ML20140G862
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January 29, 1969
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D. kiler, Chief Reactor Project Branch 1, DRL AGENDA AND SCHEDU12 FOR MIDLAND SUSCGOGTfEE METING, FEBRUARY 4,1969 Dr. Monson, the Midland Subcommittee Chairman, has identified the follow-ing specific subjects of interest and has recommended that the applicant be ready to discuss them during the February 4 meeting:
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Meteorology 2.
Dose Calculations 3.
Containment Leakage Rates 4.
Iodine Removing Systems for the Containment 5.
Evacuation Procedures Please recognise that discussion usy arise on other subjects even though i
they have not been specifically identified here.
A tentative schedule for the meeting is as follows:
11:30AM Meet with Regulatory Staff 10:30 j
11: 30AM-12:30PM ' Lunch 2:30PM Meet with Applicant 12:30 2:40PM Break 2:30 2:40 3:10PM Caucus Finish Meet with Applicant 3:10 The meeting will be in Room 1046 at E Street. The standby room for the applicant is 1062.
Originnt signed b7 J. I E3Td J. E. Hard Senior Staff Assistant
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EXCERIT FROM
SUMMARY
OF 103RD ACRS MEETING OCTOBER 31 - NOVEMBER 2, 1968 MEETINGS WITil Tile DIRECTOR OF REGULATION AND Tile REGUIATORY STAFF 10.
Dow Midland Plant Mr. Price reported that DRI, had received Part B of the Dow Chemical applica-tion to construct a PWR (f4W) for electrical power and 4 million pounds of process steam for manufacturing pharmaceutical chemicals. Dow has encountered a problem with the Food and Drug Administration and the DeLaney clause of the Pure Food and Drug Act which prohibits any man made radioactivity in drug products.
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1/3/49 Frg. lect: Midland Plant jgggg Population related site evaluation S S 9 9
General Bescription:
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y.3, p Twin BdN reactors of conventional design, owned and operated by Consumers Power Company, are to be employed for generation of electrical power and process steam. b process steam is to be directed to the Dow Chemical Company plant located immediately to the north of %e reacters. A perties ef the Dow plant is within the reactors' exclusion area. h sity of Mid-land is located just north of the Dow plant and within 1 1/2 miles of the reactors. h entire city is within the Low Population Eoes defined by the Applicant (radius = 3 mi). Combined business and residential population numbers are larger than for the Indian Point site at all radii est to about l
7 miles.
Results of Site Visit:
since only two subcessaittee members were able to visit the site, a follow-up Subcommittee meeting was planned. h concerns meted by the Regulatory Staff (see below) were discussed with both the Staff and Applicant. Evacua-tion plans for the Dow plant and for Midland received considerable attention as did the matters of possible contamination of the process steam and control of plant activities in the reactor exclusion area.
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Items from DRL Report:
i This report has no conclusions; the subcommittee has been told that the con-clusions would be presented orally by Mr. Price at the February ACRS meeting.
The Staff points to difficulties they are havir.g in the areas of evacuation, application of 10 CFR 100, population density, meteorology, and post-accident doses. h Staff also observes the lack of iodine removing safeguards in the containment and notes that the containment design leakage rate is higher than at other plants.
l Schedule and Amanda:
8:30 10:15AM Executive session 10:30AM Break
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10:15 11: 30AM Meet with staff l
10:30 i
11:30AM - 12:30PM Lunch 2:30PM Meet with Applicaat 12:30 2:40FM Break 2:30 3:10FM Caucus 2:40 Finish Meet with Appliennt 3:10 b applicant is to be particularly ready to discuss the following:
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c.o tainment lea a.e rates sumA=, Iodine.. removing.. systems for the containment JEHard:emb 5.
Evacuation proicedures
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JEE:amb 2/3/69 1E23251: Midland Flant 1g333 : Population related site evaluation k'.,.,*.
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,sv General temeristion:
Twin SW reactors of conventional design, owned and operated by Constasere I
Power Company, are to be employed for generation of electriasi power and The process steam is to be directed to the Dow Chemisal process steam.
A portion Company plant located immediately to the north of the reacters.h city of Mid-of the Dow plant is within the reactors' exclusian area.
land is located just north of the Dow plant and within 1 1/2 miles of the h population Center Distance as defined by 10 CFR 1001sw600 feet and nearly the entire city is within the Low Population Esse (radius =
reactors.
4 Combined businese and residential population muebers are larger than 3 mi).
for the Indian Point site at all radii out to about 7 miles.
Items from DEL Reoort:
This report has no conclusions; the Subcomeittee has been told that the con-clusions would be presented orally by Mr. Price at the February Acts meettag.
h staff points to difficulties they are having in the areas of evacuation, application of 10 CFR 100, population density, meteorology, sad post-accident N Staff also obcarves the lack of todine removing safeguards in the doses.
containment and notes that the contalement design leakage rate is higher than g
j at other plants.
Results of 1/22/69 Site Visit and 2/4/69 Subconnittee Meetina h Subcommittee identified the three particularly different aspects of this applications population distribution and related questions, sooling pood for heat dissipation, and the use of steam from the reactor steam geeerstors in the Dow processes, bn the population distribution is analysed using the Erges-Mosson amethod, f
Evacuation the site looks acceptable in souperison with the reference site.
plans exist for the Dow plant and are being discussed for Midland; the subcein-
-mittee felt that more discussion is required on this question and the applicant was read a list of evacuation related questices prepared by Dr. Thompson.
Fost-accident off-site doses have been calculated using meteoretogical assump-tions that have been questioned by the gesff and by Dr. 01fford. h applicant i
I reports acceptable doses even though the containment leak rate is high 10.21./ day)
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and even without iodine-removing engineered safety features. h applisant Dr. Manson aboerved stated that there was room for compromise in his design.
that the Ezclusion Area radius and Low Fopulation Eoes distsace both saa he redused seasiderably if additional safeguards are employed.
FILE: Midland project file
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- E entation., describing the et,te ACRS h"$)dsame will-be -ready.-with-.a.ama-hour.. pron use of the reactors.
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<L EXCERPT FROM Sl?CLARY Or 106TH ACRS MEETI AG FEBRUARY 6-8, 1969 SPECIFIC PROJECTS i
Midland Plant - The Committee completed its preliminary site review of the applica-tion by the Consumers Power Company for authorization to construct the Midland' Plant.
The Committee concluded and reported orally to the applicant that: "The Committee has reviewed the proposed Midland site primarily from the standpoint of population and populatica density. The following remarks therefore reflect Committee conclu-sions based primarily on that one aspect of site evaluation.
"The Committee considers the site proposed to be unacceptable for use with reactor plants designed and analyzed as presently described in the PSAR. However, it believes that the site may be acceptable for use with reactor plants of the proposed power rating if:
(1) The facility is equipped with adequate engineered safety l
features and protective systems; (2) the facility is analyzed sufficiently conserva-tively - particularly in respect to: determination of exclusion area and low popula-I tion zone; assurance of low potential doses at short distances from the reactor in the unlikely event of a serious accident; evaluation of the number and location of people who could be safely and quickly evacuated in such an event; and, use of 1
assumptions, for example those related to meterology, in dose calculations; (3) the facility is designed, constructed, and utilized sufficiently conservatively; and (4) the facility is provided with thoroughly structured, effective emergency plans, including evacuation plans."
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Significant factors in the Committee's considerations were the high population within four niles of the site, the minimal engineered safeguards proposed in the j
application, and the use of less than conservative assumptions in the dose calcula-tions.
The Ccmmittee discussed the items listed below and the results of those.di,scussions
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are indicated: #
a.
The cost of energy in the Midland area - the costs of coal are high and the demand is not large enough to enjoy the price benefits of unit-train transpor-tation.
b.
Dow Chemical needs process steam as well as electrical power - the nuclear A ant. needs to be located close to the Dow plant to minimize steam line costs and l
saoem losses. There is some indication that Dow may phase-out the Midland works if a new source cf less costly energy is not found.
Exclusion area and low population zone - the exclusion area extends 1100 c.
meters from the proposed plant and includes a portion of the Dow plant, including
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- 333 Dow coployees'; the low population zone extends to three miles and includes all of the Dow plant and part of the city of Midland. The site received a '34 index i
rating when compared to the hypothetical reference site (considering the maximum population in the Dow complex).
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d., Evacuation - Consumers Power has made arrangements with Dow to evacuate The 353 Dow employees all Dow employees on order from the nuclear power plant.
within this exclusion area can be evacuated within ten minutes of notification, 907. of the remainder of the Dow employees would be evacuated within twenty minutes, The and ' complete evacuation could be accomplished 45 minutes after notification.
city of Midland has an evacuation plan and has conducted partial evacuations of some segments of the city.
Engineered safety features - the applicant does not propose to install e.
en cooling an iodine clean-up system in the containment s ucture..The co tai
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sprays are p borated additive,
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Protection of the nuclear plant and operators from accidents originating within the Dow compicx - the release of chlorine gas appears to be an accident that could be hazardous to the nuclear plant. The applicant did not believe that a 1
chlorine accident could cause shut-down and evacuation of the nuclear plant. The ventilation system could be provided with chlorine filters and emergency air pacs l
and protective clothing could be provided for the operators.
Other aspects - the Committee mentioned but did not explore in any depth:
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the suitability of BEM reactors for marginal sites, protection required against reactor vessel splits, cavity flooding systems, and the use of process steam in products to be coisumed by people.
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MIDIAND PLANT Excerpt from Summary Letter dated February 14. 1969, revised February 17,1969 (106th ACRS Meeting, Feb. 6-8, 1969)
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Specific Projects The Comittee considered population-related factors of the application by the Consumers Power Company to construct the Midland Plant. The Comittee had the benefit of discussions with representatives of the I
Consumers Power Company, its contractors and consultants, and the AEC Regulatory Staff. The Comittee provided preliminary coments to the applicant and the Staff regarding this site and site-related matters.
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BORON CONCENTRATION MONITORING H. B. Robinson Unit No. 2, 4/16/70 Point Beach Units 1 and 2 "As methods for continuous monitoring of boron concentration and a more definitive determination of gross failure,of a fuel element are developed, consideration should be given to their implementation in this plant".
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FUEL - SPENT (Damage)
Indian Point 3, 1/15/69 "In the event that an irradiated fuel assembly is dropped or otherwise damaged during transit f rom the reactor vessel to the spent fuel pit, the cladding on the fuel rods may be ruptured with a consequent release of radioactivity.
In view of the relatively high population density close to the Indian Point site, the applicant should review the assumptions made in analysis of a refueling accident to see whether additional conservatism is warranted in assessing its effects and the provisions to cope with the accident. The matter should be resolved with the Regulatory Staf f".
Hutchinson Island Plant Unit No. 1, 3/12/70 Further study is required with regard to potential releases of radio-acitivity in the unlikely event of gross damage to an irradiated subassembly during fuel handling and the possible need for a charcoal filtration system in the fuel handling building.
This matter should be resolved in a manner satisfactory to the Regulatory Staff.
Point Beach Units 162, 4/16/70 "The applicant has deter =ined that turbine failure could release
.uissiles that might damage fuel elements in the fuel pool.
He has stated that, for each turbine, a second, completely independent speed control I
system designed to meet nuclear protection system criteria of redundancy, separation, and reliability, will be installed to reduce the probability of an overspeed condition.
In a related matter, in the evaluation of re-fueling accidents, studies pertaining to reduction of fission product releases have not been completed.
The Committee recommends that irradiated fuel not be handled outside the containment building until these matters are resolved in a manner satisfactory to the Regulatory Staf f".
Millstone Unit 2, S/15/70 Further study is required with regard to potential releases of radio-activity in the unlikely event of~ gross damage to an irradiated fuel assembly in the spent fuel pool.
This matter should be resolved in a manner satisfactory to the AEC Regulatory Staff.
Oconee Nuclear Station Unit 1, 9/23/70 In order to protect against the postulated consequences of the acci-dental dropping of a fuel element, the applicant has stated that either he will install filters in the fuel pool building exhaust system, or the equivalent control and. protection will be assured by another method.
This matter should be resolved to the satisfaction of the Regulatory Staff within the first year of power operation.
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INSTRtotENTATION (POST-ACCIDENT)
Palisades Plant, Jnnuary 27, 1970 The Committee recommends that attention be given to the long-term ability of vital components, such as cicctrical equipment,and cables, to with-stand the environment of the containn.cnt in the unlikely event of a loec-of-coolant accident.
This matter is applicabic to all large water-cooled pouer reactors.
Hutchinson Island Plcnt UniL1, March 12, 1970 Beaver Valley Pcuer Statier.. Unit 1, March 12, 1970 Information on a number of items, identified in previous reports of the t
Committee, is to be provided by the. applicant to the Regulatory Staff during <o'nstruetion.
These include:
b)
Review of development of systems to control the buildup of hydrogen in the containment, including an appropriately con-servative estimate of possibic hydrogen sources, and of in-strumentation to monitor the course of events in the unlikely event of a loss-of-caalant accident.
?oint Bench Nuclear Plant Units 1 and 2, April 16, 1970 l
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B. Robinson Unit 2, Acril 16, 1970 Studies by the applicant are underucy on two probices identified in prev-ious reports of the Cor:m:ittee, as follous:
(b)
- j Review of development of systems to control the buildup of hydrogen in the containment, and of instrumentation to mon-
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itor the course of events in the unlikely event of a loss-of-q coolant accident.
l Dresden 3, July 17, 1970 Problems experienced during preliminary operation of Unit 2 have shotm the need for some improvements in the plant and in operating procedures.
These improvements will also be incorporated in Unit 3.
One of the improvements consists of instrumentation to be installed in the primary' containment for remote monitoring of temperature and pressure over the i
full range of postulated acciden::s.
The Committee believes that instru-mentation should also be provided for monitoring high radiation levels by means more rapid than camplini; and laboratory analysis, Surry Power Station Units 1 and 2. December 17, 1971 The applicant should assure himself that instrumentation for determining he course of postulated accidents is on hand at the station and that ppropriate calibration methods and calculated. bases for interpreting instrument responses are available.
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Ovster Creek 1, Dec. 12, 1968 "It is recommended that supplemental and potentially more sensitive methods of primary system Icak detc ; tion be studied, evaluated, and impicmented if they provide significant improvements in measurement of leak rate, in the time needed to measure Icak rate, or in distinguishing the nature of the Icak. The study and evaluation should be completed within a year."
Nine 1111e Point, April 17, 1969 "The applicant plans to study supplemental and potentially more sensitive methods of ptimary system leak detection and to implement methods uhich provide significant improvements in measurement of leak rate, in the time needed to measure leak rate, or in distinguishing the nature of the leak.
The applicant should report to the Regulatory Staff his progress in this area witnin a year af ter start of power operation."
Duane Arnold, December 18, 1969 "It is important that nc leakage from the primary containment bypass the secondary containment and the associated filtering systems in the event af an accident. The applicant should study the ef fects of Icakage through
'ossible bypass paths, with particular emphasis on the main steam line
. solation valves, and should propose measures to deal with any such bypass leakage. This catter should be resolved in a manner satisfactory to the Regulatory Staf f during construction of the plant.
The Committee wishes to be kept informed of progress in this area."
Monticello Unit 1, June 15, 1970 "The Committee has on several occasions stressed the importance of inservice inspection and leak detection. It recommends that the Regulatory Staff develop a schedule of inspections for safe ends. The operation of the Icak detection and location sys'tems should be reviewed and m'odified as appropriate to obtain the maximum speed and sensitivity for detection of leaks. In addition, the applicant sho'uld study other techniques of detecting Icaks."
Millstone Station Unit 1, June 16, 1970 (Same as above)
Nine Mile Point, June 16, 1970 "The applicant is studying improved leak-detection methods. The Committee believes that detection and location of small leaks. is an essential part of the surveillance program. The applicant should expeditiously install
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uch leak-detection devices as seem likely to give improved sensitivity ar speed of Icak detection. The Committee rec'ommends that at least one leak-det,cction system in addition to the proposed sump accumulation rate
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and dew point systems be installed within a few months and wishes to be i
kept informed of progress in this regard."
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HEUTRON NOISE ANALYSIS Point Beach NucIcar Plant Units 1 and 2, Anril 16, 1970 The applicant has presented a program for preoperational tests of the plant including proof testing of the containments.
The applicant is performing studies to determine the appropriate number of tendons and the interval for tendon incpection.
The applicant is following the work of others for in-service vibration monitoring and loose parts detection so as to evaluate the applicability and appropriateness of impicmenting such means when de veloped.
Neutronic and external accelerometer signature measurements of the reactors during initial operation should be considered in order to pro vide a basis for comparisen with possible future monitoring results.
matters should be resolve 6 in a manner satisfactory to the Regulatory Staf f These Indian Point Nuclear Generatinn Unit No. 2, Seotember 23, 1970 "The applicant stated that neutron noise measurements will be made periodically and analyzed to provide developmental information concerning the possible use-fulness of this technique in ascertaining changes in core vibration or other displacements.
On a similar basis, accelerometers will be installed on the pressure vessel and steam generators to accertain the practicality of their use to detect the presence of loose parts."
_0conee Nuclear Station Unit No. 1, September 23, l'70 9
i "The Committee suggests that developmental techniques, such oc neutron noise analysis and use of accelerometers, be ccasidered as an aid in ascertaining displacements, changes in vibration characteristics, and the presence of loose parts in the primary systems.
tinuing use of some thermocouples in the core."The Committee notes the desirability I
I1 ndian Point Nucicar Generating Unit No. 2, September 23, 1970 The applicant cally and analyzed to provide developmental information conc
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p,ossible usefulness of this technique in ascertaining changes in core vibration or other displacements.
be installed on the pressure vessel and steam generators to ascertain t practicality of their use to detect the presence of loose parts.
Quad Cities Station, Units 1 eno 2, March 9, 1971 The Committee recommends taken as part of the start-up ord power ascension test pro 2 ram -that a c should be resolved with the" Regulatory Staff.
This matter It is also recommended thtt consideration be given to the use, on a developmental basis, of neutron noise easurements, accelerometers, or other devices to provide information con-arning the occurrence of excessive vibrations,. structural damage, or loose parts.
The Committee wishes to support and encourage continuing efforts by the applicant to develop improved methods of inservice pressure vessel in-spection.
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kl 2Lk POST-ACCIDMET OPERATICM Palinades Plant, Jcnurrv 27, 1970 "The Committee recommends that attention be given to the long-term ability of vital components, such as electrical equipment and cables, to withstand the envircnment of the containment in the unlikely event of a loss-of-coolant accident. This matter is applicabic to all large water-cooled power reactors."
Hutchinson Island Pinnt Unit No. 1, March 12, 1970 Beaver Vallev Power Station, Unit l'o.
1, March 12, 1970
- Information on a number of items, identified in previous reports of the Committee, is to be provided by the applicant to the Regulatory Staff during construction.
These include:
b)
Review of development of systems to control the buildup of hydroacn in the containment, including an appropriately conservative esti-mate of possible hydrogen sources, and of instrumentation to monito:
l the course of events in the unlikely event of a loss-of-coolant accident.
Point Beach Nucionr Plent Units 1 and 2, Aoril 16, 1970 H. B. Robinson, Unit 2, April 16, 1970 Studies by the applicant are undcruay on two problems identified in previous reports of the Committee, as follous:
(b) Review of developm'ent of systems to control the buildup of hydrogen in the containment, and of instrumentation to monitor the course of events in the unlikely event of a loss-of-coolant accident.
Fort St. Vrain Nucicar Generating Station, May 12, 1971 "The applicant has identified electrical equipment which is required to operate during and. follouing postulated accidents, and has detcrmined the environmental conditions to which this equipment will be exposed. Test procedures have been developed, and tests are being made to provide assurance that this essential equipment will perform its functions.
The results of these tests should be reviewed by the Regulatory Staff prior to operation of the plant."
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QUALITY ASSUP.ANCE i
Pilgrim, April 12, 1968 "The Committee recommends that the Boston Edison Company assume an active role i
in quality assurance in all stages of fabrication and construction.*"
i Kewaunce, May 15, 1968 i
t "The Committee continues to emphasize the importance of quality assurance in fabrication of the primary s:tstem as well as inspection during service lif e, l
j and recommends that the applicant implement those improvements in quality that are practical with current technology. The Committee also calls attention to those matters previously emphasized, which it deems to be important for all large water-cooled power reactors."
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Zion. July 24, 1968 The Committee continues to emphasize the need for quality in the manufacture, i
O The storage, and installation of the reactor and primary system components.
applicant described the quality assurance program that he and his contractors In this connection, the applicant i
- ntend to carry out for this purpose.
escribed the testing program for engineered safety features, including a full l
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l flow test of the eme*gency core cooling system delivering water to the reacter
/essel. The Committee recommends that the applicant give further consideration l
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to testing the containment spray systems with full flow to the spray nozzles at least once at an appropriate time during construction."
l Russellville, Sep. 12, 1968 i
"The Committee emphasizes the importance of the implementation and management l
of the quality assurance and quality control programs necessary to achieve the design, construction, and operation objectives."
a1 Donald C. Cook. December 1968 "The Committee continues to emphasize the importan*cc of the implementatien and management of the quality assurance program."
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i Indian Point 3, Jan. 15, 1969 "The Committee also emphasizes the importance of independent action by the i
' applicant.to assure quality in the construction of the facility."
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"*The Committee b,elieves that the industry should continue to pursue an l
orderly. program Icading to further improvement in the quality of pressurc vessels and other ecmponents of the primary system such as valves, pnnps, J
and piping."
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Monticello, Jan. 10, 1970 and Millstono 1, Jan. 15, 1970 "Continu!.ng research and engineering studies are expected to lead to enhanecmen of the safety of water-cooled reactors in other areas than those mentioned, fo; exampic, by the determination of the extent of the generation of hydrogen by radiolysis and by other sources in the unlikely event of a loss-of-coolant i
accident, developm.cnt of instrumentation for in-service monitoring of the pressure vessel and other parts of the primary system for vibratien and detection of loose parts ir the system, by the development of further means of preventing common failure modes f rom negating scram action and of design features to make tolerable the consequences of failure to scram during anticipated transients, an3 evaluation of the consequences of water contaminati by structural materials and coatings in a. loss-of-coolant accident. As solutir to the prob 1 cms develop and are evaluated by the Regulatory Staff, appropriate action should be taken by the applicant on a reasonabic time scale."
palisades, Januarv 27, 1970 f ontinuing research and engineering studies are expected to lead to C
enhancemer.c of the safety
~ water-cooled reactors in other areas than those mentioned:
for example, by determination of the extent of the generation of hydrogen by radiolysis and frca other sources, and development of means to control the concentration of hydregen in the containment, in the i
unlikely event of a loss-of-coolant accident; by development of instrumentation for inservice monitoring of the pressure vessel and other parts of the primary system for vibration and detection of loose parts in the system; and by evalua-tion of the consequences of water contamination by structural materials and coatings in a loss-of-cooland accident. As solutions to these probicms develop and are evaluated by the Regulatory Staf f, appropriate action should be taken by the applicant on a reasonable time scale."
Beaver Vallev, March 12, 1970 "Information on a number of items, identified in previous reports of the Committee, is to be provided by the applicant to the Regulatory, Staff during construction. These include:
a) A study of means of preventing common failure modes from negating scram action and of design features to make tolerabic the consequences of a failure to scram during anticipated transisnts, b)
Review of development of systems to control the buildup of hydrogen in the containment, including an appropriately conservative estimate of possible hydrogen sources, and of instrumentation to monitor the course of events in the i,
unlikely event of a loss-of-coolant accident."
Hutchinnon Island,ifarch 12, 1970 (Same as above) 10.7-5
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" Continuing research is expected to enhance. safety of water-cooled reactore in other areas than those mentioned, for exampic, by the determination of the extent of radiolytic decomposition of cooling water in the unlikely event of a loss-of-coolant accident, development of instrumentation for in-service monitoring of the pressure vessel and other parts of the primary system for vibration and detection of loose parts in the system, and eval-uation of the consequences of water contamination by structural m'aterials and coatings in a loss-of-coolant accident. As solutions to the problems develop and are evaluated by the Regulatory Staff, appropriate action should be taken by the applicant on a reasonable time scale."
Dresden 2, Sco. 10, 1969
'hany improvements in safety features and-procedures have evolved since the Dresden Unit 2 provisional construction permit was granted, as a result of the work of reactor suppliers, the AEC, and others.
Some of these improvements hav been discussed in recent ACRS construction permit and operating license report.
The applicant has agreed to incorporate several of these improvements in Dresden Unit 2.
These include an improved emergency cooling system, flooding protection for the cmergency cooling pumps, provision of an interlock to pravet depressurization by the automatic pressure relief subsystem if low-pressure emergency core ccoling pumping capability is lost, and installation of a stror.
motion seiscograph."
l Duane Arnold, Dec. 18. 1969 and Shoreham, Dec. 18. 1969 also Sequevah. F<S.
(except (c))
"Information on a number of items, identified in previous reports of the Committee, is to be provided by the applicant to the Regulatory Staff during construction. These include:
"(a)
A study of means' of preventing common failure modes from negating scram action and of design features to make tolerable the consequence of failure to scram during anticipated trans-icnts.
"(b)
Review of development of systems to control buildup of hydrogen in the containment following a loss 'of-coolant accident.
"(c)
Anhlysis of method.s to limit damage to the spent fuel pool and to reduce release 'of fission products in the event of a drupped fuel cask.
"Other problems related to boiling water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports. The Committee feels that-resolution of these items should apply equally to the Arnold plant."
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SUMMARY
OF 196th' o
E stEETING 8/12-14/76 Y
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- 5. Midland Decision Guidance from the Co mission should be sought before the Connittee expands upon the generic concerns noted in its construc-tien perialt report on the Slidicnd Plant.
o D. Midland Decision _
The Chairman noted that the Federal District Court has found t
the statement made by the Comittee with regard to generic items in the It was anticipated that Midland report lacked necessary specificity.
the Cocaission wauld request some action by the Comittee this deficiency.
a request was received from the Comission.
A. Court Decisions on Midland and Vermont Yankee L. V. Gossick noted that B. C. Rusche would inform the public via s press conference of the NRC positions resulting from the District Court of Appeal Decisions regarding the Midland and Vermont Yankee licensing actions. He noted that for the intnediate future, NRC was not planning to issue any full power licenses, but did plan to permit Salem and Calvert Cliffs Plants to develop up to 1% full power in order to carry out a number of their preoperational tests. He stated that the NRC Legal Staff believed that these decisions will prevent the issuance of OLs, cps and LWAs until the matters concerning waste nanage-ment have been satisfactorily addressed.
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Excerpt from:
Minutes of 202nd ACRS Meeting ltecting Dates: February 10-12, 1977 VIII. Executive Sessions (Open to Public)
A. Request for Further Clarifications of the ACRS Report on the Midland Plant
%e ACRS Executive Secretary noted receipt of a memorandum from F. J. Coufal, Chairman of the Atomic Safety and Licensing Board, 28, 1977, requesting further clarification of the dated January Comittee's report on the Midland Plant, Units 1 and 2 (see Mr. Coufal is of the opinion that the Committee's Appendix XXVI).
supplemental report on Midland Plant, Units 1 and 2, dated 18, 1976, does ot meet the requirements set forth by November District of Columbia Circuit Court Case, Aeschliman vs. NRC.
he Comittee agreed to seek the advice of the General Counsel of NRC, and to consider a reply to this request at the 203rd ACRS Meeting to be held in March 1977.
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lI A Excerpt from Minutes of'203rd ACRS lteeting Meeting Date: " arch 10-12, 1:-
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r V. Meeting With the NRC Staff on Recent Ooerating Ex6erience, Licensing Actions, and Future Agenda f
D. Midland 1: Containment Liner Problem K. Seyfrit, NRC Staff, discussed a recently occurring problem at the Midland Plant, Unit 1, where a bulge developed in the containment liner. his plant is currently stdl in the con-struction stage.
(For details of the bulge, see Appendix XXVIII.)
He stated that the cause of the problem is believed to have been a leak in a water line used to cool the containment concrete during curing. Had the curing process been completed, this water line would have been grouted. He noted that this is a preliminary report, and that details will be reported at a later date.
I K. Seyfrit stated that the Applicant has cut sections of the bulge away to clear debris. The method to be used to renair the liner has not yet been determined. We Applicant has not yet analyzed the pressure which caused the bulge to see whether or or not it was adequate to fail the liner welds.
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