ML20056B655

Expresses Appreciation for Opportunity to Meet W/Nrc on 661110 to Discuss Need for Availability of Monticello Generating Plant in May 1970

ML20056B655
Person / Time
Site:	Monticello
Issue date:	11/15/1966
From:	Mcelroy D NORTHERN STATES POWER CO.
To:	Price H US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9102070675
Download: ML20056B655 (28)

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NSP NORTHERN STATES POWER COMPANY M I N N K A!PO L,l e. M I N N E S OTA U S401 Ibvmber 15, 1966 Mr. Harold L. Price, Director of mgulation U. S. Atanic Energy Chunissicn Washingtnn, D. C.

20545 durrIFFT.Tn IRlCLEAR GEtERATING PIAITT E5979 Incket 50-263

Dear Mr. Price:

We appreciated the opportunity of meeting with you and Dr. Abrris in your office on November 10 to discuss our need for the availability of the bbnticello Generating Plant in May 1970, and possible means for expediting the licensing procedures so that a construction pennit can be obtained as early in 1967 as feasible.

During our visit with you, we left a copy of the attached written discussion, which contained on page 5 a proposed schedule of subnittals and meetings.

During our visit, you and Dr. Fbrris suggested scme revisions, and as a result we propose the following revised schedule of suhnittals and meetings:

Receive questions fran AEC on non-vessel Week of Novanber 21, 1966-subjects Forward to AEC AT:endnent No. 2, which Ibvember 21, 1966 will cantain detailed infonnation on site assanbly of reactor wssel Ibzward to AEC answers to questions Decanber 7,1966 on non-vessel subjects Abet with DRL staff and ACRS sub-Week of D2canber 5 or 12,1966 catmittee to discuss A'nendnent Ib.

2 (possibly at site)

R3ceive questions fran AEC on Week of Decmber 19, 1966 Aw.ndment No. 2 Ebrward to AEC answers to questions Ibt later than January 9,1967 on AT_ndment Ib. 2 Meet with DRL staff and ACRS sub-Waek of January 16, 1967 ominittee l

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N HERN OTATES POWER PANY e.

4 Mr. Harold L. Price 2

lbvember 15, 1966 Send DRL staff report to ACRS Prior to January 23, 1967 Meet with ACRS (full-day meeting)

Week of Ebbruary 6,1967 Hold public hearing Week of March 13, 1967 Ie ive construction permit Week of April 24, 1967

'Ibe new proposed schedule shows only ane full-day meeting with the ACRS l

rather than the two shown in our original discussion dated Nove:nber 10.

i We recognize that two neetings may be required, depending on the thorough-ness of our presentation and upon your staff review. We are hopeful that one long meeting with the ACRS will supply all the infonnation they need l

to recamend granting of a construdion permit.

After you and Dr. hbrris have had an opportunity to review this schedule, I would appreciate hearing fran you so that we can expedite our work and that of General Electric Ocmpany.

Yours very truly, I

D. F. McElmy Vice President - Engineering AVD/ DIM /do cc:

P. A. Fbrris L. C. Koke D. E. Nelson A. V. DienhcL4 J. B. Violette (2 extra copies to Mr. Price) l l

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Northern States Power Cbnpany Itnticello Nuclear Generating Plant

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'IHE NEED FOR PLANT AVAIIARTLTTY IN 1%Y 1970 m

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h operation of the Monticello Nuclear Generating Plant. in May.1970,.,

i as a part of the Northerri States Power Canpany system is a vital necessity.

h following discussion will highlight the factors supporting this state-ment.

NSP's latest systm planning studies show that the sumer mav4== de-mand for the NSP system in 1970 will be 2,944 megawatts. 'Ihis forecast considers the previous trend in systen growth, adjusted for unusual weather conditions. h trend curve is attached as Exhibit A.

The estimated maxi-I I

mum demand for 1970 has been revised upward from the projected 2,821 m shown on page 5 of NSP's application for license and construction permit:

(Docket 50-263) transmitted August 1,1966.

In addition to meeting the maximum demand on the NSP system, the empany also is obligated by its contractual agreements with its pcrer paoling associates to provide a reserve capacity of 12 percent.- Conse-quently, in 1970 NSP will have a total capacity obligation of 3,297 m.

1 Ctmpletion in 1968 of a 600-aw fossil plant now under construeden will bring NSP's total generating capability to 2,758 w.

Subtracting this frun the 1970 obligation of 3,297 m will. require additional gener-

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ating capacity in 1970 of 539 m.

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'.4 In arriving at its generating requirments, NSP is obligated to con-j sider the generating and transmission plans of all of the neighboring i;

utilities with which it is associated in the Upper Mississippi Valley Power' r

Pool and in the Mid-Cbntinent Area Power Planners (MAPP). 'Jhe requirunents of these many power producing organizations located throughout upper mid-l continental thited States are factored into NSP's plans for fdture gener-i I

ationt and the plans of all of the members of the power pool are contingent upon actual performance by each member in accordance with these plans. A

' delay in the construction of NSP's 1970 generating unit, the Mxiticello Plant, would affect electric pcraer producers throughout a large portion of the Upper Midwest region. NSP will be purchasing 225 nw of capacity in 1969, and if the 539 ma of additional capacity is not available on NSP's system in 1970, the only recourse will be to purchase the additional capa-I city. However, aside from questions of econanics, there is evidence that i

this large anount of capacity will not be available frtrn the other members t

l of the power pooling organizations or frcrn any other source on a firm basis in 1970.

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l In addition to providing the source of the needed capacity to the NSP 1

system, and to the systems of its neighbors, the Monticello unit in 1970 will add to the reliability of NSP's system because of its strategic lo-cation with reference to other generating units in the NSP area. 'lhis relationship is shown in Exhibits B and C, which also indicate the ties between the NSP systan and adjacent utilities serv.ing the metropolitan centers to the east and south of the Twin Cities.

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When the decision was made by NSP early in 1966 to construct the M:ntiy..:

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4 cello unit as a nuclear plant, preliminary site discussions were held withr w jf4 t yi agn[d$

members of the staff of the Division of R3 actor Licensing of the Atomic Energy Chmtission, and at that ' tire a tentative schedule was devel I

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called for a construction pe' unit to be issued bh May 1967... During meetings f*JM.I e

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with the staff, they indic5ted Ehat.this appea$ed to be a reascnable sMME ^[:

aq a-if the Facility Description and Safety Analysis 5% oculd be subnitted

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by August 1966. Such a suhnittal was made, and NSP's. plans are proceeding on the basis that a construction permit will be issued by May 1967 and that the unit will be aviil:ble for catrercial operation May 1,1970.

From the outset, all parties concerned with the developt of the Manticello site for a nuclear generating plant realized that site assembly of the reactor vessel is inherent in the developt, because geographic considerations make it infeasible to transport a shop-fabricated vessel to the site. Studies by General Electric Cbnpany indicated that a site.

i assembled vessel was practicable, and designs, material proctnunent, and fabrication planning are proceeding on the basis of a site-assembled vessel.

j The plant in-operation date of May 1, 1970, and the corresponding can-I struction permit receipt in May 1967, are predicated upan the site assembly of the vessel, because the point of no return has been passed for the procure: tent of a shop-assenbled vessel for use at another site. In other words, NSP's generating unit scheduled for 1970 operation cannot be a

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l nuclear plant unless the construction permit is received by May 1967 based on the use of a site-assembled reactor vessel.

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i The significant points in the o,nstruction schedule developed by the

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4 plant constructor, (bneral Electric Q:mpany, are as follows:

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Septraber 1_966' Ocznplete major excavation Dece< ber 1966 q' ~

April l'967[ g; Begin foundation construction Begin aanstruction of containes August 1967 Begin asscrnbly of reactor vessel September 1967 Ctruplete construction of cantalment February 1968 Ornplete construdion of reactor vessel Ibbruary 1959 Cbmplete basic building construction June 1969 Ornplete essential nechanical and November 1969 electrical systems Ctriplete reactor assembly January 1970 Cbnplete erection of turbine generator February 1970 Cbmplete systems testing February 1970 Initial criticality February 1970 Omnercial operation May 1970 In order to achieve the proposed constructim schedulo, it is essential that the ennstruction permit be received by May 1967 at the latest NSP g

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-.w propossthat the following schedule of suhnittals and neetings be followed prior to my 1967:

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s mceive questions fmn AEC on non-vessel November 15, 1966 ?

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Ibrward to AEC hxt:ent #2 which will catain 4 Novernber 21( lh66 - ---

detailed infonnation cn site assanbly of reactor.

s-vessel Beceive questions frcIn AEC cn A%t #2 Decanber 12, 19667 Suhnit replies to all questions frctn AEC W%r 30,1966 met with DE Staff and ACRS Subcor:mittee January 9, 1967 Send DE Staff report to ACRS January 13, 1967 Meet with ACRS Ibbruary 7, 1967 Meet with ACRS a second time, if required Mardi 1,1967 Hold public hearing April 10-11, 1967 Beceive construction pennit my 15,1967 or earlier NSP believes the pro r " chedule is workable and realistic. 'Ihe in-operation date of the Manticello Nuclear Generating Plant depends upon adherence to such a schedule and is vital to the welfare of the NSP system, its custruers, and the systans avi custmers of adjacent utilities.

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.{' l l ~ y La r i = 4 . ? ? V 1 a Docket No. 50-263 i I 5 is 1 .l Northern States Power Company 414 Nicollet Avenue Minneapolis, Minnesota 53401 Attention: D. F. McElroy Vice President, Enginee ring Gentlemen: On October 11 and 12, 1966, representatives of Northern States Power Company met with the staff of the Division of l Reactor Licensing to discuss your application for a Construction Permit and Facility License for a nuclear power plant at a site near Monticello, Minnesota. During the recting, it becare evident that channes to the engineered safeguards, as described in Facility Description; and Safety Analysis Report, were being developed. To assure an accurate understanding of the engineered safeguards, as well as related systens and analyses, please provide answers to the questions listed in the attachment. The staff will be available to discuss and amplify the meaning of any of these questions, should this be necessary. A supplement to the original application on the subject of field erection of reactor pressure vessels was recently sub-mitted and is currently being reviewed. Staff questions on. the supplement are not included in this letter. Sincerely yours, g r iv Peter A. Morris, Director i Division of Reactor Licensing i 1 .w

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(( List of questions .,a t e t a. 4:6 n 4' c. a t 4

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j-i REQUEST NO. 1 FOR INFORMATTON ON NORTilLRN ST/.TES PO'..'ER CO'IPANY !!ONTICELLO NUCLEAR U' I T NO. I 1.0 Accident Analysis Sections 5.1.3.1, 5.1.3.2,13.3.4.2, and 13.3.4.4 of the Preliminary Safety and Analysis Report should be supplemented. The following is required: 1.1 A technical description of the analytical blowdown model. 1.2 An assessment of the adequacy of the analytical blowdown model referenced in section 5.1.3.1 and the basis for sane. 1.3 A presentation in graphical form, as a function of tire after MCA, of the following information along with progran input assumptions, justification where necessary of these values, identification of operating engineered safenuards and the error band on the calculated values. 1.3.1 Coolant mass in reactor vessel (RV). j 1.3.2 Coolant mass expelled fron RV. (Identify flow to recirculation i pumps and steam to turbines until isolation valves are fully closed.) 1.3.3 Coolant nass introduced to RV. 1.3.4 Energy stored in core, reactor vessel, and internal structures. / 1.3.5 Energy in reactor vessel coolant. 1.3.6 Energy in coolant expelled fron RV. 1.3.7 Energy added to RV internals. { 1.3.8 RV injection water t enperature. 1.3.9 RV injection water pressure. i 1.3.10 Itass flow rates throup,h the core. 1.3.11 Reactor vessel core exit plenum pressure. 1.3.12 RV core entrance plenum pressure. 1.3.13 Pressure drop across bottom of fuel bundle shrouds. . it II 1.3.14 Reactor vessel water level. ,1: i' 1.3.15 Reactor vessel water temperature. 1 ji a 1 i k'_h

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~ . - ( %.*_. I t a - i 1.3.16 Maxirum fuel clad temperature. I 1.3.17 Maximum fuel (00 ) temperature. 3 1.3.18 Drywell atmospheric temperature. w. 1.3.19 Drywell pressure. t 1.3.20 Drywell steel vessel temperature. 1.3.21 Drywell containment vessel-to-concrete gap at critical locations or relative expansion of containment vessel due to pressure and temperature increases. l 1.3.22 Suppression pool water temperature. 1.3.23 Torus vapor space pressure and vapor content. 1.3.24 Peak and average blowdown ~ forces over the core cross-sectior.. Specify symmetry or degree of assynetry due to one of two i recirculation loops double ended rupture wit. resultant effect on 10 of the 20 jet pumps. l 1.3.25 Blowdown forces acting on jet pumps or other internal vessel structures such as jet pump seal annular ring. 1.3.26 Energy absorbed by containnent spray (if cperating). 1.3.27 Containment spray flow rate, i j 1.3.28 Recirculation pump speed -- identify the beginning of cavitation. I l 1.3.29 Core reactivity. (Stress the consequences of failure of rods j to scram on the reactivity balance throughout the post MCA period and the core flooding stage and the ef fectiveness, if any, of the liquid poison injection system.) l 1.3.30 Minimum fuel pellet-to-clad gap. 1.3.31 Maximum internal fuel tube gas pressure. 1.3.32 Maximum clad stress. 1.3.33 Allowabic clad stress at the location of peak stress. 1.3.34 Gap conductivity (fuel to clad). l 1 l' 1.3.35 Clad heat transfer coefficient to coolant. H + P 1.3.36 Amount of metal-water reaction. ~ i. w" * ] L + h I j + l- .;j

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t p f Continue the analysis in time until the fuel temperature is completely under control as reflected by a permanent reduction ! in temperature to a stabic. safe value.- Show significant changes' which occur during post-MCA recovery, for example, balanced containment pressure resulting from vacuum breaker action, switching from condensate storage water to suppression pool ? water, initiation of suppression pool cooling, initiation of 4 core cooling spray, containment spray, etc. 1.4 A comparison of accident recovery conditions for the above accident with and without high pressure coolant injection. 1.5 A comparison of the accident recovery conditions for the above accident if containment spray coolers are inoperative (No suppression pool cooling). What is the elapsed time following MCA until con-tainment pressure limits are reached? What is the capacity of other heat sinks? 1.6 A comparison of accident recovery conditions for the above accident with and without containment spray. What is the naxinum permissible suppression pool tenperature when the plant is pressurized and at operating temperature? 1.7 An assessment of the maxinum blowdown vibration forces and thermal transients which can be tolerated during a prinary coolant bicwdown or post MCA conditions without damage or loss of function of internal reactor vessel structures (forces on the core or fuel assemblies, fuel shrouds, spray ring spargers, steam separators, driers, jet pumps, seal ring, core supports, control rod guide tubes, etc., also reactor vessel and nozzle stresses due to thermal transients following activation of injection water, spray headers, liquid poison, feedwater, etc.). 1.8 A description of the tests, calculational method and nuncrical results to show that portions of the shroud which may have collapsed on the upper ends of withdrawn cruciform control rods will not: 1.B.1 Prevent control rod scran. 1.3.2 Cause fuel or fuel bundle shrouds to move when control rods are scrammed. i 1.8.3 Cause buckling of control rods or damage to cladding and poison pins. ~ 1.9 A numerical evaluation of the blowdown forces generated in the upper reactor vessel region following maximum steam line rupture; i.e., separators, baf fles, driers, etc. q {}. - 1.10 An analysis of delayed core cooling af ter MCA. Specifically, what drywell pressures, blowdown forces on vessel internals, etc., are e W' } m 3 s, we "6

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9 o a 2.0 Engine c red Ra f e"'Jards Discuss di f f erences in tonceit and degree of redundancy between th. safe-f "u a rd.s nroposed for this plant end those proposed for the Drese!cn and ] Dund-Citics reactors. a 1 4 > 2.1 provide the f ollowin; informatien for the contain ent spray water . S,,, I c systen. 1 1 2.1.1 nasis for and numerical values of the heat transfer coefficients. h 2.1.2 Heat rejection capability as a function of containnent spray flow g rate and suppression pnol water termerature. i~ 2.1.3 Description of tiie cor.tainnent ; pray uuter exchanner includine design pressure, heat trans fer vurf ace, tube size. pitch, enterial, l lot.tio,. clevation, cnd a cessibility (i.e., unen will t eactar . ii li r.r cr.t ry be pe rni t ted .,h at radiation t'osen "ould c.ccc mant ucq v.ch entry inta-tae buildiac!) durin us;t-?CA conditi ons. What as t h r. basis for the nur.ber c i hc.,t e xc'ian n,e rs provided? l i l 2.1.4 Sprcy wate r f ron suppree. ion pool t o :.in t ai nmen t ecclers inclu.iing J de s i c.n f l ov. rate, press n e drop, inlet tegerature..ird exi t j tenperature. i i 1 2.1.4.1 Containt.ent spray punp desien pressure, flow rate, and <! rive j notor horsepower. s l 2.1 m, m nc~ 4. .w. ! attr-fr re?a:Icn te contain-d - t y q...4-re. v.- s i 2.1.4.3 Location, eh c.aion, anc ama ssi 1;t" S t h.' a "rA tj conditions. c1 2 2.1.5 Service water to the containnent coolers includine, desinn flow rate, inlet temperature, evtt terperature, c.nc service water inlet pressure. 2.1.*i.1 punn suction protectien. 2.1.5.2 Capacity and nunber of service water nurns. Scc Figures VI-9-1 and X-5-2. (Also number required durine post "CA). 2.1.5.3 Location and elevation of service water c z's and drives and the horsepower of each purp drive. 2.1.5.4 Capacity and nurber of service water booster punps. (Also number recluired during pest MCA). 4 s. e J a... % d q{f?f~?

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2.1.5.6 Location and method of operation of valves. p'sg: 2.1.6 Provide the design basis for the containment spray rines, i.e., jj ' what is the heat absorption load and mechanism of heat transfer! s, l What are the basic considerations in specifying the spray rinr. 3 j design such as number of nozzles, volume of water, pressure, etc.? 2.2 Provide the basis for the core spray systen design, in particular. [{ L I 2.2.1 How is the required spray pressure established in relation to the post-MCA blowdown pressure? The blowdown as presently underr,tood is nearly completed (30 seconds after MCA) before nressure is low enour,h for the core spray system to deliver water to the reactor vessel. i i 2.2.2 How does the core spray coverage over the core cross section varv as the sprav rinp header pressure dif ferential above reactor vessel pressure increases? The c P is lou, increasing from near zero at the time spray action is initiated to the maxinun when i reactor vessel pressure approaches containnent atmos phe ri c conditions. 1 2.2.3 What is the selected desien specification for the nump and drive motor? Include a typical purp perfornance curve. !!ow many punps are to be provided? 2.3 Provide the design basis and specifications for other hich pressure injection systees such as the RCTC. Include location and elevation of equipment. Provide typical turbine-puna performance characteristics, pipe sizes and vessel injection nozzic locations, i 2.4 Provide an assessment of the maximum expected leakage of the enr,ineered safeguards which remove suppression pool water fron the containment in order to cool it and/or redeliver it to the reactor vessel or cont ain-ment systen. Identify each of the leahane sources, radioactivity levels, j and resultant radiation hazards. ' bat provisions are nade to dispose of such leakage? What is the maxirium amount of leahane which can he trierated? What is the storage capacity of rad waste? 2.5 Provide the design basis for low pressure reactor vessel coolant injection systers. 2.5.1 Nunber of pumps. 2.5.2 Punp capacity. I l' I f ~ J. .,e i h h di '

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2.5.5 Punp pressure differential. 2.5.4 Desi gn temperat ure. 2.5.5 Injection location. 2.5.6 Location and elevation of pumps. Provide the design criteria and specifications for energency diesel 2.6 generator power. Include: 2.6.1 Power rating. 2.6.2 Starting mechanism. 2.6.3 Loading priority. 2.6.4 Location. 2.6.5 Elevation. 2.6,6 Fuel tanks and storar.c canacity (translate to nininun pernissible fuel and equivalent full load diesel ennine escrating tine): fuel pump pe. er source, rFaracteris tics, location, etc. 2.6.7 List horsepower ratinp of the various electrical loads to be powered by the energency generator during the post-MCA period. If all the equipment cannot de operated sinultaneously after MCA, present the sequence of events and the nethods for establishinn such a series of operations. 2.6.S identity the inst runentatien and elect ricci conponents within containnent which are required to be operable in the event of a bloudown accident, h'h a t are your criteria reparding the desien qualification testing of such components and the associated wiring to ensure that such devices will not be disabled by lonn term exposure to the accident environment? Is any.on-vital (expendable) wirinq connected to circuit breakers which also f eed enqineered safenuards systens such that a short circuit within the non-vital wiring would open the circuit breakers and disable the safeguards?

.~ s e 2.7 Engineered Safeguards Instrumentation-2. 7.1 Provide a complete list of the signn!s which can initiate operation of the engineered safeguards. 2.7.2 t'rovide a list of sensors, lot.ition of same, and location of signal indicators and recorders which can be used to monitor post accident recovery to assure that engineered safeguards are performing as predicted and that any radioactive release to the t open atmosphere, buildings, river, and surrounding countryside l is not excessive. A partial list would include: l 2.7.2.1 Radiation nonitors. 2.7.2.2 Temperature sensors. 2.7.2.3 Fressure sensors. 2.7.2.4 neactor vessel water Icvel sensors. 2.7.2.5 control roa position indicators. 2.7.2.6 Containment vessel water level sensors. i i 2.7.2.7 Water flow rates. Identify the maximum primary system break for each of the enginee' red 2.8 coolant injection or spray saf eguards (or combinations of same) which could be tolerated without uncovering the core. Show R.V. pressure icyc1 and temperature as a function of time for the limiting cases. 2.9 Identify the accident range and status of engineered safeguards that would require depressurization to pcrmit core flooding before core is uncovered. Present as a function of time and delay of depressurinction for various size breaks: 2.9.1 R.V. pressure. I 2.9.2 Maximum clad temperature. I 2.9.3 Water temperature. 2.9.4 R.V. water level. l 2.10 Describe alternate sources of reactor vessel injection or spray water l if suction from the suppression pool during post MCA recovery is lost. What other reliable sources of water ere available to cool the core and how long can this cooling method continue? What drywell and suppression pool water levels can be tolerated without causing excessive pressure? L l r

(- _9 o 2.11 provide the basis for and reliability of steam isolation valve per-formance. What consideration has been given to performing tests on a steam line isolation valve prototype to denonstrate required closure time and leakage performance under postulated stean line break accident conditions considering that the isolation valves would be the only barrier between the primary system and the environment in the case of a steam line break? Consideration of these tests should include: l 2.11.1 Demonstrating closure ability against the steam-water flow 'i expected during a steam line break, s 2.11.2 Cold leakage tests af ter a hot seating, which might be the sequence after a loss-of-coolant accident within the drywell. 2.12 Resolve the following with respect to facility design: 2.12.1 On page II-6-4 a statement is made about using the spectra of Figure 11-6-5 and danping values f rom Table 11-6-3. Later in the same paragraph of Section 11-6-3.1 the statement is made, "if computerized methods of dynamic analysis are used, the mathematical model may be subjected to an excursion through the Taft earthquake of July 21, 1952 North 69 West component with an applied factor of 0.33." The statenent then goes on to indicate that the structure should be examined under values of twice those given in Fig. II-6-5 as well, or a dynamic excursion with an applied f actor of 0.66. Clarify the meaning of this factor in terms of its use in the procedure, and whether the maximum carthquake corresponds to values twice those indicated in the spectra of Fig. I1-6-5. 2.12.2 on page 11-6-5 it is noted that for Type 2 structures and equipment a minimum seismic horizontal coef ficient of 0.10 with a one-third allowabic increase in basic stress will be used in the design. State the reason for selectinc,this value and its consistency as compared to the procedures adopted for the Type I structures. Provide the basis of the response acceleration spectrum of Fig. 11-6-5. In order that we may analyze note readily the short period range of this spectrum, please provide this portion of the spectrun on an e>.panded scale or provide a logarithmic plot of the spectrum. 110w is the uncertainty con-sidered in calculation of period using a response spectrum showing a large change in acceleration response for a sna11 change in period? 2.12.3 Damping values are listed in Table 11-6-3 on page II-6-5. The damping level for reinforced concrete structures is listed as 5 percent critical. We believe that the damping value is a function of the stress level permitted either under design conditions or for safe shutdown, and this should be clarified. u d W 8 . 4

~ -b = u 2.12.4 The reactor building is described on page V-2-2 as a " steel frame with insulated metal siding" building which is over a P[; substructure of poured-in-place reinforced concrete. Describe the leak tightness of the insulated metal siding and submit descriptive information concerning the manner in which the joints J i are made in such siding and the provision for prevention of ' $i opening of cracks under earthquake response. Some indication is needed of the size of the cracks that can open and still meet the leakage requirements. 2.12.5 What provisions are taken to insure stability of cranes during an earthquake? I 2.12.6 Please provide a description of the stack, how the analysis of ) the stack is to be carried out, and whether there will be any j danger of failure of the stack causing a collision with vital elements of the plant. How of ten and to what extent will 4 l stack be inspected? l 2.12.7 Some clarification of the treatment of Class I equipment contained within Class 11 structures is necessary. Are there any { 4 such items required for safe shutdown? If so, how is the analysis handled for these items and how is their response af fected by the response of the Class II s tructures to which 's they may be attached? 2.12.8 Provide a description of the inspection procedures to be followed for the construction of the containment structures or other critical structures, and indicate the responsibility for quality control. l 2.12.9 State the thickness of the drywell and suppression pool containment vessel steel. 2.12.10 State the thickness and steel reinforcement requirements, if any, for the biological shield surrounding the containment vessel. ll 2.13 Provide a numerical evaluation of the containment vessel and engineered safeguards protection from internal and external missiles. Include assumptions and input information as well as a description of the calculational techniques employed. Express for a representative range of missile sizes the striking energy, location, velocity, and angle of approach required to penetrate the containment vessel and relate these values to missiles which could conceivably be generated within the containment during MCA or on the exterior during tornadoes or as a result of turbine failure. 2.14 With respect to the two reactor building to suppression pool 20 inch } vacuum breakers, state: ) O .Yo t Y 4 y.

(. 1 2.14.1 Their decinn function. 2.14.2 Design criteria used to determine number required, size, leakage, re-seating reliability, etc. 2.14.3 Location in the building, 2.14.4 Vacuum breaker activation pressure differential. 2.14.5 Operational characteristics of the vacuum breakers, i.e., how does the valve operate mechanically in relation to sensing devices etc.? t 5 ? .4 ph e ka 4 (

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I p, i . 3.0 Site Analysis j The following information should be supplied: i rf 3.1 A statement and evaluation, based on the 1980 population projections, 'C l, i of the low population distance for this site. sd ' l ^?: t 3.2 A description and evaluation of the anticipated meteorological j '.I ' program,to be initiated at the Monticello site. i ; l 3.3 An analysis showing the minimum dilution to be expected between-the it l l condenser discharge outfall and the intakes for the nearest public drinking water supplies for both an accidental slug release and - continuous release associated with (1) normal, and (2) low flow in ~ the river. State the maximum amount of liquid radioactive vaste which will be stored in the various on-site containers in relation to the capacity of these same containers. l 3.4 Data on storage, capacity and an estimate of the length of time j withdrawal of drinking water can be suspended for the municipal water supplies down river from Monticello. Provide data on the j water storage capacity and use rate of population centers within j 50 miles downstream of the site. 1 i I 3.5 Preliminary procedures f or the operation of the cooling towers. Can the failure of the cooling 'owers involve safety considerations for the nuclear plant? 3.5.1 Describe how the operating procedures and design of the liquid rad-waste system take into account the operation of the cooling towers in meeting the requirements of 10 CFR 20, 3.5.2 What are the maximum radioactivity concentrations that will be released to the river associated 61th operation of the cooling towers? 3.5.3 What is the capacity of the cooling tower basin' 3.5.4 What is the circulating water arrangement at the cooling tower in relation to river water coolant? f 3 1 3.5.5 What is the transport time of the circulation water from the i condenser through the cooling towers and condenser? I j 3.5.6 Do the main circulating pumps discharge to the top of the cool-i ; o i ing towers or are there other lift pumps? - g l 3. 5. 7_ llow many induced draft blowers are required and what is the 1 7,,- horsepower of each? b g, . f f .g sY -n 'b 4 5

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g. ^T di:. charge l 3.6 The justification for the estimated liquid rndleactivity rat es of I re/ day normally and 250 mc/ day wit'i fuel leafs. ) 3.7 Design criteria and description of the reactor building st andby gas Provide the following information: treatnent system. J in y 3.7.1 The type of filter media or charcoal and amount of charcoci each unit. 3.7.2 Ignition temperature. 3.7.3 Ded depth or filter thickness. 3.7.4 The naximum curie inventory that could be trapped on the filters. 3.7.5 Description of the f rane and shicidinn. 3.7.6 Testing procedures and frequency. at the filters 3.7.7 The acrospheric conditions that would be present as a result of the occurrence of the MCA. 3.7.8 Stack exhaust velocitics. Nunber of blowers and filters in service during post-MCA. 3.7.9 3.7.10 C%r< oal t erpe t at are at. a :;r tI n of fed or fnvectorv vith blescrs 11. I 3.7.11 L' hat provision 3 will be n..fc : n c.su nrub

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M J t: es I occur? i 3.7.12 Control room dose and ventilation systen description. 3.8 1he detailed calculation, described in Section 5.'.?.2 to estimate exfiltration rates in high winds, includinn justification of assumptions. Ue do not believe that the nethod proposed to test the secondary containment is adequate to show that there is no out-leakage at any point in the buildinn during low wind conditions. Please discuss the nessurenent of nressure dif ferentials at points on the external walls as a nethod for initialle det ernining that ll the in-leakage specification is met. In addition, provide the test procedure planned to demonstrate periodically the lenk-tinhtness l-4: of the secondary containment. the instrumentation te be used in such meast terents and the frequency of testina. .lus t i f y the suita-j7 bility of these tests to demonstrate that a 0.25 inch of water will exist at any location of the entire peripheral area of the secondary d; containnent structure. N

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to its integrity ac.ainst the cifects of a severe tornado, including 2 the capability to withstand rapid depressurization. Discuss the 9I potential damage due to tornado induced missiles, including the velocities needed for various types of Inissiles to penetrate the secondary containnent building and to damage the various comnonents z and equipnent necessarv for the safe shutdown of the reactor plant. c Include the justification for the assurptions nade in this discussion. 3 .g 4.0 "ower Plant 4.1 i'xplain why spare condensate and feedwater pumps (compared to previous ir.!R plants) are not planned for this facility. 4.2 Explain the increase in the boron concentration (50 0 ppm vs. 3800 ppm for Dund-cities) in the temporary control curtains. What experience uarrants this chane,e? Discuss integrity over the pro-pesed irradiation period. 4.3 Describe tue work which assures that cavitation in the ict pumps will cause no adverse vibrational ef f ects durinr, nornal or abnormal operation such as loss of feedsater. 4.4 Describe the methods used to determine the individual ennineered hot channel factors. What is the accuracy or error hand? !!nw were these values used in the core ther al performance calculations? 4.5 Section 11.2.2 of the Facility Description and Safety Analysis Report states that the main condenser will acconnodate a 1 W load rejection. Section 11.'.3 recognizes that the bypass valves will pass up to 15% of the throttle steam directiv to the nain con-denser, and that the conbined capacity of the bypass valves (three 57 capacity valves) and relief valves is sufficient to keep the reactor safety valves from openine in the event of a sudden loss of full load on the turbine generator. The increase in neutron f lux, however, according to section 11.3.1 causes a reactor scran. Nith reference to the forecoing, please provide the follow-inn infermation. 4.5.1 When the plant is in the automatic rode of o,eration using variable speed recirculation pumps, what is the maxieun step change in load that can be tolerated without scram and what cffeet does the 157 bypass capabi1ity have? Show reactor power level, pressure, tecirculation flow, bypass steam flow, as a f unction of time af ter step power reductions. 4.5.2 What is the step power reduction capability without scram when pump speed remains constant? What are the response characteristics of the turbine bypass valve 7 6 ','j >V j $ d' z l 4.5.3 % E gQ' j - [ 3 x;gn. i _ _ [ 5.' A l e ?{ gV W.,, UW t,,,k l 1 a ~ ~ 'l - b G; % s , f j f f Q f & $ % ? h ;f' [~ Y T3 % .;y.GQggf ~'

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.:tu 4 e - # M' %, Aa;r ,y cv t.44 4.5.4 Provide powcr distribution evaluation of the Northern Staten ; Q .k MF Power Company network assuming large permanent load losses occur, possibly causing reactor scram, to show that a scram ' f the La; o reactor can be tolerated without causing a complete electric ,f$ f ailure at-the site; i.e., identify the other power plants l& ~i on the grid and their respective response characteristics to 'k a 472 Kle loss of generating capacity as a function of time af ter the initial step load reduction (the load reduction which initially causes the reactor scram). Included should be: 4.5.4.1 Frequency. 4.5.4.2 Voltage. 4.5.4.3 Power. Transmission line and breaker locations and ratince should he Identified where necessary for an understandinc of the evaluation. 4.6 Uith regard to the core tieermal analysis and consi.'ering the ef fect of flow distri autions produced by orificing, what will be the maximun exit quality in the hot tes t channel? Identify the ?'CilFR of 1.5 with regard to quali ty, flow rate. and location in the core. What confidence is there that this ratio will not be Icssened by axial power distributions other than the reference dist ribution considered in your analysis? If the ?tCilFR of 1.5 is reached, what will be the power nargin to reach the critical heat flux? 4.7 Provide an analysis of the buildup of tritiun in the prirary coolant over the life of the plant. Consider such sources of t ritium as diffusion of fission product tritium through the cladding, i activation of additives or irpurities in the urinary coolant, if any, neutron reactions with boren and uhotonuclear reactions.

r. valuate the hazard f rem tritium inventerv in the prirary coolant in terms of a stear line rupture.

2:b at teans of tritiun nonitoring will be provided to ensure that excessive concentrations are not I reached in the reactor coolant or rad waste systcr? 4.8 Provide the excess reactivity for the following cenditions in relation to nuclear design data in Table 1-4-1 TD&SAR: 4 4.8.1 llot critical, l i 4.8.2 Ilot full power. 4.8.3 Scram of half of the rods withdrawn during full power equilibrium xenon condition. Present the following as a function of time I .s y until a stable equilibrium condition is reached. 4.8.3.1 Net reactivity balance after half of total withdrawn rods scram. S' 6 ~ 3

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^ ,j J V c' " 4.10 Is the circuit breaker for diesel operation accessible and manually j ' operable? i 1,..p', , e. " 6 4.11 What instrumentation senses a steam line break and causes the steam Vae g-- c line isolation valves to close? jv 4.12 Show the relationship between power level and flow during natural circulation in contrast to forced circulation over the same power range. Also show average core voids and reactivity balance for this comparison. 4.13 It is stated that the plant may be controlled either by recirculation flow control or control rod movement. 4.13.1 Operation of the flow control by signals from the main turbine speed governor enables load-following operation of the plant. Is it planned that the plant be operated initially in this r.anner? If v so, what further test data or operating experience will be pfo-vided prior to plant operation to support the feasibility of automatic power control; What changes must be made in the control system and power transnission system to convert the plant between .j base-loaded operation and load-following operation? 4.13.2 Describe the operating characteristics of the recirculation pump-drive motor fluid coupling and motor generating set. 4.13.3 Describe the design of the pump seals which enable the pumps to be operated at variable speed without excessive vibrational damage. 4.13.4 During manual operation of the control rods, what will be operational status of the flow controller? t' 4.13.5 Explain the statement on page IV-2-3 FD&SAR, "one third of the flow passes out of the reactor vessel" in relation to page IV-2-5, " injection nozzle functioning at least nozzle flow itself will be injected into the bottom plenum of the vessel. This corresponds to 1/2 of the rated jet pump flow." g Q I d-1 u. } k .~ 8 . "g

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y i~ 7- , t t* b. i ' s' e" 4.13.6 Present the relationship between total jet purp flow, pressure drop across jet pump, and recirculation pump flow. 4.13.7 How are jet pump malfunctions detected? 3 - g4 ~ 4.13.8 How is total. reactor coolant flow measured? s -j 'l 4.13.9 By reducing reactor power using flow control, what will be the ~ maximum average and hot channei quality reached at the core exit? Compare a 30% chanbe in reactor power using flow control to the corresponding change in the critical heat flux ratio. What is the lowest power where flow variation vill be permitted? 4.14 How will the failure of any of the in-core neutron sensors restrict power operation? If other installed instrumentation is substituted for a f ailed sensor, what evaluations are required to show equivalence? 4.15 How is accidental control rod movenent out of the approved operating pattern detected automatically and what corrective action occurs? For example, a control rod drif ts or unintentionally moves when other selected rods are actuated while in automatic node of operation. If undetected by the operator, what would prevent automatic features from compensating for the local disturbance and

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