ML20244C768
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SEP 21 1964 f
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U. S. ATOMIC ENERGY COMMISSION y
JW DIVISION OF REACTOR LICENSING
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]k//q REPORT TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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IN THE MA"TER OF j
NIAGARA MORAWK POWER CORPORATION 4-APPLICATION FOR CONSTRUCTION PERMIT Mote by the Director. Division of Reacter Licensing I
l The attached report has been prepared by merabers of the Division of Reactor Licensing for consideration by the Advisory Cocunittee on Reactor Safeguards at its October,1964 meeting.
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. Introduction 1-
. II.
Suassary 1
III Safety Considerations 2
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A.
Site 2
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Water Usage 3
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Meteorology 4
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Geology and Hydrology 4
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4 Seismology and Seise.ic Design 5
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Effect of Liquid Effluent on Aquatic Life 7
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Environmental Monitoring 7
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B.
Containment 8
C.
Engineered Safeguards 11 4
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Isolation Valves 11 2.
Containment ' Heat Removal Equipment 11 D,
Electrical Power 13 E.
Instrumentation 14 F.
Re. circulation Flow Control 15
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IV Safety Analysis 16 e
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Routine Operation 16 f
1-1.
Radioactive liquid waste 16 2.
Radioactive gaseous waste 17 3.
Radioactive solid waste 17 B.
Accident Evaluation 18 i
1.
Steam Line B:eak Accident 18 l.
2.
Loss of Coolant Accident.
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3.
Refueling Accident 21-
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4.
Control Rod Ejection Accident 22 l
5.
Zirconium-Water Reactions 23 l
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V Conclusion 25 4
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Introduction The Niagara Mohawk Power Corporation applied for a Construction Permit on
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l April 1, 1964.
In suppcrt of this application, a report entitled Preliminary Hazards Summary Report Ene Mile Point N? clear Station was submitted for f
review.
In this case, the applicant is Niagara Mohawk Power Corporation and the S2neral Electric Company is the nuclear subcontractor. Niagara Mohawk will act as its own archievet-engineer.
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In addition to the staff technical review, meetings were held between the applicant and the Reactor i.icensing staf f on June 3 and 4,1964. An ACRS
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Subcommittee meeting was held near the proposed site in Oswego, N. Y. on June 10, 1964. As a result of the discussions during these meetings, ques-tions were sent to the applicant on July 10, 1964. Answers to these questions were received on August 17, 1964 and were incorporated into the First Supple-ment to the PHSR. A further meeting with the applicant was held on September 17, 1964, to discuss the answere to the July 10, 1964 questions as well as some additional technical 1,roblems.
II.
Summary g
The Niagara Mohawk Power Corporation proposes that the Nine Mile Point l
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facility be located on an approximately 1500 acre site on the southeast shore y
of Lake Ontario, approximately 7 miles northear t of Oswego, New York.
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The reactor is to be a 1538 Mw(t) boiling water direct cycle unit from which I
approximately 525 Mw(e) gross will be generated. The unit will be installed
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in a pressure suppression containment structure similar to that used in j
Humboldt and proposed for Bodega and Jersey Central. This structure will be designed to withstand the pressure which would result from a release of
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the thermal energy which is stored in the primary system. With active engineered safeguards, it will also control over an extended period of time
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the temperatures and pressures which would result from release of fission l
product decay energy. A reactor building will be built above the contain-1 ment structure and leakage from the containment structure will be into the h
reactor building.
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The applicant has considered the potential consequences to the environs of i
a number of accidents, including a main steam line break, a loss of coolant accident, a control rod ejection accident, and a refueling accident. Although we do not agree with every detail of the applicant's assumptions used in evaluation of the consequence of these accidents, we believe that the plant can be operateo within the requirements of 10 CFR 20 and 10 CFR 100.
1 We have reviewed in some detail the requirements and design criteria for engineered safeguards which would maintain the facility in a safe condition in the event of a serious accident. We believe they are adequate and con-sistent in principle with those already considered acceptable for other facilities of a similar size with concrete containment structures, provided an additional drywell heat removal system is added. We understand that the applicant is considering such an addition.
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We have Latified the following problem areas which we believe should receive the attention of the Committee and which are discussed in following sections of this report:
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Containment a.
Criteria for time of ventilation isolation valve closure.
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b.
Retest pressure.
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Combustion of hydrogen evolved from a potential zirconium-water
_g reaction.
2.
Adequacy of engineered safeguards to remove fission produced decay f
I heat from containment.
3.
Recirculation flow control.
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4.
Inclusion of a period trip in the nuclear instrumentation.
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Means of limiting the consequences of a control rei ejection accident.
l III.
Safety Considerations B
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The proposed Nine Mile Point facility is similar or identical in major respects a
to both the Bodega Bay and Oyster Creek facilities which have been reviewed recently by the Consmittee.
The same general problems concerning the facility i
4 design are in evidence in the proposed Nine Mile Point facility.
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-i since the applicant, Niagara Mohawk Power Corporation, has not previously I
appeared before the Committee, and since this application is being considered V
on its own merits, the principal safety considerations are discussed in this
,j report even though they are similar to those considered for Bodega and Oyster I
f Creek.-
A.
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The Nine Mile Point site occupies approximately 1500 acres of land in a l
rectangular plot on a broad promontory on the shore of Lake Ontario, seven miles northeast of the City of Oswego in Scribs township, Oswego County, New York, and 36 miles northwest of Syracuse, the nearest largs city to the site. The site is generally flat with a gentle slope toward the lake. The station elevation will be 260 feet MSL, and the normal lala level is 246 feet. About 200 acres along the lake shore in the central j
portion of the site is cleared of trees, and all structures. of the pro-i posed facility will be erected in this area, with the reactor located approximately 400 feet from the lake shore. The shortest distance to
_p the site boundary will be 4000 feet to the west, and the nearest. residence will be on Lake View Road, approximately 5200 feet in the same direction.
Distances to the ether boundaries are over a mile to the east, and about a mile and half to the south. The land seen surrounding the site is
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largely used for pasture and light agriculture.
The following tabulation shows the population distribution within 20 miles of the site.
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svl 1 {:. a g' POPULATION WITHIN TWENTY MILES OF THE REACTOR BUILDING t Or <3 1960 Census Figures ik ) Distance (Miles) Populatfon in increment Total population within radiusi i i 1 0 '0 2 300 300 i' 3 500 800~ !L 5 1,100 1,900 .i 10 28,900 30,800 20 44,000 74,800 r-t % l-. 9 As can be seen from this table, and scrutiny of the data presented in the y' PHSR, the population density of the area surrounding the site is low, t being less than 100 per square mile in all sectors within five mU es. The nearest village, Lycoming, is 2 miles from the site and has s' popu-The next ne' rest significant population is the City of l' lation of'125. e j. Oswego at a ' distance of seven miles, with a 1960 population of 22,155. l Adjacent to the nearest site boundary (to the west), along the lake shore, ] is a 12 acre plot known as Lakeview, operated as a summer camp eight weeks k j per year. _There are up to 500 people at the camp during the week, and { ( 1500 on week-ends.. There is a scattering of summer cottages along the t g lake shore, the nearest 'of which is about 8000 feet to the east.. g ?. i S. A comparison of the 4000' feet exc1 sion radius available f O( TID 14844 exclusion distance' fa roximatel 0.92 mile adicates that I engineered safeguards more effec ve than as d in 4844 by a j factor of approximately '7 will be required in order for this reactor j 4 f k( t$ be 'within the guidelines set [ orth in Part 100. Similarly, a factor / ~ i L} i of 21 jvill be required for this h actor to be within the guidelines set' 'O sforth'in Part 100 regarding the cit /Mistance, if the City of Oswego j' 7 Gosulation 22,155) is used for a reference. Otherwise, the factor will
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og gg 4 ,H i 1. Water Usaae TI D ; /7<7 .f, [ Lake Ontario, one of the nation's mejor water resources, is used as a source of water supply by variea-amicipalities and industries ] v along the shore, for shipping, fer commercial and sports fishing, i and for extensive recreation. The asarest public water supply intake is for the city of Oswego, and is located eight miles west of the proposed facility.- The mouth of the Oswego River empties into the lake between the plant and the Oswego water supply intake. As described lateir, ^ the temperature and current patterns established by this river, as well as the breakwaters extending into the lake near the river mouth, would tend to provide protection to the water intake from reactor' ef tluents that might be released into the lake in the condenser ' ( cooling water. It is anticipated that use of the Oswego intake
4 ~ t.L, 3' 1 ( ( 7 { po 0/ will be expanded to supply more of Onondaga-County, includ'ng'tho' i Sity of Syracuse. By the time this has been accomplished, this, intake ]- may very well furnish an essential and irreplaceable source of water j-for a very large number of people. Other intakes from the 1che are over 35 miles away, and do.not represent any problem. .{. 2. Meteorology A long period of record of wind direction and ~alocity is available q from a station ten toiles from the site in Os.uy In addition, one year of data has already_ been taken tnd analynd at the site. It 4 j has been found that the average wind velocity at the 203 foot level is 14.7 mph, and the frequency of calm is zero, indicating that the + wind velocity at this site is generally somewhat higher than it is in central parts of the country. The maximum e qected wind velocity is estimated to be 110 mph. Due to the influerv e of the lake, the frequency M Anversion varies drastically during the year, from a f: x minimum of 6% in the winter to 75% in June,,with an annual average I of 35%, similar to many other sites. The frequent passage of vigorous low-pressure systems tends to prevent the occurrence. of long periods n. f' with constant wind directio7. During 1963, 'there was only one occasion l when the wind persisted in one octant for more than 45 hours, and ( >j three occasions when it persisted for more than 36 hours. The data has been analyzed by the applicant with respect to wind direction,. speed, and stiabilit.y for one year, and appropriate diffusion para-g- meters for one hour, five days, and thirty days derived therefrom. l' The U. S. Weather Bureau has examined the meteorological data for t this site and has advised that the diffusion parameters utilized l- -by the applicant are appropriate. However, they also suggested that i wind Sector D be reduced in size to examine the differences between J ow.r lo rd and over water trajectories, as well as the pos'sible marked differences in wind frequencies within the r,ector. This was done ,j-in the First Supplement to the PHSR in wSt the staff believes to be (, a satisfactory manner. The Weatbar Bureau also suggested the possi-i ble occurrence of other diffusion conditions not analyzed by the applicant which might have some effect on the annual average stack emission limit. However, preliminary estimates by the applicant and by the staff indicate that the stack limit will be large enough to 4 accommodate any small adjustments which might result from such con-siderations in the future. I-3. Geology and Hydrolony i 4 [ Bedrock at the site is. Oswego sandstone. The overburden above bedrock averages about eleven feet at the boring positions. It is estimated that the Oswego sandstone extends down to about 185 feet below the ground surface in the general station area. The Oswego ( sandstone grades into the Lorraine group, the Trenton group, and the Cambrian sediments between approximately 85 and 1700 feet. All
j _ M major structures will be founded c r the Oswego sandstone, which is E ' a very substantial rock and is resistant to weathering. l~ i A detailed study by the applicant of the surface drainage patterns I in the area of the site has shown that all parface water will run off .f into Lake Ontario, or percolate down a short distance and then move toward the lake in the underlying soil and rock. j, Iil A detailed study of lake currents in the vicinity of the site has been made. This was accomplished by using both drift cards and.the clorine j 1aden water from the Oswego River..With these tracers, it was possible to relate the surface lake currents, which would carry the plant effluent, to surface wind conditions, and thereby predict the frequency of occurrence of various lake patterna. It was found that the pre-dominant lake current is to the east from the plant, turning northward near the eastern end of the lake. The minimum yearly average dilution -calculated at any beach due to this current is twenty-one. Most of c the time it will be impossible for planc effluent to move across the [' mouth of the Oswego River to the Oswego water intake to the' west. It is estimated that it will be carried to this point by the winds only .g four percent of the time during a year, and that the dilution in the ( y ~[ lake will be a factor of twenty even when this happens. These two factors would give an average dilution factor of 500. In addition, j 1 . it is estimated that the c':nwd: ' sffect 'due to the submerged location 'l. of the intake will produce an aoaitional factor of 100 dilution. Hence, j there is a possibility that the concentration of radioactivity in the affluent could be permitted to exceed Part 20 without producing i excessive concentrations in the nearest supply of drinking water. l However, such a proposal has not been made by the applicant at this i time. 7 I The maximum measured level increase in Lake Ontario due to a seiche is 2.9 feet. Wave heights as great as twelve feet have been measured, j and at this location these can come up to tha shore before breaking. c Thus, it appears that the site elevation of fourteen feet above the t average lake elevation is just suf ficient to prevent flooding during severe conditior&s. j J on the We have requested that the U. S. Geologicel Survey conssent Copies of these lgeologicandhydrologicinformationforthissite.comunents will be i a 4. Seismolony and Seismic Desian There is no evidence of foldf.ng or faulting at the site or in.the regional geology, and the region is stated to be one of the most stable areas in the county. Historically, the nearest earthquake ( occurred fifty miles to the east at Lowville, New York, and resulted in insignificant acceleration at the site. This is also the location of the w arest known fault. The next nearest quake occurred at Attica,
sb ( ( -6 _ New York,110 miles west of the site. Tts magnitude of six would have produced'an acceleration of only 0.01 g at the site. The appli-cent has estimated that the largest shock likely to occur in the .[ region would be a Magnitude 7' shock, such as occurred in 1925 in' the St. Lawrence River Valley at a distance of 280 miles northeast. If this size shock were placed st Lowville,-the> ground acceleration at the site would be approximately 0.11 g. j]E The intensity of ground motion which would be' expected at the site was calculatad by a method developed by Kanai. This procedurs takes into account earthquake magnitude, epicentral distance, and the elastic } properties of the soil and rock in the area.- Using this method, the 7 [4 rf ground motion response spectrum for the Oswego sandstone, the material 3 on which all principal facility structures will be founded, was developed S and is given on Plate C-22 of the First Supplement to the PRSE. A $r/ f,) maximum ground acceleration of 0.11 g was calculated. )~s,
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The 'U. S. Coast and Geodetic Survey has been requested to examine the i seismicity aspectr af the site and has estimated that a ground accel-e s g eration of apprc,.mately 0.07 g is the maximum that has ever occurred J A $ ? in. this vicinit.f. On the basis we believe the applicant's assumptions _v / are adequately conservative. I The Class I structuras, those whose failure could cause a significant release of radioactivity or which are vital.to the safe shutdown and gab isolation of the reactor, will be designed'in accordance with.the c following criteria: } 1. Functional load stresses resulting from normal plant operation, j when combined with stresses due to the seismic accelerations specified in the ground motion response spectrum, shall be within -{ the working stress for the particular material involved. >j: 2. In the case of the containment structure, primary load stresses plus those resulting from pressure, tergersture and seismic accelerations specified in the ground motion response spectrum shall be within working values, f These criteria are in most important respects consistent with the seismic design criteria set forth for the other large power reactors now in some phase of the Consission's licensing p: ocedure. Typically two earthquake aceslerations have been specified; namely, an accel-eration that has occurred in historical time, and also a higher l acceleration that theoretically could occur. In all cases the facilities (7 are being designed to ride through the " historical" earthquake with combined normal operating, accident and seismic stresses on all com-ponents, structures, and systems important to safety at or near the established working stress of the material under consideration. For '{ the higher acceleration theoretical earthquake, no loss of function - of all systems important to safety is permitted; however, this implies I
l }. '.I.. f h r' & U1 c ,g 7 [ that in some cases stresses may reach to or be ond the yield stress of the material involved. In the case of Niagara Mohawk, they have elected td anae4fy oni a design acceleration corresponding to thp % oretical maxi 7,. ' lI. earthquake (0.11g) but are designing the facility in accordance with l the criteria normally specified for the "nistorical earthquake." This conservative approach is, of course, acceptable. Based on our .[ evalustion of the advice from the U. S. Coast and Geodetic Survey l T concerning the-maximum ground motions that have occurred in the past, we believe that-the applicant's earthquake design criteria are adequate for the seismic conditions likely to be encountered at this site, 5. Effect of Liquid _Ef fluent on Aquatic Life J 3 i The applicant has not established at this time a numerical limit on tha amount or concentration of radioactivity to be released to Lake J Ontario, but intends to base this limit on 10 CFR 20, taking into .i secount the composition of the discharge and the dilution and concen-(- tration mechanisma in the Lake..The study of the dilution character of the Lake previously discussed, and the applicants proposal to - carry out ecological and radiological studies on the marine life will y. -be important in establishing thiu release limit. It is our under-j standing, based on discussions, thet the applicant is maintaining liaison with officials of New fork' State on the radioactivity aspect .i. of their proposed operations. i p < We have' requested comments on the effects of liquid effluents on {' ' fish and shellfish from the U. S. Fish and Wildlife Service. Copies j of these comments will be made available to the Committee, when 5 received. .I e- 'f 6, En_virotunental Monitoring i .h An environmental monitoring proEram will be initiated'by the applicant y during the construction phase prior to operation of the plant. This monitoring will continue into the operational phase of the plant. This prograd will delineate background radioactivity levels, prior re operation, and provide a basts for evaluating the effects of the opera' tion of the plant, if any, on the environment. t j! (
i ( k [ I B. Containment l i The containment system proposed for the Nino Mile Point plant is a pressure suppression system similar to that installed at Humboldt Bay, and proposed for Bodega Bay and Oyster Creek. Ihe proposed system consists of two chambers, the drywell in which the reactor vessel and recirculation equipment will be situated, and the suppression pool to 4f, which the drywell atmosphere will vent in case of a pressure rise. i i The drywell will be a spherical vessel 70 feet in diameter with a 33-foot diameter cylindrical top. It will be designed and fabricated in accordance with Section III of the ASME Boiler and Pressure Vessel Code of ASTM A-212 Grade B firebox steel. The design pressure of the drywell will be 62 psig at a temperature of 280 F. The suppression chamber is a toroid in form, the centerline diameter f of which is 123 feet. The cross section is circular and 27 feet in i j dicmeter. It nill be approximately one-half filled with water. As with the drywell, it will be designed and fabricated in accordance with Section III of the ASME Boiler and pressure vessel code. This { vessel will be designed for a pressure of 35 psig. Communication between the drywell and the suppression chamber will j be by 10 large diameter vent pipes, a manifold in the suppression chamber, and finally a large nupber (approximately 120) of down-comers whi.ch are submerged in the suppression pool water. i l The specification of the Nine Mile Point containment desiEn pressures is based upon an analytical model developed from Moss Landing test results. The design pressures are based upon consideration of a i double ended break of one of the recirculation lines which results in a blowdown of the fluid in the primary system. It has also been 1 assumed that a relatively small steam leak has prepurged the drywell of air. This analysis yielded a drywell pressure of 31.6 psig (or j-46.6 psig with prepurge) and a suppression chamber pressure of 25.3 i psig. i f Figure VII-1 in the PilSR is a graph of the calculated drywell pressure as a function of time af ter the accident. The suppression chamber should be at approximately the same pressure since the two chambers are connected by vacuum breakers. In addition, the suppression pool will be connected to the reactor building by vacuum b;eakers. Based upon our review of the design of the containment system as I well as the assumptions used in specifying the design pressures, we believe the containment system will be abic to withstand the immediate consequences of all predictable accidents. [ Two airlocks for personnel access will be provided in the drywell. The doors of each airlock will be mechanically interlocked so that
( ( 4 _9_ l I ( ) only one door may be open at a time. No plans sre made for personnel access during power operation; however, limited access will be permitted t during hot shutdown. There will be three personnel access openings to the suppression chamber. These will be sealed with double gasketed bolted covers. Although the drywell ventilation system will be normally closed during .l operation, the ventilation ducts will be equipped with dual isolation valves in serias which will be closed by a high radiation signal. j The applicant is new considering closing the suppression pool ventil-ation isolation valves at a set time delay after receipt of a high drywell pressure signal. The advantage of such a scheme would be that the containment system could blow down to the atmosphere thus lowering 3 pressures. The disadvantage of this scheme is that the containment atmosphere is allowed to blow down freely to the atmosphere concurrent r J with an accident, a time when nothing is predictable. In view of the near impossibility of guaranteeing that there would not be a con- ) current release of fission products with overpressure in the suppression j f pool, we believe that all ventilation isolation valves to the containment l should be closed immediately by the same signals which initiate such { f emergency functions as closure of main steam lines isolation valves iand closure of other isolation valves. Typically such signals would } [ come from such sources as reactor water level, condenser vacuum, or l , containment pressure. i All fluid lines penetrating the containment barrier will have one or j two isolation or check valves, depending upon the function of the line, which will close upon receipt of such signals as high drywell pressure, low reactor water level, and low condenser vacuum. Signals i vill be received from two duplicate sensors in each of two independent j channels. The function will be energized by either of the sensors in both channels. Power for the channels will be provided by indepen-dent power sources. Power to close the isolation valves will be supplied 3 by the station battery or stored mechanical energy. We believe that f this isolation arrangement is acceptable. j At the present time, the applicant states that they are considering providing the capability of testing of individual penetrations on some regular interval. This would include capability for pressurizing I between the airlock doors and providing double gasketed seals with -} testing capability on the manhole covers in the suppression pool, the i drywell head, and the electrical and mechanical penetrations. Taps } for pressure testing between the ventilation isolation valves will also be provided. We believe that such testin;; capability would be highly desirable and should be installed. ( The specified leakage rate for the combined drywell and suppression pool is 0.57. per day at the suppression pool design pressure (35 psig). This is to be demonstrated during an initial leakage rate test with all penetrations installed at a series of pressures up to 35 psig so as to determine leakage characteristics as a function of pressure.
.? .p'_ n _ _... 5 '; v >.s j ! l- %. Retests are planned at a pressure of at least 5 psig. We believe that unless a retest oressure higher than 5 osia is used. It will be difficult 'c to extrapolate the measured leakage rate to the specified desian oressure - or n psig; however, this point need not be settled at this e b-- r As with Jersey Central, the possibility of. a metal-water reaction as 'jn i, a. consequence of an accident cannot be completely discounted. If this' reaction were to occur, the evolution of hydrogen would commence 11 - at some time after the initial pressure peak in the drywell and would [ not alter the design pressure. - However, the pressure would not decay q to zero as shown.in Fig VII-1 but would approach some higher value due to hydrogen evolution. Another consideration is the possibility of burning' or explosion from the excess hydrogen in the containment. We believe that it in aae==c===--- *e rel==
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a*nblems at this stage of the licensing proceedings; however, study should continue j, 'through constricElon~E5 assure that appr%~prTaH~ design or equipment '~j>' 'conNTderations are instaTied to prevent a _nforyn-oxygen reaction. J.la wouic violate containment integrity. __ ). The drywell for this proposed facility is relatively small in size - [l ~ and surrounded by an insulating shell of concre'.;e. As such, it' r shares the problem of all power reactors which are situated in con-crate containment. systems. That is, means must be provided to remove _ all phases.Appe ration, the y s sion 4raducLde cay _ hee t _inergy t henu nh _and principally after accidents. If this capability does not exist, I Mut'egrity of"th'sTntiinment system af ter an accident cannot be j guaranteed. This problem _ vill be discussed in greater detail in the " Engineered Safeguards" portion of this report. _ ~}j Th'e reactor' building will be a reinforced concrete ox-steel frame I structure with reinforced concrete walls. It will be situated I' directly above the containment system. The majority of piping, elec-j .t-trical and personnel penetrations will communicate with this building. 'l The exceptions are 2 main steam lines, 2 feedwater lines, and 4 extra penetrations which communicate with the turbine building. 3 In the event of an accident, an emergency ventilation system capable 7'[ of one air change per day will serve the reactor building and will provide a controlled path of release from the reactor building, via filters, to the stack. Leo complete full-capacity sets of equipment will be installed. The building is being designed for an inleakage of ' I: 1007. per day with a pressure differential of 1/4 inch of water. Thus, all radioactive materials leaking from the containment system, except those which may leak via the penetrations to the turbine building, will a enter the reactor building and subsequently be exhausted through-charcoal and absolute filters to the stack. i ( Except for those areas of concern noted earlier, we believe that the design of the containment system, including the reactor building, is adequate to fulfill its intended purpose.
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.g .,, o.. r1 ..1 [ !i m q 1 [h!... .C, Engineered Snfer,uards e . In the event of an accident, and an noted pievin'.i r, in order for the + naineered safe-Li containment integrity to be maintained, too act 'E . mordw mus t funct ion. These are inolation
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,.i containment ? - ; -heat removal equipment. The fact that the in,,o i ywell is. O relatively small and surrounded by concrete t.d e it unnditory that p
- t. cons be provided for. removal of ' fisslon.produt decay heat. :Specifi.
j celly, the two active engineered safeguarn a m.b :1;ned as follows: r 1. Isolation Valves: luolation valves will be installed in all fiold linet, which penetrate t he drywell and suppression chambo c, nn,va!ve in series which will.' close automatically, or by manual actum.1 <a, 2111 he in.atclied on lines which comunicate directly wit t. th' :.ne t rar steam supply T
- ystem or which open to the containment mi4 pho o.
On inlet 11on:, one of the two.valver coy h.>.. t' 4 v.d n* 1.ints uhich 4
- fi rm a ' closed loop inalde the centa n. ou c.nt;h could discharge
, C failurei will- . {'. 'i.ucti /a mt.c rial o.it side as i i. 1 1 l i nco connected quipped with one isolation v.alv. s (,
- e I he - nucl car etenm supply
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i' ni > n an valve. 4
- t mon to these Chich vill be cloned I y r.nnual net.. o ic i.
cr iteria are as follows: ~ 4 ) ie '..m J. a. No isolation will be provided en t'. co,. i hydraulic system lines. .r h b .Uc isolation valves will be provi.'- h ti nu. na! cent cooling system and in the lines betw m C suppr.m mn E ii' pool and core spray. . t' 'e Ci nenH:e d Fouction of the ventilation system 1. 4 1" 3_I. in s.ac containment ection ut tbi- ,t b ( j ac. int:rined the crit er in thnt itm-ir-I dividual it2ms of equipnent. n ;n i i i mlation -[ valves will be redundant..1%n r r er c l. i :elc r ion I valves will be supplied by the plant bu* O .to.u mechanical energy. We believe that the criteria for insi.1'c t la w'
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j isolation valves, except as discusned in th Contr.:n mnt Section two ccequate and of this report for ventilation system valve n s !~ consistent with previously approved reactors of similar nature. 2. Containment Heat Removal Equipment ( The containment heat removal equipment consist s of the core spray system and the suppression pool cooler. s ______-__-.m___._._____m__ m_
7. __ 4h4 a ..m.;e4.,, .s. ( ( .j C. Engineered Safeguards 'In the event of 'an accident, and as noted previously, in order for the e - i l containment integrity to be maintained, two active engineered safe-guards nust function. These are isolatten valves and containment heat removal equipment. The fact that the proposed drywell is relatively small and surrounded by concrete makes it manditory that means be provided for removal of fission product decay heat. Specifi-ll' cally, the two active engineered safeguards are designed as fol.'iws: .6 1. Isolation Valves 3 Isolation. valves will be installed in all fluid lines which penetrate ' the drywell and suppression chamber. Two valves in series which will close automatically, or by manual actuation, will be installed on lines which communicate directly with the nuclear steam supply system or which open to the containment atmosphere. On inlet 5 lines, one of the two valves may be a check valve. Lines which r. form a closed loop inside the containment but which could discharge ~ radioactive material outside as a result of a pipe failure, will .j be equipped with one isolatica valve. Instrument lines connected to the nuclear steam supply system will have one isolation valve j ( which will be closed by manual actuation. The exceptions to these i criteria are as follows: No isolation will be provided on the control rod drive a. (I hydraulic system lines. i b. No isolation valves will be provided in the post-incident ~l cooling system and in the lines between the suppression i, pool and core spray. b I Function of the ventilation system isolation valves is discussed in the containment section of this report. The applicant has -i' f maintained the criteria that the sensing devices and other in-f. dividual items of equipment required for actuation of the isolation valves will be redundant. Power for closure of the isolation a k valves will be supplied by the plant battery or by stored mechanical l energy. We believe that the criteria for installation and actuation of the isolation valves, except as discussed in the Containment Section ~l of this report for ventilation system valves, are adequate and j consistent with previously approved reactors of similar nature. ?. 2. Containment Heat Removal Equipment 1 The containment heat rcemoval equipment consists of the core spray { system and the suppression pool cooler. l +
<n . ( e ) The core spray system as now visualized by the applicant will consist _ of two independent full capacity units. In each unit, water will 1 ( be drawn from the suppression pool by a core spray pump which will I discharge to a spray header in the reactor vessel. Power for each j pump will be supplied by an independent bus. One pump will be I automatically switched to emergency power upon loss of outside j i power. The other may be manually switched to emergency power. l J The system will be actuated automatically by a low reactor water I l-level and will function when the system pressure is below 150 g psig. \\ The design cri':erion for the core spray system is that sufficient J cooling be provided by either onit to limit the fuel temperature l to prevent rapid oxidation of the zirconium fuel cladding. A { minimum flow of approximately 1850 gpm will be required to meet ) this criterion. 1 l The suppression pool cooling system will consist of two independent j i full capacity unitst Each will consist of a post incident cooling l pump of about 2000 gpm capacity and a post incident heat 6xchanger. h Each heat exchanger will be capable of removing from between about -( 3 and 13 N of decay heat energy depending upon the suppression ] pool temperature. This appears to be barely adequate, since the 1 y decay heat energy will be in the order of 13 N within 3 hours of j t. shutdown. The heat sink for the post incident heat exchangers will either be the service water system or the raw water cooling l sys tem. Power will be available from the facility diesel gener-l ator to operate this post incident cooling. i I j With reference to the interrelation between the core spray and suppression j pool cooling systems, it should be noted that both of these engineered l j safeguards must function to maintain the integrity of the containment j system in the event of a steam line break accident or a loss of coolant t. accident. This results from the fact that the only method of rejecting l{ heat from the containment system is through use of the suppression pool coolers. To underline the importance of these systems to public health and safety, we requested in Question 1 under Section VII, an evaluation of containment system pressure as a function of time assuming that cer-tain combinations of active engineered safeguards (i.e. core spray and post incident cooling loops) do not function. The principal result of I this evaluation is given in Figure VII-10 in the First Supplement. Note by reference to this figure and the text that in order to guarantee.the l integrity of the containment beyond 40,000 see (about 12 hours) it will be necessary that both the core spray and the post incident coolin loops function. There is no back-up to these two systems. g For comparison, consider Connecticut Yankee. Three active engineered ( safeguards were specified by Yankee, namely: (1) safety injection, (2) air recirculation, and (3) containment spray. Any one of these three, each of which is independent, and of a completely different I
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conedpt,'would protect the~ containment. In addition, each of these . systens has redundancy provided -in major components within the - syst em.' 'Even if the' safety injection system is discounted, since - it isn't operated by emergency power, one independent redundant 4 safeguard system out. of two would protect.the Connecticut-Yankee containment. 1 era for voltage and j eignal loss monitorinE and (5) adequccy of che variable range picoaxieters in lieu of period channels connected to the at.tomatic l/ safety system. It is, however, our criuton that these can be resolved with the applicant during final design and construction of r the facility. F. RecirculatLori Plow Control j 4 i The reactor power level is to be adjusted by positioning control i! rods and by varying the recirculation f1cv. The rods will be positiotted manual.ly in accordance to a program. The recircula' tion system as nca conten. plated will consist of 5 ( eeparste loops. Each wil1l take suction from the downcomer region and will discharge to the inlet plenuru belov %e core. The flow in each loop may be varied from 307. to 10C7. f ull flow. This will be accomplished..by using-induction mtors to drive the pumps. The ...+ ./ A.Y '
^2,o ( (. 1 ( frequency of the AC power used to drive the pump motor will be i varied by use of a drive motor connected to an AC generator by a variable hydraulic coupler. Flow c4atrol will be manual. The applicant is also considering, in the event of a severe , ;f decrease in distribution system frequency, use of an under frequency signal which would increase flow. e i. For this application, we have the same problem which was related 1 in the Jersey Central report. That is, during periods of operation ~! at reduced recirculation flow, the overpower scram set points are j .not correspondingly reduced. Thur, if through operator error in control rod positioning the power level were increased to the scram l set point of 1207. rated power while the recirculation flow remained low, some burnout of fuel would occur. In response to Question 111-13, the applicant has calculated that under these circumstances, f some 77. of the core would experience burnout. The applicant's present position is that if the BOR under these conditions (the currently accepted value is 1.5) is calculated to be below the operational limit, they will provide other than procedural controls. We agree thrt this is a resonable position, and that a more definitive { solution can be established later in the licensing procedure. 2 I J IV. Safety Analysis For convenience, the effect of operation of the proposed facility on the public health and safety is divided into two categories, namely routine 1 operation and accidents. A. Routine Operation f 1 1. Radioact!ve licuid vaste t All potentially radioactive liquid waste will be discharged 4 f-to the lake after dilution with the main condenser cooling water stream of 240,000 gym. A continuous sampling device j will collect a side stream of the effluent for periodic s { analysis. The criteria for discharge to the lake is that the concentra-tions will either be limited to 10*7 uc/cc at the point of discharge, or the effluent will be analyzed to assure that the t concentrations of limitinE 180 topes are below those specified f in 10 CFR 20. The liquid waste holdup tima will vary dependinh upon the particular operation in progress; howet er, the applicant has specified that on the average a one day holdup time would be available. q (
t;c p , 8. e (. ( 4 ( 2. Radioactive gaseous waste The majority of the potentially radioactive gaseous waste will be dischstged through the facility stack which will be at i least 300 feet tall. The criterion for release of these materials is that off-site doses will be within the limits specified in 10 CPR 20. i s' The principal sources of gaseous activity during operation f in the iosence of failed fuel is N-13 and A-41. The activity of othet short lived isotopes of 3xyFer and altrogen wili decay to es.:entially zero, because a half uour delay line between the air ejector and the stact wt11 be installed. If some f ailed fuel is in the reactor, isctopes of xenon, krypton, and iodine asy also be released. Figure IV-3 in the First Supplement is a graph of relative gas activity as a function of holdup time. This graph demonstrates for noble gases that the half hour holdup decreases the release level from the stack by a factor c,f about 50. Significant further reduction in noble gas g activity level or in halogen level is not achieved unless j [ g a holdup time in the order of days is considered. Other principal sources of gaseous effluents to the stack l will be from the turbine gist.d seal condenser, the reactor building vent air and the turbine building vent air. Continuously operated instruments will monitor the stack l i effluent. In addition, provision will be made to collect samples f:r particulate and halogen analysis. The holdup line to the stack from tre air ejector will be monftored and may be valved off if a high level is measured. j. p Some equipment such as the emergency condensers which may, under accident conditions, become highly contaminated, will be normally vented to the atmosphere. However, these j; vents will be znnitored and will be closed automatically upon a high radiation signal. 3. Radioactive solid waste f Radioactive solid waste will be shipped off-site for i disposal. We have reviewed the design provisions for handling radioactive waste materials resulting from normal operation and believe that they are suitable.
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, 3,, g ~,. ( g [. B. Accident Evatustion If 'all facets of the Niagara Mohawk pressure suppression scheme p operate as assumed and concurrent fission product release and high containment pressures is considered to be incredible, one is hard put to find an accident which would result in a significant j beslth hazard. This is illustrated in Table VII-1 in the FE8E which gives maximum off-site doses calculated by the applicant j[ for varf ous accidents. Rather than accept the applicant's version of the consequences of accidents, we have evaluated the 1 consequences of each of the major accidents by using less conservative assumptions. We have determined, in each case, that the off site doses are within those suggested in 10 CFR 100 provided that some combination of engineered safeguards function. We believe that these safeguards as now proposed by the applicant can be designed to function to such an extent l that the 10 CFR 100 guideline doses will not be exceeded. 1. Steam Line Dreak Accident i g This accident is a rupture of one of the 24-inch main f ( steam lines in the pipe tunnel outside the drywell. J (In the application it is called the maximum credible } accident.) The following assumptions are made to -l compute the off-site consequences: J 1. Isolation valves start closing 2 seconds after break and are completely closed 32 seconds later. f 2, 200,000 lbs of primary coolant are released. y: 3. Outside atmosphere is at 800F, 40f, relative l humidity. 4. Radioactivity in the water is as given in Table ) VII-13 in the First Supplement. i [ 5. Hemispherical shaped cloud 100 meters in radius i is formed. 6. Wind speed is 1 m/sec. U l With the exception of the cloud shape, these are, in our ii opinion, appropriately conservative assumptions. A significant increase in the off-site dose may be achieved if a cylindrical or cigar shaped rather than a hemispherical cloud is assumed. The longer cloud would serve to increase the time of passage over the site boundary. Depending upon I the cross section assumed, an increase in dose of perhaps a factor of 10 can be computed. Even in this event, the
3.= ( (, ( 19 - thyroid and whole body doses (approximately 100 and 3 rem respectively) would be below those given in 10 CF.1100 for serious accidents. The above is nnt ame====eit r---* } di l _to criticize th= manli e== * = = = =-=n ti an ; hue to 111 atrate 2 how sensitive the calculation is to changes in th's l 1tT3ifaption. s We believe the applicant's evaluation of the consequences i of this accident, with the exception noted above, is a i reasonable appraisal of' probable consequences. For the consequences of the steam line break accident to remain within the limits prescribed above, it will be necessary for the following active " engineered safeguards" to function: 1. The control rods. 2. The main steen line isolation valves. [ 3. The emergency condenser isolation valves. !j 4. Instrumentation required for actuation and monitoring 2 the above three functions. f 1 5. Condensate surge and storage systems'. 6. A control rod drive feed pump. t l 7. A power supply. I 8. Containment isolation. I i We have reviewed these required functions in terms of redundancy, reliability, capacities, and adequacy of emergency power, and believe that in the necessary respects the safeguards should function as required. I 2. Loss of Coolant Accident This accident is a rupture of one of the five recirculation 3 loops. The following assumptions were vade: .I l 1, 1. Reactor has been at a power level of 1538 MW(t) for ) 500 days. 2. Safety system closes containment isolation valves and ~ ~ initiates core sprayr~ ---- - / 3. Core spray limits core meltdown to 10%. C ~~ ~
.', J. .d: ( 20 - 4. Containment system leaks at a rate as specified in Figure VII-1 in the PHSE (Rate is scaled from 0.5%/ day as a function of pressure decay). ) l 5. Post incident cooling functions. 1 The following active engineered safeguards are assumed to ~ i function: -l 1. Containment isolation valves. 2. Core spray. 3.- Controls and instrumentation necessary to actuate and monitor the above. "1 4. Post incident cooling. 5. Reactor building cooling water. 6. Service water. 4 1 7. A power supply. To place the potential consequences of this accident in j i perspective, the following table starts with the TID assumptions and consequ6 aces, and show, as the various active engineered I safeguards are added, the reduction of off-site doses. 4i r T i , u e L ( t
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Site Boundary Condition 2 Hr. hyroid Dose ~ 1. TID 14844 assumptions 05%/dayleakege 2000 R j I 2. TID 14844 assumptions,10% core meltdown 0 5%/ day leakage 200 R [' 3. TID 14844 assumptions,10% core J meltdown, leakage decreases as a function of pressure 170 R 1 I i 4. TID 14844 assumptions,10% core meltdown, leakage decrease as a function of pressure, 70% per day } ? 10 R exfiltration l L
- 5. ~ Applicant's assumptions Less than 0.1 R he above indicates that a factor of 7 to 10 is needed to bring consequences down to those suggested in 10 CFR 100. To do this tLon is available (1) core spray, (2) reactor building, and (3) f
( the emergency ventilsition system. We believe that the combina_, 7 tion of these three safeguerds will accomplish this. -Q-t 3. Refueling Accident he refueling accident is the dropping of a new fuel assembly into a near-critical core. The ensuing reactivity input would be j 0.023 he resulting power transient was calculated by the applicarit to release 2950 W-sec-of which 60 W-see is due to Assumingthatthereactorgadbeen i zirconium-water reaction. l shutdown for 24 hours prior to the accident 0.93 x 10 curies i 6 j of noble gases and 0.70 x 10 curies of halogens were calculated i-to be released to the water. We find no reason to disagree with i these assumptions or with the methods of calculation of the power j transient and consequent fission product release. r t i The following assumptions were then made concerning the amount of p j noble gas and halogen fission products released to the reactor g building. 1 \\ t 1. All of the noble gases are released to the reactor cl building. 2. The halogen concentration in the reactor building is in j A ratio equilibriumwiththatdissolvedintgewater. of air to water concentration of 10- was assumed. ( 3 ' 100% per day of the reactor building volume is discharged ] i through the emergency ventilation system which has a l halogen a moval efficiency of 95%. l
,y 3w" g' { { t .y 3 a Webelievethatthesecondassumptiongboveissub,jecttodoubt. It would appear that the factor of 10- would be valid if equi-1 librium were-reached between the iodine concentrations in the water and air and if the released iodine did not come out of.the i . core and through the 50 ft. head above the core in a steam- ~]. bubble. - We believe that it cannot be guaranteed that equilibrium would be reached, nor that a steam bubble would not occur. }C g 4 The removal of 100% per day of the reactor, volume via the g emergency ventilation system requires (1) integrity of the reactor building not be violated and (2) the emergency ventila- -i '] tion system continue to function through the course of the accident. + In oziler to provide an estimate of an upper limit of consequences 1' to the public without the various mitigating assumptions,.we have a calculated off-site th oid doses for a 2. hour exposure assuming a alease of 0.70 x 10{euries of halogens and TID 14844 Lv assumptions. 'These are tabulated as follows: Assumptions Dose (Rem, Thyroid). ( l '. Ground level release from reactor l building in short time 30,000 Rem u; I 2. Ground level nicase from reactor i. building at rate of. 70% per day 2,000 Rem Ih 3 Release from reactor building at rate;of100%perdayviastackand' d
- f filters which are 95% efficient 50 Rem; i
a [ 'Ibe above illustrates that unless the integrity of the reactor L[t. building is maintained during refueling and tne emergency ventila-l tion is operable, doses in excess of those suggested in 10 CFR 100 j. may be received. We believe that both of these conditions can be _ met with detailed consideration of design. 5' 4. Control Rod E.jection Accident During nozinal operation, to remain within the thermal design criteria ctated for this core requi2es lov individual rod worth patterns to achieve allowable peaking factors. Thus, the per-i, .missible maximum rod reactivity worth decreases from 0.025 at low power levels to 0.015 at full power, and operating pro-eedures will be necessary to ensure that thermal criteria are met. Because of these 3 imitations, the applicant has stated that the nazimum insertion et reactivity resulting from a control rod ( e,)ection is 0.015 at full power. The applicant has stated that some other meatus, in addition to procedural control, such as a rod worth minimizer, or other suitable alternative will be used to limit the zwactivity available for insertion by any rod.
u--..--,. ( (_ i I I I - i j Two cases were analyzed, the first being a 0.015 reactivity insertion 3 at full power. Using conservative nuclear parameters, the energy released during the excursion was _ calculsted to be 3820 y Mw-sec. In addition, a 0.23% zirconium water reaction yielded 650.Mw-see and 2000 SCF of hydrogen. This results in a reactor i. vessel pressure peak of 1130 psia, approximately two seconds- ) after the initiation of the accident. Detailed calculations for the fission product release were not made, but the applicant P stated that the consequences would be much less than those for i the loss of coolant accident with 10% core meltdown. The Staff has reviewed the above accident and believes a realistic evaluation has been made. The energy input from this transient was impossible to validate without the use of a computer code; however,
- A the calculated pressure peak was found to agree with the input
'.b ~ energy. We also agree that the consequences would be less thkn the loss of coolant accident. 3 f( The second case analyzed was a reactivity insertion of 0.025 ht. .I 5 percent rated power. The excursion energy release was not given, 'j ,s but it was stated that the safety valves are open for about 3 seconds f at a maximum flow of about 50 percent ~of their total capacity. The ( \\- - resultant UO2 dispersion and mcmentum transfer to core componente; ) could not be predicted due to lack of confirmed analytical models j and/or direct experimental data describing the modes and efficiency i_ energy tr s er,from the fuel to coolant. After completion of the i f de'structivf 'SPERT.UO2 tests and the TREAT capsule tests, it should [ and vessel (darige.become pos ible yo provide more detailed answers reg .} We cannot fully evaluate this accident, because a detailed analysis ,i. was not presented. In addition, the UO2 dispersion and momentum i transfer could not predicted. The applicant has stated that if [ the results of this accident are not acceptable, the maximum amount of reactivity that could be inserted would be reduced.by { restricting rod patterns. Because there is a means to reduce the amount of inserted reactivity, it is our opinion that the applicant can proceed with the design and fabrication of the core. This matter of whether the worths of individual control rods are 4 excessive will receive further review prior to issuance of the operating licence .;f ^ 5 Zirconium Water Reactions In response to our request for additional information, the applicant ( discussed the magnitude of the energy and pressure contribution of any zirconium-water reaction to the drywell pressure as function of time for the 100% core melt case and its effect on the drywell leakage rate. j l
4n - 2h - ( Three analytical cases were chosen: 1. Core spray initiated normally. 2. Core spray 6elayed 15 rdnutes, t 3 Perfectly insulated reactor core with 100 percent meltdown. k In our opinion the model used in these three cases for the calculation of the percent zirconium water reaction yielded results consistent with data in the current literature. The model, also defines the relationship between volume of hydrogen evolved, time, and absolute temperature, which agrees well with available data. Case 1 There was no reaction because the peak clad temperature did not exceed 1600 F. 0 Case 2 This assumed one core spray pump inoperative and the other had a 15 minute delay before actuation. A total of 0.46% j of the clad vae calculated to react. The increase in j ( pressure is less than 0.5 psi due to hydrogen formation. 3 I Case 3 f In this ense only enough steam to supply the zirconium-vater reaction vs.s assumed present. Calculations were performed to obtain fuel temperatures as a function of time after the j accident. The reaction was censidered to have a threshold clad temperature of 18000F. A total of 23% of the zirconium in the core was calculated to react. The maximum increase in pressure due to the hydrogen is 22 psi at 3 5 hours, which when added to the transient pressure at that time falls below the design pressure of the suppres-l sion chamber - 35 psig. After condensation of water vapor in the dryvell, the resultant equilibrium presaura due to air and hydrogen is 10 psig. The maximum percent of hydrogen free volume during the accident was stated to be 54%. If all j of the oxygen in the free volume were combined with part of I the hydrogen evolved (either burn or explode) the pressure generated would exceed 100 psig. Containment integrity could not be maintained under these conditions. Since the containment integrity cannot be guaranteed for Case 3, we believe that some trovisic>n ghould be included in the design 7 to preclude a hydrogen oxyg5n reaction. tb ( '/ g pU a
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3 l L, V. R o ( ( l -y l ' V.- ' Conclusion .l Based on the foregoing considerations, and assuming that the reservations outlined in the sumanary portion of this report are adequately answered, we believe that the proposed facility can be constructed and operated at the Nine Mile Point site without undue risk to the health and safety of the public. t I. O t 4 1, 0 !,j t j r'i q 7 - - _ - _ -}}