'e been brought into the plant from the transmission grid.
Mr. Minr.ick observed that he did not really think that two sets of ,
wires t.trung side by side in the same yard really improved the situat. ion.
Containment Design i
! Mr. Ireland described the proposed containment. It will be
}
< a reinforced concrete structure with a steel liner which functions j as a vapor' membrane. The containment will be designed t' i resist accidents, earthquakes, and wind loadings. The containment g
reinforcement will be primarily No. 18S bars, and there will be negligible stress on the reinforcement under operating conditions.
The bars will be designed to resist all forces without counting on the liner. Stresees have been calculated to be below 907. of
, yield in every case. Cadweld splices are used primarily, although
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- some butt velds will be used hreirq;-
3 the same strength requirements.
Close to the base mat, there will be additional vertical bars ins talled to accomodate the momemts induced by the fixed base.
l Radial-diagonal stirrups will be used to resist shear forces at i the base. Mr. Ireland felt that there would be a pattern of horizontal and vertical cracks opened during pressure testing.
Earthquake induced. shear L1 resisted by combined
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DRAFT - 99th Meeting Part V, Page 22 dowel and aggregate interlock action.
Mr. Ireland pointed out that Holley had performed an analysis I of the resistance to shear with cracked concrete, etc. He had also performed tests of the dowel action and showed that the analytical results were reasonable. Amendment 10 had included additional subsequent testing which showed that the dowels alone could with-stand the earthquake forces without the aggregate interlock mechanism being taken into account.
Small penetrations which fit within the rebar pattern will
'. not require additional reinforcement. Penetrations too big to I .
I fit within the pattern but not bigger than about 3.5 feet will I
require additional bars. The equipment and personnel hatches j will be fitted with reinforcement rings to which the bars will be t
i anchored.
Dr. Siess questioned the values for bending stresses under the assumed rebar distributions. His computations had indicated
- these would be about 16000 pai and therefore not really negligible.
>cim %s Dr. Siess continued .c(o.d yvinced. out that Dr. Holley's report 1
. ; had aid the c.;;r2ch "2 only he4*g applied to the shell where there are no radial shear or bending moments. That is, it did not 1
apply to the bottom 20 feet. He asked what would be the approach )
l for that part of the containment. I l
The tensile stresses will be much lower at the bottom i h
part of the containment that is the case higher up, and no concrete j cracking is expect concrete will also be l
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DRAFT - 99th Meeting Part V, Page 23 reinforced by additional vertical bars. If there is cracking, the additional bars will carry the loads. Horizontal bars would carry the loads if there were vertical cracks.
Dr. Siess observed that, in the upper region where the concrete will be cracked, the analysis does not assume that the liner participates ;
in carrying shear loadings. The liner will, however, be deformed under such conditions. He asked if an analysis had been performed l to show that the liner will accept these deformations without failure.
?
Mr. Ireland replied that the calculations indicate this is the case.
, The liner connections, S/8th inch studs spaced 24 inches apart, and I '
t 12 inches apart in the lower portion, will also accept these l
deformations.
.I There was some question ev%er the documentation provided on i this point. Mr. Beckley thought that the analysis of liner deforma-tion had been documented. The analysis indicating that the icwer part of the containment would not be cracked under the pressure loadf ng had not been documented, but he indicated that this could be done.
Under questioning by Dr. Siess, it developed that the prediction of loads to be carried on the bars in actual use from the tests which were made had not taken into account the peak loading value based on a dowel action analysis, but had only assumed t N arbitrary stress distribution. Dr. Siess suggested the point would be worth I
checking.
DRAFT - 99th Meeting Part V, Page 24 Dr. Holley commented that, when the report which he had prepared was submitted, the containment base detail was as yet undeveloped. That area was onitted from consideration therefore.
He had not meant to imply that the same approach would not be applicable in the base. In addition, at that time the detailed spacing of the bars was not known so that two vertical bars per foot was used in his analysis. Actually there will always be at o
least 2 1/2 vertical bars per foot. In addition, tvo high a shear
, value was used. It has since been reduced from 37 to 22 d os
, per foot. Because of these things, much lower stresses will be j
, developed than the 10 to 17,000 psi Dr. Siess had suggested.
t l In answer to a question by Mr. Ern ington, Mr. Minnick pointed out that a visual inspection of the containment while under the test pressure loading is included in the acceptance test specifications for Maine Yankee.
Dr. Hanauer asked how far down the Connecticut Yankee vessel had cracked under it's pressure test.
The vessel had cracked to within about 25 feet of the base mat. The Maine Yankee crack pattern is expected to be approximately the same, altbough a higher internal pressure will be used. There will be additional steel reinforcement to compensate for this, however.
Dr. Siess asked why a different approach had been taken in
, designing the Surry containment. In that case diagonal bars had been used.
Mr. Ireland recalled thrt, at the time the Surry containment
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W' d - 1 DRAFT - 99th Meeting Part V, Page 25 was being designed, questions had come up in connection with the Connecticut Yankee design, in which earthquake shear loadings were carried on the liner. De Staf f's consultants at that time had indicated that they would not like to see another such design ;
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with a liner of comparable thickness. For that reason, diagonal bars to carry shear loadings were included in the Surry design. )
At the present time a better analytical approach has been developed so that diagonal bars were not thought to be necessary for the
- Maine Yankee containment. Mr. Ireland thought that the Surry
} containment would be designed,1f it were being done at present,
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at Surry was .15g. He felt that it would probably be reasonable to go to earthquake accelerations as high as 0.3g without including e
diagonal bars to carry shear loadings.
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Dr. Holley agreed that one could probably carry much higher l 1
shears that apply to Sutry without diagonal bars.
I
- Dr. Isbin asked if the radwaste treatment was typically better i
in the Maine Yankee design than in the operating plants under the purvue of the Yankee engineering group. Mr. Minnick did not think j it would be appreciably better than in Connecticut Yankee, for example, and is in fact essentially the same. Boric acid recovery is employed,and the Yankee engineering group continues to feel that this is a good approach. nere will be minimal waste discharge.
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DRAFT - 99th Meeting Part V, Page 26 Condensate, however, is not re-used in the Maine Yankee design or at Connecticut Yankee.t The plants can operate well within any limits established, and in fact Yankee-Rowe has been operated within Part 20 by f actors of 1,000 with no problems. The problem with re-using condensate is that there is always a residual amount of boron in the condensate, and one does not really want to recycle this. Mr. Vandenburgh thought that if one wanted to re-use the condensate it would have to be put through additional demineralizers, and more equipment probably would be necessary. It is thought that demineralized water can be made for less cost. The amount of water to be treated depends on the number of shutdowns per
? year.
t Dr. Hanauer asked if there was any additional information available concerning the design of the scram bus. Mr. Beckley replied that Maine Yankee had agreed with the Regulatory Staff to modify the scram bus arrangement so that it will comprise two independent systems. The ability to test the breakers will be included as well as the ability to separate the rod drives into 1
two groups with respect to their power supply. There is an agreement to submit the design to the Regulatory Staff before ima fabrication,and all instrumentation will be designed to IEEE standard No. 279.
The design will include complete separation of control and
- protection instrumentation. In addition, the drives to be used will have been proven through experience or have had sufficient testing and operati cornon failure modes
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DRAFT - 99th Meeting Part V, Page 27 will have been identified. They will be thoroughly tested prior to startup. During the first 100 to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> while equipment is energized, it is thought that any component failures,due to 3*
manufacturing defects, etc. should be revealed. In addition, thesystemisdesignedfailsafek ,inatripmode)inso far as is possible.
Dr. Hanauer suggested it was too optimistic to assume that common f ailure modes will have been identified. He suggested that i
the subject needed additional thought.
]
Following a brief caucus, the applicant was informed that the Comittee believes it can write a letter favorable to construc-a i
tion of the Maine Yankee plant. It is likely to mention a number of the subjects which had been discussed. With respect to off-site power, the Comittee was still deliberating and had not established its position.
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CONTINUATION OF THE 99TH ACRS MEETING JULY 21, 1968 WASHINGTON, D. C.
Attendees:
C. W. Zabel, Chairman S. H. Hanauer, Vice Chairman S. H. Bush H. Etherington F. A. Gif ford i J. M. Hendrie H. S. Isbin H. G. Mangelsdorf H. O. Monsen A. A. O ' Kelly l
D. Okrent ,
N. J. Palladino l C. P. Siess W. R. Stratton H. E. Plaine, Legal Advisor
, R. F. Fraley, Executive Secretary
. J. E. Hard, Senior Staff Assistent l J. C. McKinley, Staf f Assistant L. Blische, Administrative Of ficer
, Reviewers: W. L. Faith D. Okrent !
Letters: Zion Station ,
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TABLE OF CONTENTS l
i Page I. EXECUTIVE SESSION General Letter on Reactor Vessels............................ 1 II. Z IO N S TAT I0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 III. AP POI h'IMENT O F D R . L . SQUI RES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 IV. MISCELLANEOUS A. Future Mee t ing Da t e s . . . . . . . . . . . . . . . . . . .................. 8 B. R ema rk s b y D r . Z a b e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 t
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t Continuation of the 99th ACRS Meeting Minutes I. EXECITTIVE SESSION Dr. Monson said that he had attempted to rewrite his draft letter to include Dr. Okrent's ideas without reference specifically to the Zion site, and he suggested six steps that should be taken to resolve the pressure vessel failure problem. He said he felt the only one that would be controversial is the requirement that the designer examine the size of failure against'which protection can be provided. He said that he did not anticipate reqairing backfitting as a result of this study. He went on to state that if the Comittee could accept a letter in this vein there may be no additional remarks on the Zion letter. This was confirmed by Dr. Okrent.
Dr. Hanauer asked if the Comittee would have any problem of prejudging sites on which it had not seen any data or reactor design. He said he did not think the Comittee had determined how much protection is enough.
Dr. Monson distributed draft No. 2 of his proposed letter. Dr. Zabel suggested that the Committee read this letter and determine what action to take.
Dr. Hanauer said that he had reviewed the drafts that were included in the members folders as well as Dr. Monson's draft, and each gives him a problem since he feels that all of these letters prejudge future sites.
( Mr. Etherington said that he was concerned by a letter that would say "Do what
}
t you can, and we won't protect against anything else",
Dr. Okrent advised the Comittee that he had made revisions to his additional i
remarks that he plans to attach to the Zion letter.
t Dr. Isbin said that he felt the Comittee had taken a step forward with the
- l statetpent that the additional remarks would be dropped on the Zion letter if l the general reactor vessel letter could be agreed upon. 1 Dr. O' Kelly moved that the Comittee censider all of the general letters in !
order and consider Mr. Etherington's concern that the Coenittee would be saying
' for the designers to determine what accidents they could protect against and
- accept no protection against others. He suggested that the Comittee read each letter and then have a general discussion. Dr. Gifford seconded the motion and, when it was called to a vote it carried. l 1
Mr. Etherington observed that all of the proposed letters contain three pages of blah and he suggested that the authors identify the pertinent points.
Dr. Zabel suggested starting with the first letter in the members' folders l
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Continuation of the 99th ACRS Meeting Minutes Page 2 which was the one under tab 1 prepared by Mr. Fraley. Mr. Fraley summarized the content of the letter saying that reactor plants must be designed to cope with reactor vessel failures.
Dr. Isbin asked what the letter was to tell Dr. Seaborg what does the Connittee want done and why was the letter written. Mr. Fraley said that he thought the letter would say, that for sites worse than Zion, these are the things that must be considered, these are tha things that the AEC Staf f should look at and that the industry should be aware of.
Dr. Hanauer said that he thought the publication of such a letter would be an invitation to the Nuclear Industry to move into sites that were worse than the Zion site. He objects to safety by patchwork (add something for each new site).
Mr. Mangelsdorf asked, if this letter were adopted, would it imply that if these things were met that the Burlington site would be acceptable? Dr. Bush said he felt that the industry does not know what would be a starting point for utlilizing locations in core populated areas.
g Dr. O' Kelly noted that the Cocnittee will be looking at sites worse than Zion, e
' and he thought the Committee could put together a letter that would give positive guidance to the industry for sites that are worse than Zion. Dr.
Monson expressed his own feelings that the Coenittee was asking for too much 1
.' for sites that are only slightly worse than Zion site. Dr. Zabel said that he I i was not sure that the Coenittee consensus is that for Zion or better sites that
! the designs presently proposed are acceptable. He said there is some indication that various members may want something more at sites that are as good as those l presently in use and certainly for those that are worse.
Dr. Bush read his draf t letter No. 2. Dr. Isbin asked what Dr. Bush personally l believed. Dr. Bush said that he was not able to put together a conceivable set l of circumstances that would lead to a catastrophic reactor vessel failure. He
, said that he believes in the leak before break concept, that he can't believe in a break in the belt region until the reactor receives very large fluents. He is reserving judgment pending the results of the Heavy Section Steel Test program.
He said he did not anticipate any problems for at least the next ten years. He is not concerned with neutron exposures in the range of 1019 fluents. He said that he wrote a letter on Metropolitan Siting, that he does not really believe in.
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__ ,.. W . W _ _. r-Continuation of the 99th ACRS Meeting Minutes Page 3 Mr. Mangelsdorf noted that the provision for containment cooling and cavity flooding to protect against a major core displacement due to a pressure vessel rupture do not appear to be compatible with the requiremont to provide measures to guarantee against major release of fission products to the atmos-phere in the absence of ineffective core cooling. Dr. Bush agreed that these
.. two could not occur at the same time, and therefore, were somewhat incompatible.
Dr. Monson pointed out, however, that the China syndrome must be met.
Mr. Etherington read his draft letter and stated that he was not sure about his real position. He was concerned about making GE do major redesign on their pressure vessels and in forcing the PWR's to being made larger. He pointed out that when he worked for A. O. Smith, that Company had made a number of tests back in 1928 - 1930 on chromium venadium vessels which failed in a brittle manner. He pointed out, however, that these tests were conducted to prove that the selds were better than the base material.
I Dr. Bush stated that--the weld material is generally better from a fracture point of view than the base material. He said the only reason for insisting that B&W make tests on the electro-slag weld techique is to give some data and j confirmation that the weld is as good as it should be. He said that he does j '
not expect any adverse or negative results. He pointed out that by increasing
- the PWR vessel diameter pushes the wall thickness up to about 13 inches, which is beyond the current technology (which is presently limited to about 12 inches).
4 He noted also that it pushes the wall thickness beyond the present HSST Program.
Dr. Siess pointed out that Consolidated Edison Nuclear Units 4 & 5 plan to reduce the off-site dose only for those accidents presently being considered and that thsy might be adversely affected by more severe reactor vessel failures.
Mr. Etherington said that he felt that if the Comittee was going to force the applicants into considering failures greater than the double ended pipe break, then the GE containment concepts and the ice condenser are going to be found i unsatisfactory.
Dr. Bush pointed out that the best estimate of the cost to hydrostatically I
test the Yankee Nuclear Power Station is about 1/2 million dollars per test. He pointed out, of course, that this was a function of the bookkeeping technique and the downtime involved. He said the Committee must determine what is accomplished by the expenditure of a large amount of money.
Dr. Monson went back then to discuss his July 20 draft. Dr. Hanauer asked him }
if this was to be a Metropolitan Siting letter. Dr. Monson said no that it was I not.
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.; g y Continuation of the 99th ACRS Meeting Minutes ;
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It was to be a letter relating to sites worse than those presently approved.
He did not feel that it took reactors into Metropolitan areas. He pointed out again that he felt the last item would be the most controversial.
Dr. Okrent stated that he believes that these steps should be started at Zion, but he does not object to waiting for the next round of reactor sites.
He stated that if Dr. Monson's letter would be adopted, he would drop the additional remarks that he plans to attach to the Zion letter.
Mr. Palladino said he liked the feature of Dr. Monson's letter that does not tell the designers what to do. Dr. Isbin pointed out an inconsistency between items 5 and 6 wherein 5, the reliability must be proved while in 6, 1007. reliability is not required. Dr. Hanauer asked if the designers had met all of these things would they have received app oval for the Burlington site.
Dr. Bush said that he thought this letter appeared to be more of an engineered safeguards letter than a reactor pressure vessel letter. He asked the
,' I Conraittee what kind of a letter it would propose to write. Dr. Okrent then i discussed the two draft letters that were contained under tab 4 of the folder for this meeting.
Dr. O' Kelly suggested that the Committee change its objective to that of
- writing a reactor safeguards letter such as Dr. Monson has proposed rather than t limiting it to reactor pressure vessels. Dr. Siess seconded this motion and pointed out that Dr. Monson's letter does not make reference to the 1965 pressure vessel letter.
Mr. Etherington pointed out that for the last two Comittee meetings, the members have talked about reactor vessels and a general reactor vessel letter.
He said he thought the Comittee might be mired down if it attempted to go into a general safeguards letter.
Dr. Hanauer said he felt the proposed letter was an invitation to the utilities 9 to push already large and unproven extrapolations of reactor design into Metropolitan areas.
He said he objected to offering a blueprint for the next step. He doesn't believe that technology is as far advanced as the reactor designers claimed.
He opposes the motion.
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Continuation of the 99th ACRS Meeting Minutes Page 5-Dr. Siess said that he favored the motion in that it could be modified to be a warning to applicants that sites only slightly worse than Indian Point and Zion will not be accepted without additional safeguards.
Dr. Isbin suggested that there might be a possibility of handling this in another manner such as a Subcommittee action.
Mr. Etherington said that he thought that he would like to have an opportunity to sleep on any letter of a broad new concept. He said that he regrets his part in the hasty 1965 pressure vessel letter.
Dr. Monson said that he thought the problem was with population density and not with poor quality of reactor vessels.
Mr. Etherington pointed out that the authors of the various draft letters were asked to draft reactor vessel letters and not reactor safeguards letters and that reactor vessels were all that he considered.
a Dr. Zabel called for a vote on the motion - seven members favored it, five
( opposed it.
t He ruled that the motion carried and that the Committee consider i Dr.
or diverging. draf t No. 2 for the next half hour and see if opinion is converging Monson's t
Dr. Monson then read draft No. 2 completely through. Dr. Zabel then asked for a 1
discussion on the first paragraph.
Dr. Hanauer said that the Cormittee needed to get down to the safety aspects and the improvements that should be made in safety sssurance and the safety related items.
Dr. Siess suggested that some place in the letter the Committee indicates that t
these are minimum requirements. '
Dr. Hanauer said that he felt the plants should be designed to minimize the l consequences of small accidents as well.
{* Hr. Etherington compared this to agreements made by the late John L. Lewis when he said for the United Mine Workers "We want more". He said he felt the Committee should specify what they want in the way of something more.
Dr. Hanauer pointed out that the public could not tolerate a radioactive waste accident and that the use of Part 100 for small accidents should not be permitted.
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f Continuation of the 99th ACRS Meeting Minutes Page 6 Dr. O' Kelly said he felt the Comittee should identify the responsibility of organizations for the R&D Program. Mr. Etherington said that when the Committee reviews the application for a provisional operating license, it will be determined if the design R&D, etc., is adequate to protect the health and safety of the public. Dr. Isbin pointed out that a number of reactor vendors think the R&D is complete and only the degree of conservatism needs to be
, verified.
Dr. Isbin said that he did not want to ask for more margin in the design when he did not know what the present margin is. Dr. Bush agreed with Dr. Isbin saying that he felt that the Comittee was continuously asking for more and more . Dr. Monson agreed that the letter could be more specific about where !
margins should be approved.
Mr. Palladino said he felt the letter should include a paragraph with regard g to the inspectability of the reactor vessel. Mr. Etherington said that he
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thought possibly the Committee should look at the seventy design criteria and i
see where they were inadequate. Dr. Bush said that possibly in such letters this could be interpreted as telling industry that they could ignore the seventy design criteria. He said he liked the concept of the realistic approach, ,l fj ( but he felt that industry may be treated like a ping pong ball between initially ;
' coming in with realistic assumptions being challenged on these and being forced l to go back to more conservative assumptions than having the Committee say "Why l didn't you look at realistic accidents"? I j Dr. Siess asked if it was the Committee's intent to say that the things that l were suggested in the November 1965 letter are now required items. Dr. Monson and Mr. Palladino agreed this was the intent.
Dr. Ha.nauer said that he objected to the invitation to industry to come in at sites that are worse than the Zion site. He said that he would have to dissent to this letter due to its general tone.
Dr. Okrent answered that the Comittee must address itself to sites worse than Indian Point and Zion. He pointed out that if the Comittee does not of fer guidance that reactors with the present safeguards will be proposed for sites worse than Indian Point and Zion or the siting of reactors in metropolitan areas will be taken over by some other group or individual.
0 Dr. Siess said he thought this letter was intended as a warning, that Zion type reactors and sites worse than Zion will not be acceptable without additional safeguards.
Dr. Okrent noted the difference between his position and that of Dr. Hanauer, where Dr. Hanauer says that no worse sites can be considered while his position f is that worse sites may be acceptable if these specific problems or areas
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Dr. Hanauer asked if the Comittee thought that Part 100 was appropriate to the Zion sites where the population cannot be evacuated. Dr. Hendrie said that he thought Dr. Hanauer's point was valid since he felt tue reactors needed substantial improvement if the low population zone cannot
.' be evacuated. Dr. Gif ford said the Comittee was getting into a much broadet question than reactor vessels and dafeguards. Dr. Monson asked if Dr. Hanauer believed that the last paragraph could be modified and other changes made to the letter to make it acceptable to him. Dr. Hanauer replied that it was possible but he was not sure.
Dr. Monson then observed that if there is going to be a dissent on this letter then the Comittee will not get it out at this meeting. If so, then he suggested that the Comittee drop consideration of the general letter at this time and proceed with the Zion letter and permit Dr. Okrent to simply add t',.emarks to that letter, f Dr. Zabel said that he felt there was suf ficient latitude to reach agreement (
g on a general letter, so he suggested a redraft. He suggested that the 4 ( Comittee break for lunch, then Dr. Monson should prepare a redraf t and per- l
' mit the Committee Members to read the remarks prepared by Drs. Hanauer and '
Okrent.
l Dr. Stratton said that he objected to issuing a letter under this extreme pressure. He asked if Dr. Okrent is considering withdrawing his additional J
coments on the Zion letter if he feels that the Comittee will get out a general letter, he said that another application is coming in to which .
additional remarks could be attached. Dr. Okrent replied that he had elini-nated his references to worse sites and reduced the length of his general l remarks; however, he still plans to make the additional remarks.
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? Dr. Handrie said that he felt that the Committee should proceed with the Zion letter after lunch.
Mr. Palladino suggested having the general letter redraf ted and attempt to reach agreement and see if Dr. Okrent would then withdraw h!s additional I remarks on Zion.
Dr. Okrent said that he felt that if the Committee accepted Zion with the proposed letter, he could not debate the problems on Indian Point 3.
Dr. Zabel then directed that the general letter be redraf ted as quickly as possible and that the Comittee take up the Zion letter and the additional remarks next. Subsequent to this discussion, Dr. Monson redrafted the general letter; however, it was not discussed further at this meeting. Fur-ther consideration was postponed until the 100th ACRS meeting.
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1 Continuation of the 99th ACRS Meeting Minutes Page 8 II. ZION STATION Dr. Isbin read draft No. 7 of the Zion letter. Dr. Okrent read his latest revision of his additional remarks. After a number of wording changes and rearrangement of paragraphs, Mr. Palladino asked if there were going to be any additional remarks by members other than Dr. Okrent. Mr.
Mangelsdorf replied that he had no additional public remarks at this time.
He thinks that this Zion letter is the best approach and suggests spending no more tiee on the general letter. Dr. Stratton seconded Mr. Mangelsdorf's motion that no additional time be spent on the general letter at this meeting.
The motion carried without opposition.
Dr. Isbin read draft No.8 of the Zion letter. Dr, Monson suggested a final sentence to this letter saying, "The matters discussed by Dr. Okrent were considered by the Comittee during its meetings. The Comittee believes tha t the status of these matters is satisfactory a as they pertain to the Zion units". ,
With the addition of this phrade, it was/ greed to show the letter to Mr.
Schroeder, and the letter was declared final, t
III. AP"0INTMENT OF DR. L. SOUIRES
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} Dr. Zabel and Mr. Mangelsdorf reported that the duPont Company has reported I
that Mr. Squires has persuaded himse? f that he could join the ACRS. This represents a change in both Mr. Squires and his management's stand on this matter. The Comittee agreed that Dr. Squires would be valuable addition to j the Committee. ;
E IV. MISCELLANEOUS A. Future Meeting Dates i
Dr. Zabel announced that there would be a three-day meeting in August and that the dates for future Comittee meetings were reviewed for October-November-December. These were:
ACRS 100th/Mee ting August 8 - 10, 1968
! 101st ACRS Meeting September 5 - 7, 1968
' 102nd ACRS Meeting October 3 - 5, 1968 1 103rd ACRS Meeting October 31 - November 2, 1968 104th ACRS Meeting l December 5 - 7, 1968 l
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M Continuation of the 99th ACRS Meeting Minutes Page'9 The following schedule was established subject to confirmation at the 100th ACRS Meeting.
105th ACRS Meeting January 9 - 11, 1969 106th ACRS Meeting February 6 a 8, 1969
- 107th ACRS Meeting March 6 - 8, 1969 108th ACRS Meeting April 10 - 12, 1969 109th ACRS Meeting May 8 - 10, 1969 110th ACRS Meeting June 5 - 7, 1969 B. Remarks by Dr. Zabel Dr. Zabel stated that he had been reviewing the hearings that had set up the various boards (1961), and he noted that in these hearings there have g been many very interesting coments and he encouraged all members of the i Comittee to read these hearings. He also noted that Comissioner Ramey had made a speech in the 90th Congress and that this was pertinent to the problems l the Coccittee was f acing. He went on to state that there seems to be a strong move afoot to relieve the Comittee of some of its power. He said that the u ( divergencies within the Comittee may destroy it, and he predicted that Dr.
Hanauer would be the last Chairman of the Comittee. He then chided the f Comittee on its performance on the Zion application.
i f Dr. Okrent expressed his opinion that if it fails to act the Comittee will not be abolished but will have certain of its important fonctions given to
, others such as the Bolsa Island Blue Ribbon Panel and/or a new Metropolitan Siting Group. Dr. Zabel said that he felt the additional remarks made by Dr.
Okrent should hot have been permitted to be inserted into the Zion letter.
Dr. Okrent responded with another question asking how a Coccittee member can express a strong feeling when the Comittee actually takes action by taking no action; it sets a precedent by not saying something in regards to an application.
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