ML20127L018

Rept 2 to ACRS in Matter of Util Application for CP for Monticello,Mn,Nuclear Unit 1

ML20127L018
Person / Time
Site:	Monticello
Issue date:	03/22/1967
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20127L015	List: ML20127L013 ML20127L018
References
NUDOCS 9211230385
Download: ML20127L018 (32)

Text

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March 22,1967 i

--- U. 5. ATOMIC ENERGY C009t1SS10#

DIVISION OF R.'. ACTOR LICENSING 4

REPORT TO ADVISORY COMMITTEE ON RF. ACTOR SAFEGUARDS l

I IN THE MATTER OF' NOR11UtIGI STATES POWER CONFANY

_ APPLICATION FOR CGl&TRUCTION PERMIT FOR THE MONTICELLO. MINNESOTA. NUCLEAR UNIT No. 1 p rrT NO. 50.263 i

REPORT No. 2 i

( .

Note by the Director of the Division of Reactor Licensine The attached' report has been prepared by the Division of Reactor Licensina for consideration by the ACRS.at its April 1967 meet!nm.

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TABLE OF CONTENTS i

Abstract M

1.0 INTRODUCTION

g 2.0 2

NEW OR SPECIAL PLANTELLOFEATURES FOR MONTIC 2.1

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15% Turbine Bypass 3 2.2 3

i Power Rsliability and One Emergency Di

! 2.3 Cooling Towers esel Generator 5 i

2.4 7

! 3.0 Field Assembled Reactor Vessel -

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SITE CHARACTERISTICS 9 z

3.1 Site Description 4

3.2 Meteorology 3.3 Hydrology i

• 10 3.4 Environmental Monitoring 11 3.5 Geology and Seismology 12

, 4.0 ENGINEERED SAFETY SYSTEMS 12 5.0 13 6.0 REVIEW OF PREVIOUS ACRS CONCERNS ACCIDENT ANALYCIS 20 6.1 Control Rod Drop Accident 22 6.2 Refueling Accident 23 6.3 23 6.4 Steam Line Break Outside the Reacto r BuildinR 24 Loss of Coolant Inside the Drywell 7.0 25 s

ITEMS REQUIRING CONTINUING REVIEW 27

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@ ; 10C0 A L Q1 301 @ hl Abstract The Northern States Power Company hun proposed to build a single cycle boiling water reactor of 1469 Wt (designed for future operation at 1674 Wt) at Monticello, Minnesota. The plant >

similar except for size to Dresden and Quad-Cities plants, will be furnished by General Electric.

A 15% steam turbine bypass capabi!.ity to the main steam con-denser will be provided. Since this is smaller than the turbine bypass capability for previously reviewed plants, we examinsd this particular feature extensively and concluded that it will not result in unsafe operation.

One emergency diesel as proposed by NSP is contrary to our present view requiring that on-site emergency power meet the no-single-failure criterion. -Accordingly, we believe that two sources of on-site emergency power should-be provided.

Cooling towers are provided to dissipate reject heat from the main condensers when river water flow is seasonally low.

Although a new feature for CE boiling water reactor plants, it creates no safety problems.

The reactor vessel which is to be erected in the field is a "first" and because of this received a considerable evaluation effort. We concluded that the integrity of the field assembled reactor vessel is expected to be at least equivalent to that of a shop fabricated reactor vessel.

Based on infonaation' supplied by the applicant in response to our questiets- concerning the+ original Facility- Description and Safety Analf sis Report (reference 2) and aided by cral pre-sentations and field trips to reactor vessel manufacturer shops, we have concluded that t'nis facility can be built and operated at the proposed location without undue risk to the heelth and safety of the public.

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1.0 nrrRODUCTION The Northern States Power Company, Minneapolis, Minnesota, has submitted an application dated August 1, 1966, for a license and construction permit, for a single cycle boiling water reactor of 1469 MWt (designed for future operation at 1674 MWt) to be located on a 1325 acre site on the west bank of the Mississippi River in Wright County, Minnesota, approximately 3 miles northwest of Monticello, Minnesota. General Electrir will furnish the complete nuclear power plant which is sbnilar, except fe' "c.a;, to Dresden Units 2 and 3, and Quad-Cities Units 1 and 2.

The infonnation submitted by the applicant and considered in our review is listed in Section 9.0 of this report in addition to other correspondence with the applicant and ACRS. Also included in the same section is a tabulation of meetings and field trips related to the proposed !!onticello nuclear power plant.

. Except for the plant features reviewed in Section 2.0 of this report, the Monticello plant is similar to other BWR plants that have been eveluated by us and therefore a comprehensive review in this report of all safety aspects and engineered safety systems would be repetitive. Accordingly, in Section 4.0, we have prepared a comparison of safaty related items to illustrate this similarity.

An item of current interest is the new GE Critical Heet Flux correlation APED-5186, July 1966, which was used in designing the ' ' nuclear plants. .This new correlation was not used in the Monticello core thermal hydraclic calculations.

The Jansen and Levy Critical Heat Flux correlation reporteo in APED-3892 (April 1962) used in the thermal hydraulic calculationa for BWR's which preceded TVA was used in the design and safety calculations for the Monticello core.

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2.0 NEW OR SPECIAL PLANT FEATURES FOR MONTICELLO

) Northern States Power Company has proposed that the tionticello plant be operated with:

! (1) Turbine bypass capability to the main steam condenser limited to 15%

of rated steam flow.

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(2) One diesel driven caergency electric generator to power one of two i core spray systems and three of four LPCI sub-system pumps following l a loss-of-coolant accident.

! (3) Cooling towers to remove heat from steam condenser circulating water.

(4) Field erected reactor vessel.

2.1 15% Turbine Bypass During periods when the reactor is at hot standby or is starting up or shutting down. decay and sensible heat is removed by generating steam in the reactor core and condensing the steam in ethe ' main ' condenser. To f accomplish this, a turbine bypass system to the main condenser is provided to accoannodate up to 15% of rated full load steam. In addition, this bypass I

capability would be used to restrict overpressure transients resulting from i

sudden turbine control valve or stop valve closure. The bypass. valves are l operated on an overpressure signal from the IPR. Rapid-partial load rejection

. can be accommodated with the bypass system.

We also note that this is the smallest turbine bypass capability of any of the BWR's proposed:recently as can be observed belowt

! Vermont (proposed) - 100% turbine bypass-l TVA - 25% turbine bypass Quad-Cities -

40% turbine bypess

[ Millstone - 100% turbine bypass-l OFRCLAL USE ONLY l

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i I We have considered the significance of the reduced turbine bypass capability  ;

!- in terms of reactor safety, and we believe that such a reduction for the reasons ,

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t outlined below does not have safety implications. It is anticipated by NSP and V

CE that the Monticello plant will normally be' operated at or near rated power.

To avoid a reactor scram followingl loss of the'Monticello turbine would require

! a f ast responda's turbine bypass capability of approximately 105%. Turbine i

l bypass capacities less than this value will result in reactivity transients i

$ caused by overpressure which will cause the reactor to scram. Therefore, on the i

l premise that normal power operation'will be near rated power, the frequency of screas is not af fected by the turbine bypass capacity of 'less than 105% and the l~

i choice of 15% should, in itself, not increase the frequency of reactor scrans in contrast to 25 or 40%.,

l The mechanism of automatic transfer of emergency power following a. scram, I here, as in other nuclear power plants, has not been- considered in detail. If i

the power fails to transfer, it will'be necessary to rely on ~ 1) turbine driven

!- pumps to make up coolant lost through relief valves ~o'r 2) the-emergen :y diesel l generator power supply system. ~We intend'to consider the ~ ability of -the NSP  ;

power network to-survive the ' complete loss of'the -Monticello' power generator

!- following reactor scram and' continue to supply reliably of f-site power when we 1 again review'this facility for an operating license.

[ In response to our. questions,- the applicant informed us -(ref. 8) - that .

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turbine bypass capability of less than 105% will not preveut a reactor scram if

i- the - generator trips. The choice of bypass capacity is made by the power c.:spany.

NSP has determined that they can tolerate the very infrequent'scrans which they might encounter and has therefore specified.the~ smallest turbine bypass available l

from CE; i.e., 15%. Electromagnetic relief valves' (sised to relieve 40% _ rated

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i steam flow), chosen by NSP instead of a larger turbine bypass valve, will relieve system pressure tellenting a loss of generator scram without lif tina satety valves.

If the reactor fails to scram, following loss of the generator, as a result I

i of the overpower transient or overpressure signals the safety valves are canable of relieving 100% steam flow without excessive reactor vessel pressure. We are g satisfied, therefore, that a reduction to 15% turbine-bypass capabilite doen not l result in unsate operation.

1 2,2 Power Reliability and One Emeriency Djesel Generator

.! It is planned to deliver the electrical output of the Monticello plant to i

a 345-kv nwitchvard which will have a ring bus configuration with oostrions for

. connecting the generator output, two transmission lines and a 345-115-kv' auto-i l

i transformer as snown in Figures VIII-1-1 and 2 (ref. 2)._ . The two 345-kv trans-i mission lines will be routed separately on either side of the Mississinpi River i

to the Twin Cities metropolitan area where they will be connected to tne 345-h l loop around the metropolitan area. Three 115-kv transmission lines will be-interconnected to the-345-kv system throuah the 345-115-kv autotransformerc It is also planned to install a distribution line which will be used ;at the site ,

! during construction and retained after completion of the plant as another-source of standby power. There will be, therefore, a total of six transmission lines l:

consisting of two 345-kv and four 115-kv lines connected to the Monticello site.

The applicant feals that the six power sources for the auxiliary power .

supply (including-engineered safeguards) will be sufficient in number and of i

L such electrical and physical independence that no single probable event;could a .

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interrupt all' auxiliary power at one time. Nevertheless, the station auxiliary buses will be connected by appropriate switching to a standby diesel driven OFFHCHAL USE ONLY

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generator located at the site. This standby diesel tenerator will be shysically indenendent of any normal power source and will supply the power required to shut a

down the plant and maintain it in a safe shutdown condition in the e*ent of a total loss of normal power sources.

4 Because the consequences of the MCA are so sensitive to delays in delivering cooling water to the reactor vessel such as might be experienced in starting a diesel generator within a very narrow time band, we have requested of the dppli-cant a more intensive effort to establish the reliability of site power, We asked fort 1 An evaluation of delayed core cooling (e g., consequences of addina water to the versel 1-1/2 minutes after MCA when ruel temperatures have reached 5000*).

2- The justification for relying on automatic bus transf er of auxiliarv power to off-site sources in contrast to a normal condition of l

auxiliary power from off-site sources and the effect of this bus trans-fer action on power reliability to engineered safeguards.

3. An evaluation of the independence of each of the six power transmission lines to the Monticello siteo 4- An appraisal of off-site power for Monticello if the Twin Cities metro-politas grid fails. Can power be sunplied to Monticello without inter-i ruption on a preferred load basis from'other isolated steam or hydro-electric generators? What is the maximum expected interruntion of site power following a network failure? How does load shedding affect the

Monticello power reliability?

5: A power availability evaluation followine loss of the Monticello generator (we have inquired of NSP whether they will perform controlled OFFHCHAL USE ONLY

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@El Monticello load dropof f tests when the plant is coupleted- or rely on other methods to verify the design adequacy).

6 A correlation of off-site power availability at other existing power plants which have experienced step' load reductions.

We have been' informed that the last NSP power outage occurred in 1957 and that since that time, many improvements have been made to prevent a similar re currence. Also, by 1970 NSF expects to be interlocked in a network of such size that the loss of the 460 MW-not electric power output of the Monticello generator would represent a loss of approximately 1% of the total power beine generated at that time. 11nder these circumstances, frequency would drop by .02 cycles per second and be restored to within .01 cycle per second of normal in less than 10 seconds.

The NSP response to our oral and written questions is documented on page 4-1 through 4-5 of ref. 8 and in general relates the reliability of the Monticello of f-site power to the size of the network that Monticello is normally " locked" into. We are not satisfied that the subject of power reliability has been evaluated as comprehensively as we think it should be and have identified this as an effort which should be considered again when we review this f acility for an operating license. At the present, CE contends that the power sources at this site .are suf ficiently reliable to justify a single diesel. . Considering:our.

single-failure criterion, our present view is that another source of on-site emergency power should be provided to supplement the existing diesel generator.

2.3 Cooline Towers The use of cooling towers to remove heat from the Monticello turbine con-denser circulating water is necessary- because of low Mississippi River water flow or high water temperature at certain times of the year. 11nder normal operatinn:

conditions, the cooling towers will not be used during nine months but will be

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i used f or partial cooling during two months and total cooling for one month of the

! average year. Figure I-1-1 of ref. 2 is an artists' concept of the completed ,

j Monticello plant which shows the location of the cooling < < 1ers with respect to {

the nuclear power plant buildings. The description of the cooling towers is pro-I vided on page 3.5-1 of ref, 6 In brief, there will be two half-capacity j induced draf t cooling towers located approximately 650 feet east of the menerattna plant on ground approximately twelve feet lower than the main plant ground. level.

l- Four electrically driven booster pumps will be included to move the circulat.ing l .

I water through the towers. Electrically driven fans will move the air through I

i che cooling towers, thereby cooling the condenser circulating watero j Figures 3.5-1, 2, 3.- 4 of ref. 6 illustrate- the four basic modes of operation 3 l Undar all operating conditions, the cooling water affluent from-the plant will

! tiow through a discharga canal to rejoin the river about 1500 feet-downstream of i

l- the intake.

The service water pumps which supply water for emergency systems as well as

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l f normal service water, take their water directly from the river intate structure and in no way rely on the cooling tower complex. -

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Because the pressure on the steam side of the condenser'is lower than that of the cooling tower circulating water, any- condenser tube 'leake- would cause water l-f rom the cooling side to casa into the condensate system! thus, there is little

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! probability of cooling tower water being contaminated with reactivity as 'a result l of' condenser tube leaks. ,Since the plant liquid wastes are released in the dis-4 charge canal downstream of the cooling towers, recirculation and concentration of l

these vastes in the cooling- tower basins 'and connecting canals should not occur, Our evaluation of the release of radioactive vastes' to the river is nresented-in -

Section 3,32 OIPMCHAL USE ONLY

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The applicant has made provisions for a continuous blowdown of the cooling towers, whenever the cooling towers are in service, to maintain the solids content of the recirculated water at an acceptable level. Makeup for evapora-tion losses only would result in excessive concentration of solids in the water, We are in agreement with the applicant that operation of the Monticello plant in any of the four modes of cooling water operation should not involve nuclear safety considerations significantly dif ferent from a plant using river or ocean water only as the heat sink.

2,4 Field Assembled Reactor Vessel We concluded in our first repcrt that the integrity of the field assembled reactor vessel is expected to be at least equivalent to that of a shop f abricated reactor vessel.

3eO SITE C11AR_ACTERISTICS_

3c l Site Description The proposed site for the Monticello Nuclear Generating Plant contains i

approximately 1.325 acres owned by Northern States Power Company and is located partially in Sherburne County (on the east bank of the Mississippi River) and partially in Wright County (on the west bank of the river). The sitt is approx 1=

mately 22 miles southeast of St. Cloud (1960 penulation 33,815) and 30 miles northwest of Minneapolis (1960 population of metropolitan area inc1 ? ding Sto Paul t approximately 1,400,000). The nearest residence is approximately J000 feet from the proposed plant location, j The area surrounding the site is rural in nature with only a few email l

villages within 15 miles of the site. Based upon 1960 census data, the population i

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within.5 miles of the proposed site is approximately 3900, with approximately.

! 4800 projected to 1980 The 1960 population within 5 to 10 miles is approximately 4

5800, with 7100 projected to 1980 Based upon the population information whfch i

[ indicates low population density out to a distance in excess of 10 miles and i-1 denser population beyond this point, we consider the low population zone to extend for a distance of 10 miles. The population center. distance is 22 miles, j the distance to the. edge- of, St. Cleedi We have reviewed the i applicant's analyses and have determined that the Part 100 guidelines for this S . .

site, with respect to the exclusion, low population, and population center

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a j distances (1600. feet, 10 miles and 22 miles, respectively), can be satisfied.

j 3.2 Meteorolorv _

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The dif fusica and storm climatology of Monticello area are typical of that to be expected in the north-central United States.

f l On the aversee, the dif fusion climatology of the north-central United States  ;

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is somewhat more f avorable than that expected for the rest of the United- States 4 due to the rather frequent frontal passames which are acconnanied by cloudiness.

and high winds. - Ground-level inversions occur with ~ rather low frequency - (approxi-l mately 30% of the time). The applicant has presented a range .of meteorological j parameters for estimating the dilution to be expected. associated with varying i

i atmospheric conditions at the site. We believe that the-diffusion parameters

! estimated by the applicant are adequately conservative for the evaluation o'f the consequences of the accidental release of radioactive asses from the plant s5 .

290-foot stack. The plume rise assumptions'made by the' applicant in the analysis-

. of the consequences of the steam line break accident are not realistic.in our 1-f opinion; however, it should be noted that even .if ground release assumottons are made for this accident, the resulting doses would still be well within the i

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j-- 10 CFR 100 guideline level. The Environmental Meteorology Branch, Institute for

!- Atmospheric Sciences, ESSA, has also reviewed the site meteorology and its l-

)' comments, which have been sent to the Committee, support our conclusions.

i j The intense frontal passages that occur in this area of the' country result l in rather high wind-velocities, intense snow storms, and occasionally' tornadoes '

(2% probability in 40 years). To take this into account,.all Class I structures l

i vill be designed: (1) to withstand a steady wind velocity of 100 mph 'with- gusts to 110 gh, (2) consistent with other GE reactors, to-assure that safe shutdown

! of the reactor can be achieved considering the effects of possible damage to these I c structures when subjected to the forces of. short term tornado loadings with winds l~~~ up to 300 mph,~and (3)' to withstand an internal pressure of seven inches of water -

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l (0.25 psi) without failure and without- pressure relief. - The reactor building wi11 j also be capable of relieving a pressure .of 1 poi in approximately five seconds. -

2 All external structures-are designed to withstand a snow load of 50 lb/ft .

! In view of the foregoing, we believe that adequate provisions have been made

! in the-design to account for severe- weather conditions.

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!. 3.3 Hydrolony i

The applicant has stated that the liquid- rad waste system will be capable of maintaining releases: at 10 CFR 20 levels on 'a batch-by-batch basis. The liquid waste will be discharged into the condenser cooling water downstream of the -

F l cooling towers. The~ nearest community using.the Mississippi River as a source l of public water. supply is at Minneapolis, 34 miles downstream: from the plant.

I The' plant site natural grade level is at an elevation of 930 feet MSL. - The i

worst flood on record was estimated to have reached a stage of 916 f aet at the t

l_ site. Therefore, there does not appear to be' a flooding problem at this site.

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j Due to the cooling requirements of the plant and taking into account low l- stream flow conditions, a cooling tower will be included for this plant. The j operation of the cooling towers does not appear to be present any unusual i

operating problems except as related to release and dilution of liquid. radio-L l active waste. During minimum flow conditions in the Mississippi River, the cooling tower will be operated on a closed' cycle with only its blowdown and service discharge going to the condenser discharge canal. At this tims, only

} a small volums of water is available for the dilution of liquid wastes in the discharge canal. The applicant-is aware of this potential problem, which will i

occur very infrequently, and he has stated that .he will be capable of operating -

l- under these restrictive conditions and still be ' capable of meeting the require-monts set forth in 10 CFR 20, 3.4 Environnantal..Monitorina

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l An environmental monitoring program will be initiated two years prior to i

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the scheduled completion of the Monticello plant to provide information on the l background radioactivity levels. . Samples will include air, river water, well i

ll water, soil,' vegetution, and milk. We believe that the proposed program will 4

provide an adequate basis for evaluating-the effects of reactor operations on i

the environusat.- The comments of Finh; and Wildlife Service concerning environ-

[ mental monitoring have been sent to the Committee. -

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! 3.5 Geoloav and Seismoloav l .

l_ According to the information_provided by_the. applicant,-the site is under-

. lain to-a depth-'of 'about 50 feet by unconsolidated alluvium and drift consisting

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l predominantly of clay, silt,' sand, and gravel that is saturated with water.

This material is underlain _by approximately 10_to 15 feet of mandstone which l overlies a granitic formation.

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@i10C0AL U$$ @El The applicant proposes to support the reactor facility on a layer of com-4 pacted fill using a spread or mat type foundation presumably designed to provide i a soil bearing pressure within the supporting capacity of the -fill.

l We have discussed the stability of the proposed plant location with the

,l i USGS relative to the potential for major movement or sliding of the nuclear.

plant towards the river due to undercutting or erosion of the river bank. They j' believe that this problem is prevented by the intake structure which is founded on bedrock along the riverbank. In other respects, there does not appear to be

, any unusual sagineering problems in founding the plant in compacted fill.

l The applicant has stated that no known seismic activity has originated in the area of the site. Based upon their evaluation of the seismic histoty of the i

site area, the applicant has proposed that the design of the structures and i

l equipment important to the plant safety features be based on a around motion due to an acceleration of 0.06g and, in addition, no loss of function during a ground I acceleration 0.12g. The USC&GS believes that these proposed around accelera-tions are adequately conservative for this site. We hope to have formal comments I

( from our geologic, seismic, and seismic desian consultants by the ACRS Subcommittee meeting.

l 4.0 ENGINEERED SAFETY SYSTEMS i- Previous reports have provided detailed evaluations of each of the engi-neered safety systems for the General Electric boiling water reactors. In this type of plant, the initial pressure increase following release of primary coolant energy to the drywell is limited by releasing the vapor to the suppres-sion pool where the energy is transferred by steam condensation to the cold water. Residual and decay heat from the core is removed by core spray water and reactor vessel flooding water pumped from the suppression pool.

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@lv0C0AL U$$ @MLi A very brief comparative tabulation (Table 1) of general plant information f precedes the comparison of Monticello and Quad-Cities engineered safety systems

! (Table 2) in order to illustrate the similarity of these two BWR's.

i TABLE 1 CENERAL REACTOR CHARACTERISTICS FOR MONTICELLO AND OUAD-CITIES -

Items cuad-Cities Monticello i

Reference ' design thermal dutput - MWt 2300* 1469 Operating pressure psig 1000 1000

$ Steam flow - full power 1bs/hr 8 62 x 10 6 5.86 x 10 6 Racir. flow rate at full power Ibs/hr 98. x 10 6 57.2 x 10 6 Average power density KW/ liter at 100% P 36.7 35.7 2

Heat transfer surface area - ft 63,527 42,469 Max. overpower (120%) heat flux BTU /hr it2 418,700 408,000 Max, overpower UO2 temperature F 5080 5080

100% power heat flux BTU /hr ft2 349,000 339,000 i

Max. UO2 temperature at 100% power 3725 3710*F Core subcooling BTU /lb 20e3 20. 1 Equivalent core diameter inches 182 149, Active fuel length inches 144 144 Number cruciform control rods 177 121

- Number temporary control curtains 324 '216 Excess reactivity uncontrolled at 68'F 22 6 k 22 A k Worth of control rods 17 A k 17 0 k Worth of borated control curtains 09 A L 09 0 k

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@vf0C0AL U$E @Ll TABLE 1_(continued) 1 Items _

,Oju-Cities_

Monticello Total worth of control ,26 6 k 26 6 k Reactivity of core with all control rods IN 96 A k aff .96 d k eff Reactivity of core with strongest control 99 6 k eff 99 6 k eff

- rod otTr

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Approximate reactivity coefficients Same for both plants i

Two reactors, each rated at 2300 MWt 4

) TABLE 2 ENGINEERED SAFETY SYSTEM CHARACTERISTICS FOR MONTICELLO AND OUAD-CITIES CONTAINMENT i

Ouad-Cities Monticello Material SA-212 or 201 SA-212 or 201 plates manu- plates manu-factured to factured to A-300 A-300 requirements requirements Dimensions diameter lower spherical section - f t 66 62 diameter upper cylindrical portion - f t 37 33 over-all ht - ft 112 105 torus major diameter - ft 109 98 torus cross sectional uiameter - ft 30 30 3

drywell free volume - ft 158,000 134,000 torus, suppression pool, tree volume - ft 3 119,000 99,000

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TABLE 2 (continued)

Quad-Cities _

Monticello Dimensions suppression pool water volume - ft 3 106,000 75,000 design pressure - peig 62 56 design temperature *F 281 281 max. accident pressure - drywell psig 39 33 max, accident pressure - torus psig 21 21 initial suppression chamber temp. rise 'F 50 35-drywell free volume / primary system vol. 7.6 9.74 primar" system volume / pressure supp.

water volume 0.197 0.184 primary break area / total vent area 0.0 19 0.019 RESIDUAL HEAT REMOVAL SYSTEM Low Pressure Iniection Pumps total number 4 4 number required -3 3 Pump Iniection Characteristics reactor vessel cressure poix 0 200 20 275 finw each- gpm - 5,350 2.675 4,000 2,000 flow total Apm 16,000 8,000 12,000 -6,000 head- f t 263 565 415 700 power each - hp . 600 490 580 -525 power total- hp- 1,800 1,470- 1,740-- 1,575 NPSB available - f t 33 42 25 35

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@310C0AL 03$2 @N s1 TABLE 2 (continued) g et Exchanters Ouad-Cities Monticello number 2 2 i

- heat load each-BTU /br 102 x 10 6 57.5 x 10 6 primary flow - gpm 10,700 8,000 secondary flow- gpm 7,000 7.000 primary inlet temperature *F 165 165 i

secondary inlet temperature *F 95 85 shell design pressure- pain 375 375 tube design pressure- psig 375 375 Service Water Pumps number 4 4 capacity esc' - gym 3,500 3,500 l head- f t 435 675 l

j containment sprav ilow - rpm 7.000 4.000

' 200 suppression ecol spray flow - gpm 350

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minimum break size for svatem alone ft 2 0,4 0.4 u1GH PRESSURE _ COOL. ANT INJECTION Turbine l steam pressure inlet -psig 1125 1'25 1125 155 exhaust-psig 70 65 10 65 steam temocrature *F 558 360 558 360 speed - rpm 4000 2000 -4000 2000 t;ower -no 5000 1000 3000 500 i

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@u 70C0AL U$E @GLf 3 TABLE 2 (continued) i Ouad-Cities Monticello no, stages 2 2~

emergency starting - esconds 25 25 steam flow pounds /hr 125,000 to 6600 70,000 to 45,000 Pump discharge pressure peig 1165 to 150 1165 to 160

) flow - gem 5600 2750 NPSP - f t 25 25 max, bottom break size for 2

system alone -

ft 0.16 0.0L 1

REACTOR CORE ISOLATION COOLING operating pressure - psig 1100 to 150 1100 to 150 l

number of turbine-pure units 1 1 turbine power output- hp 350 350 flow rate -gpm 350 2 3')

water used in 8 hrs of operation- zal. 90,000 60,000 CORE SPRAY SYSTEM Pumps number 2 2 speed- rpm 3600 3600 flow. gpm 4700 3020 head - f t 580 708 power- hp 860 830 NSPh (availablek ft 34 30

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Sprav Headers l

Guad-Cieles Montinello number 2' 2 nozzles / header 130 204 i nossle diameter 1" 3/4" l minimum bottom break size. for 2 i system alone - ft ,15 .15 1

j FEEDWATER SYSTEM 1

Standbv Coolant Supply System number of pumps 2 of_3 2 of 2 flow rate into vessel- tem 15,000 13,000 POWER SOURCES Elaetric Pruer

- 125 volt dece power Itatterv for

control and heavy duty buses -- ampere hrs. 912 912 of f-site power transmission lines- kv - 4-345 2-345

! 1-138 4-115 Diesels number required 2 of 3 'l size - KVA not known 3125 basic engineered safeguard load 3 of 4 RRR 3 of 4 RHR i pumps- nunna t 1 of 2. core- 1 of 2 core i sprav puans serav eumos shutdown systems for other unit i

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le Turbine generator orientation with respect to the containment building I

i is identical to Qued-Cities; i.e., 'the turbine axis is perpendicular to a line from the turbine to the center of the containment building. The containment i

l vessel is completely enclosed by a reinforced concrete structure havina a thick =

i ~

! ness of 4' to 6 feet' which in addition to serving as the basic bioloeical shield 1

for the reactor system also provides a major mechanical barrier for the protec-tion.of the containment vessel and reactor system against missiles generated external to the' primary containment. We stated in the Ouad-Cities ACRS report

! dated November 21, 1966, that if proper consideration is given to protection of-l the primary containment and reactor shutdown systems from low trajectory missiles

we could see no advantage in re-orienting the turbine on an axis directed away I

j from the containment. There is no new information which would change this con-clusion.

l 2 The primary system surveillance program-deacribed by the -aoplicant.

l includes vessel irradiation surveillance samples, and -like the Quad-Cities pro-c.

t .

! gram, portions of the reactor vessel and primary piping thermal. insulation will

! be removable to parait visual inspection and nondestructive testine of external -

l l

surfaces. One or a combination of visual, ultrasonic, mannetic particle, or

liquid penetrant examination methods will be used. The reactor vessel internal j surf aces and internal structures .can be inspected under water by use of remotely

! operable lights, visual aids such as boroscopes and underwater TV cameras, and f

l remotely placed' ultrasonic transducers; however,J complete internal inspection i

j apparently cannot be accomplished without removal of fuel,

~

The areas listed in Table 3 will be inspected periodically under a program l essentially the sama as that for the Dresden and Quad-Cities plants.

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i u ba TABLE 3 PRIMARY SYSTEM IN-SERVICE INSPECTION Resetor Vessel

Top head Studs and nuts Full penetration nozzles over 2-inch pipe size Highest stressed circumferential weld One longitudinal weld Portions of highest stressed support skirt f

Reactor Internals -

j Shroud and shroud belts Core plate Dryers and separators Relief and Safety valves JLettreulation System Selected valve and pump piping connections over 2-inch pipe ' size 4

Selected piping fittings and connections over 2-inch pipe size Supports and hangers S te am _, Feedvater Core Soray- RCICi HPCI, LCPI, CRDH Svetems Selected valve and pump piping connections over 2-inch pipe size Selected piping fittings and hiah stresses areas over 2-inch pipe size Supports and hangers 3, The possibility of coolant flow blockage due to f uel clad swelling or rupture has not been considered in cetail, although in response to a staf f request, we expect the applicant to discuss this witn the Conanittee.

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@ U0C0AL j $2 @NLJ 4, As a result of staff and ACRS concern, a program to test steam line isolation valves under simulated accident conditions is being developed. This

] proposed test program is in addition to the tests conducted at the manufacturer's plant and will be performed during a plant startup under flow and pressure con-ditions simulating a main steam line break. Isolation valves of the same design as the Monticello isolation valves will be used. This commitment represents a a

chan ge , Previously, in the December 1966 Quad-Cities report, we noted that no prototype tests with steam line break conditions were proposado 6.0 4CCIDENT ANALYSIS The applicant has evaluated four basic accidents with a potential for releasing fission products.

(1) Control rod drop (2) Refueling accident (3) Steam line break outside the reactor building

! (4) Loss of coolant inside the drywell 3 The significant considerations and results of each, consistent with the December 1966 Quad-Cities ACRS report are revituad below. Potential off-site j doses for each of these accidents are given in Table 4 TABLE 4

! POTENTIAL OFF-SITE DOSES (REM)

Course of Accident Accident Two Hours at 0-3 Mile at 10 Miles

! Thyroid Whole Body Thyroid Whole Body 1

l Control rod drop 100 9 17 1 I Re f ueling* 8 (1 14 (1 i Steam line break 17 (l (1 (1 l Loss of coolant

• 9 (1 100 1

• Standby gas treatment halogen removal efficiency of 90% assumedo

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- 23-L 631 Control Rod Droo Accident A rod worth minimiser will limit -the maximum rod worth to- .025 Mk and a

}

rod velocity limiter will limit rod drop velocity to 5 ft/see maximum. The two

, limiting cases considered were hot and cold standby.- -Hot standby is considered to be more severe ~ than a rod drop while at power because at hot' standby the mechanical vacuum pump for condenser vacuum may be in operation which provides I

j a more direct release path since the vacuum pump exhaust is not automatically

! isolated by high radiation as is the air ejector exhaust.

l For the hot standby case with an initial temperature of 547*F and a fission power level of 10 of rated power, the power transient energy was calculated' to be 4000 W seconds, causina approximately 330. fuel rods to have enthalples i

- in excess of 170 cals/gs, the threshold of fuel cladding-damate. The maximum i

! enthalpy was calculated to be 220.cals/gm which is at the UO 2 melting threshold i

i of 220 to 280 21/gm.

A rod drop power transient from the cold initial condition- and 10-8 og_

l' ~

[ rated power would generats 2500 W seconds with enthalpy in 200 rods in' excess -

of 170 cal /gm, The maximum UO2 enthalpy was calculated to be 250 cal /ge.

For each of these accidents, it was assumed that scram waa" initiated within

0.2 seconds af ter the neutron flux reached 120% of the full power value. - Oar l . .

calculated off-site doses from this accident are presented in Table 4 6,2 - Refueling Accident l-I -

l  :

During refueling operations ,. only the- secondary containment - buildint .is available to confine and control the release of- fission products to the atmosphere if any fission products escape from the fuel, Procedures and interlocks - prohibit the withdrawal of control rods while fuelois being added to the core.: The-

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@dQC0AL -24U$2 @El reactor core is designed so that it remains suberitical with one control rod j removed even if it is assumed that a fuel assembly-is dropped into an empty fuel space. Thus, no nuclear excursion would result.

. The maximum number of fuel rods that could be mechanically damazed by dropping i.e., the cladding on 445 fuel elements a fuel assembly into the core is 445; might be perforated, releasing fission products,. For calculation purposes, it j was conservatively assumed that- 1) the reactor fuel had an average irradiation -

i l of 1000 days at the reference design thermal output up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the.

! fuel assembly drop, and 2) a maximum of 1% of the noble nas activity is in the fuel rod plenums and a maximum of 0.5% of the halogen activity is in the . fuel 9 -.

) - .

plenues. The quantities of fission products calculated to be released from the i

failed fuel to the water under the circumstances described are 18.4 x 103 curtes noble gases- and 10.6 x 103 curies halogens.

The halogens would-be absorbed in' the water and eventually reacn an equi-1- librium with the air. Assuming 100% of the secondary containment building volume per day is -discharge rate throuah the- standby gas treatment system -(actuated auto-matica11y on high area radiation in the reactor building) to the 290-foot stack, and a 90% halogen filter efficiency, we have calcul'ated the, of f-site doses given f.

in Table 4 6.3 Steam Line Break Outside the Reactor Building i

The course and consequences- of this accident do not ' differ from that. described for Dresden Unit - 3 and Quad-Cities. We continue to disagree with GE concerning the -

buoyancy of the steam cloud as it . leaves the turbine building and-have thus per-f ormed' our own calculations of the consequences which are given in Table 4, i

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@ i 10C0 A L W $0i @El GE has analyzed the forces developed within the reactor vessel (pp 1.2-4 ref) for a 70 psi /sec initial depressurization rate. The . calculated results show that j the maximum pressure differential across the core for a steam line' break would be 11 psi less than the 41' pei value required to lift fuel bundles and would occur 3 to'4 seconds after the steam line break when flashing occurs in the lower 4

plenum, The maximum pressure differential tending to bulge the channel outward

! were calculated to be approximately 16 psi in contrast to a'25 pai~ pressure -

f differential for channel deflection sufficient to bind control rods.

The conclusion is drawn, therefore, that control rode will be scrammed

! before maximum forces are developed during the transient 'and also that the maxi-num forces are not suf ficient to prevent control rod insertion since the fuel bundle channels are not deflected appreciably. Also, the lift forces are not

< sufficient to change the core configurations 6.4 Loss of Coolant Inside the Dryvell GE has. calculated that the recirculation line break, 5.5 square feet double-1.

ended break, results _in the maximum fuel and coolant temperature and requires j the fastest response of engineered safety systems. The same enmineered safety i

systems as provided for Quad-Cities (see Table 2) are available for core and con-tainment coolina, i

j The maximum forces during the- blowdown reported on page 1.2-3 of ref. 6 are not excessive because the lower plenum cannot discharge directly to the atmosphere f

but must discharge to the downcomer through the inoperative jet pump dif fusers.

l Water flow through the ist pump nozzles to the atmosphere is " choked" in an area only 15% of the 28-inch OD recirculation line,' It is expected, therefore, that control rods will scram _ and core geometry will_not -channe following the 4

maximum credible accident, rupture of the coolant recirculation pipings

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~26-i During the first few seconds following the MCA, the homogeneous two-phase blowdown of approximately 44,000 pounds /second of water and steam from the 5

pressure vessel through the 5.5 f t 2break ~ pressurises the drywell and relieves thereafter through the vent pipes into the suppression pool. By venting the staan

]

to the suppression pool where it is condensed, the drywell pressure reaches a maximum of 33 psig 6 seconds af ter the accident and then diminishes., Since the design pressure for the containment is 56 peit, it can be seen that there is l

ample margin in this respect, and it can, therefore, be concluded that the energy stored in the primary coolant and also that energy transferred to the coolant l

! -- from the core during the blowdown transient is safely dissipated without exceeding containment design pressure or changing the core configuration to affect the l control rod scram capability and thus shutdown margins.

i

. Northern States Power is confident that of f-site electric power will be available continuously following the step loss of the 462 MWs Monticello gener-f l

stor and in this case the power supply for any- or all of the unmineered safety

! features (electrically driven emergency pumps for core spray, core floodina, and river water to remove decay heat) is more than adequate. GE has, however, examined i

i the sequence and time for startina these same electrically driven emergency cumps ._

f rom a power supply of limited size, without overloadingt in this case, one 3125 KVA diesel driven electric generator. The sequence of events and the horsepower i

recuirements for emergency coolina are reported in tabular form on pazes 2.3-13 and 2.3-14 of ref, 6.- We have concluded thac if there are no delays of coolant injection following MCA after the primary system cressure has decransed to about 300 psi (20-30 seconds after MCA),c fuel claddina temperatures will tenain below melting, relatively little fission product release will occur and _ post-

!- accident recovery procedores for the short and long term are adequate a

( ,

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@L10C0AL Q1$0i @Ll Nevertheless, we have calculated ~the potential off-site consequences of a coolant-loss accident which~ results in' a'100 percent meltdown of the core with release of 100 percent of the available noble gases, 50 percent of the halogens, and 1 percent of solids. Subsequently, 50 percent of the released halogens are assuiend to plateout, It was further assumed that the fiocion products leak from the drywell at a rate of aporoximately 0.6 of a percent per day directly out the 290 ft stack with no holdup in the reactor building. Ninety percent of halonens are assumed to be removed by the charcoal filters in the emergency ventilation system. The meteorology assumed for these cale11ations was derived by .drawini,-

an envelope enclosing the dilution curves for a 29Nt- release height and each of the Pasquill dif fusion types thus resulting in a curve which would produca the maxime n potential doses with distance. The resulting doses at the site bour.darf are shown 1.n Table 4 7.0 ITEMS REOUIRING CONTINUING REVIEW In our review, we have identified the several items listed below which will require further evaluation as the design progresses. We intend to continue our review of these items during subsequent licensing reviews of General Elactric

, boiling water reactors and as final information becomes available.

(1) Design and analysis of the engineered safety systems including redundancy of critical valves and instrumentation and analytical studies of core flooding requirements.

(2) The testing of steam line isolation valves for closure and leakage under simulated accident conditions.

(3) Of f-site power reliability evaluations and loss of Monticello generator tests to measure the ef fect on network.

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@uBC0AL U$$ @ML1 In addition to the foregoing, it is our view that another source of on-site emergency power should be provided. Furthermore, we recognise that the concerns of the Committee concerning post-accident reactor core cooling expressed in the TVA letter are common to all reactors, and this will be under our continued review.

8.0 CONCLUSION

~

On the basis of the information we have reviewed and evaluated in conjunction with the Northern States Power Company application to construct a nuclear power plant at Monticello, Minnesota, we believe that there is reasonable assurance that this f acility can be built and operated at the proposed location without

-undue risk to the health ar.4 safety of the publico

9.0 REFERENCES

9.1 Northern Sticar Power Company Submittais to DRL (1) March 17, 1965 NSF Request for Informal Safety Review of Site (2) August 1,1966 Application for License accompanied by Vclumes I and II of the. Facility Description and Safety Analysis Report (3) September 14, 1966 Amendment No. -1, Site Fabrication of Reactor Vessel (4) November'23, 1966 Amandment No 2, Design Fabrication and Erection of the Reactor Vessel (5) December 30, 1966 Amendment, No. 3, New Emergency Cooling Equipment and.Other Changes to the FD&SAR Volumes l 'and II (6) January 10, 1967 Amendment No. 4 Answers to DRL Staf f Questions of December 27, 1966 1

! (7) January 19,-1967 Amendment No. 5, Study and Evaluat1on of Operating Experience with Field Assembled Pressure Vessel' as requested by AEC, December 27, 1966.

l (C) March 8, 1967 Amendment No. 6, Answers to AEC Questions 9,2 DRL Staff Recorts to ACRS April 21,1966 Preliminary Site Evaluation.

We concluded that the Monticello -site is suitable from a population stand-

. point for a Gemaral Electric boiling water reactor assuming the usual engineered

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@kl safety systems for this type of plant are provided. (High pressure turbine driven coolant injection pumps and low pressure core flooding improveasnes have since been proposed in Amendment No, 3.)

September 2,1966 Preliminary Aspects of NSP Application We noted two prominent novel features proposed for the Monticello plant t (1) Cooling towers to remove heat from condenser circulating water, thus pro-viding a closed cycle cooling water system. The cooling towers will be used when river flow is low, (2) Field erected reactor vessel. Field erection of the vessel was necessitated because a shop fabricated vessel could not--be transported to the site, January 24, 1967 Review and Evaluation of Design, Fabrication, and Field Erection of the Monticello Reactor Vessel (ACRS Report No. 1)

We concluded that integrity of the field assembled reactor vessel will be at least equivalent to that of shop fabricated reactor vessels s W+ listed seven recommendations beyond CB&I, GE, and ASME Code requirements.

February 9,1967 Addendum to ACRS Report No.1 We relaxed our recosseendations as listed in ACRS report No.1 to the extent that these recommendations would apply to future field fabricated vessels to measure the improvements, if any, in achieving an independent certification of

^

the reactor vessel integrity (quality) based on ultrasonic and radiographic exami-nations.

The human f actor is prominent in present methods of assuring reactor vessel inte gri ty , We concluded that this was justified at this time because of the numerous ultrasonic examinations of the reactor vessel and the availability of experienced personnel who would'be assigned to this first field assembled reactor vessel,

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9.3 Staff tatters to Apolteant l 2 .

l May 19, 1966 Report on Reactor. Site for Northern States Power j Company. The Monticello location should be

suitable for a nuclear reactor of the type

! described if appropriate eneineered safety systesis are provided.  ;

December 27, 1966 Request No.1 for information on Northern States '

i Power Company Monticello Unit No. 1. Questions i related to accident analysis, engineered safety systems, site analysis, and power- plant.

February 27, 1967 Request No. 2 for additional information on Northern States Power Company Monticello Unit No 1, Questions: related to reactor vessel, power reliability, and site.-

9,4 Staf f Meetinp with the Aeolicant October 11, 1966 Reactor Vessel Field Fabrication Octobe r 13, 1966 Accidents, Enstneered Safety Systems, etc.

• December 2,1966 Review of DRL - Monticello Ouestions February 2,1967 DRL Recommendations Related to Fabrication of Reactor Vessel February 24, 1967. Geology,- Hydrology, Seismology of Site t

9.5 ACRS Heatinas with the Aeolicant February 3,1967 ACRS Subcommittee Meeting, " Design Fabrication and Erection of the Monticello Reactor Vessel,"

February 10, 1967 ACRS Meeting, " Design Fabrication and Erection of the Monticello Reactor Vessel."

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