. . - _ , . - -, _ . - - _ .. . . _ . -- . . - . _ .. . - . . .

ar*y%. %!,

. .,

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C .O O e-t eHC 4 t'a u > .fi.ee -4 0- V- *e4 4 4 O L dD 5| 0 4 0 4 .O 4 .C Wen enV d n, e oO d ': ?.: C'

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C c .-e(4 O.

,

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.

to yN s - (*. C.-

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by *4a

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C .C H O %dN 4L oO4 Vi L. 4.3 .* Op, W U O t O u ;

C 3 C D O pO F 0 4- p. O H% .C 4. H O C3 CN OOa e-4 N s O

< . .4.*dH

O .C C; < nG G H t. . O C t1.e4 U = 0t d CJ-" U4O 4 ~*

C CC OW n4 8 .C 4 O

rH to 4D -4 -4 c-4 -1 L Q.O % O e v4 C;a L1 > 0 0 4 .3 tS OC5 04 O O d C4k e-4 O P O4GL % d O M O e-4

, .4

j ?' 4 p C' N% O %4 4 % .O&.: O k % t2 O O1 4 O C C' *r' 4 's C .,C 6 5 4oe,e; d s 1.:; O4 <s O,

d U LO 4 T k4 e

5e O *O P O A. .O.,O

4 3 ct . wO

to .-* be A O Nt M,

1

to 4 a-1 O C C* '

4 uuWt C .e4 go. aO oO O k O

<IC T 4 .O O e4 O IN O . O Es n 4 d O l

A. g4 4N~

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r-4%4 OC

. t'). .

(Q O NeOw |*Q~D.L 0.

l, , . - - -. -

__

.

I I 1 1 I I I l- -- I ~I I

?!UTE31. RATED IU'4ER In 16, fGt?. DESIGI! FI,0W IS Y/.6 x 10 '

#DI" ~ I,1C /iR

3. TorAL rEAKIIIG FAf30H I3 f 3 0'314 CCE, IPECCIPE IS 2* f 00 I" JIG i W 1" ' ~

@ 33_ $. WATFH I??!ET.13 10 FT. 6 III . ._._.

APorr' T:U: TOP OF Tirr: AGTIE l yn ,'. , u ;

g FUEL.

$ no - -<

1

,

$3a.

{' ?0- ---

F OR PF AMING F ACTO % >108,

- ,

$l * 3 Slg g

PF I"'

R 60 - qW WHERE- Sl =5AFETY LIWli TOR Pf AK!NG > 3 08

PF = PE AMING F AC TOR - 3 08*

Sl," 5AFETY LfutT SHOTM ABOVE %

S3 - -

10 --

, _

l 1 . I I I l ! I I J_ 1

0 13 71 39 81 50 60 10 80 *) 100 1 14 120

CORE F LOW (*. OF DESIGM)

FIGURE 711 FUEL CLADCING INTEGRITY 5AFETY Listi

2.1/2 3 10RW

. - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. . . _ . . - . _ . - . -. . . -.

-

-

.

Banes Continued:

The feedvater terperature assumed was the rxtximum design temperature cutput of the feedwater heatersat the given pressures and flows, which is 376"F for rated thermal power. For my hier feedwatertemperature, sub-cooling is increased and the curves are conservative.

The water level assumed in the calculation of the safety limit was that level curresprindire to the I5bottom of the steam sep trater skirt (7" on the level instrument is equivalent to 10~6" above the topof the active fue1 nt rated power). As long as the water level is above this point, the safety lirJtcunres are applic tble; i.e. , the 1:"nunt of stem carry under would not be increane d, ml, th'*refore ,tho core iniet enthalpy uvl sub-cooling would not be influenced.

The values of the parameters involvod in Firure 2.3.3 con be determimd frv inrornation availablein the control room. Rt actor pressure and flow are recorded and the Avciac< Fev r Range Monitor

" ( APRM) in-core nuclear i nstrumentation is c'tlibrated to ren 1 in im ms of percent ;hcr.

The range in pressure an i flow used fer Specification 2.1.1. var (^ psic to ]?~)0 pnic a r! 57. to 1007.

flow respectively. Gpecification 2.1.B requires a restriction on pwc: I cyt ; hen op rstinc belnw600 psig or 5 7. flow. In general, Opecification 2.3.B will only 1 applicau e auring ctortup orshutdown of the plant. A review of all the applicable lov presruve a:ad le1 flor d st (2, 3) laashown the lowest data point for transition boiling to have a h :at f3ur of 1M,000 ITd/lih/Ft' . Toassure applicability to Monticello fuel Cecmetry and provide rme r.arcin, a factor of 1/2 sia us+ ato obtain the criticitl heat flux; i.e., critical heat flux was assunod to occar for thr.se ccnj itions f 32at 72,000 PTU/HR/Ft . Assumin6 a peaking factor of 3.08, this is equivalent ta a cose av( rart -

power of approximately 300 !T4(t) (IP;f, of rated). This value is applicable t o ci,innt prcscu~ andno flow conditions. For any greater pressure or - flow conditions, the re i s ircretnr d :.mrgin.

(2) E. Janssen "Multiro 1 Eurnout at Low Pressure" - ASME Pape r '-!fr-26, t.urist 3 % 2.(3) K. M. Becker " Burnout Conditiens for Flow of P, oiling Water in Verticsl Erl Cluators" - AE T4

(Stockholm, Sweden),11ay,1"62.

.

2.1 BASES 15FIV

-. . . ___. . . . - . -- - - - . . . . - - - - = . . . -

.

|

TABLE 3 1.1 -

RFACIUR fROTECTION SYSTEM (SCRA!() IICTRUWJ3T REQUIREMENTS

'Modes in which func- Total No. of Min. No. of Operable

Limiting tion must te Oper- Instruhent or Operating Instru-Trip Settinr;s able or Operatin # Channels per - :nent Channels Fer Required

; Trip Function Refuel ( 3) Startup Run Trip System Trip System (1) condition

1. Mode Switch inShutdown x x x 1 1 A

:

2. Manual Scram x x x 1 1 A /Ti

; 3 Neutron Flux IRM h120/125'

(See Note 2) of full scale x x x(c) h 3 Ai a. High-HiFJ1

b. Inoperative,

b. Flow Refertaced See Specif1-! Neutron Flux APRM cations |

(See Note 5) 2 3^.1 x 3 2 A or B {a. High-High !

b. Inoperative,

c. Downscale $3/125 of{j full scalet~

5 High Reactor, x(f) x(f)- 2 2 A ',

{Pressure 6 1075 psig x

6. High Drywell [l |Pressure 62 psig. x(h) x(e,f) x(e,f) 2 2 A I

~

! 7 Reactor Iowi water Level 2 7 in.(6) x x(f) x(f) 2 2 A ?

|

8. Scram Discharge !

; Volume High Level 632 gal.(8) x(a)' x(f) x(r) 2 2 Af: 9 Turbine Condenaer

Icw Vacutan d 23 in. Hg x(b) x(b, f) x(f) 2 2 A or C!,

3 1/4.13o*

!

i.3

. - -- .n ,.

. __ - . . _ . . . - - . - .- .... .- . --. . - _ _ . . - . . . _ _ - . - . . - . _ . _ . - . . - . ...

.

,

! .!~

l

|

,

Table 3.2.1 - Continued . ,;

!~

r

Min. I;o. of OperableTotal No. of Instru- or Operating Instru-

d

ment Channels Per ment Channels Per Trip RequiredFunction Trip Settings, Trip System System (1.2) Conditions

i

()b. !!igh Drywell Pressure '

; (5) 6 2 psig 2 2 D j,

| 3. Reactor Cleanup System'(Group 3) ;

a. Iow Reactor Water 210'6" aboveLevel the top of the 2 ~2 E

active fuel

i

| 4. IIPCI Steam Lines!

a. itPCI Iligh steam Flow $150,000 lb/hr 2(4) 2 Fwith 3:;60 second

2 time delay ,

b. IIPCI I!1gh Steam Flow s;300,000 lb/hr 2(4) 2 F'

|'c. IIPCI Steam Line < 2000F- 16(4) 16 F_.

| Area Illgh Temp.

5. RCIC Fteam Lines

a. RCIC liigh Steam Flow <_45,000 lb/hr 2(4) 2 G

__2000F 16(4) 16 G<| b. RCIC Steam Line Area

Iligh Temp.

>

IJ

I !

3.2/4.2 51 |REV

_,

. .. . . . . - . - .. . . - - . .- -.

.

.

.

Table h.2.1 - ContinuedMinimum Test and Calibration Frequency For Core Cooling

Pod B1cek and Isolation Instrumentation

Instrument Channel Test (3) Calibration (3) Sensor Check (3)f%

3 Steam Lino Iov Pressure Note 1 Once/3 months Noneh. Steam Line High Radiation Once/veek (5) Note 6 Once/chift i

HPCI ISOIATION

1. Steam Lino Iligh Flov Note 1 Once/3 months None2. Steam Line Iligh Terperature Note 1 . Once/3 months None

RCIC ISOIATION

1. Steam Line High Flov Note 1 Once/3 months None2. Steam Lino High Temperature Note 1 Once/3 months Ncne

REAC'IOR BUILDIIC VENTIIATION

'

l. Radiation Monitors (Flenum) Note 1 Once/3 months Once/ shift2. Radiation Monitors (Reite11ng Floor) Note 1 Once/3 months (h)

' '

OFF GAS ISOIATION

1. Radiation Monitors Notes (1,5) N to 6 Once/ shift|

NOTES:

5(1) Initially once per month until exposure hours (M as defined on Figures h.l.1) is 2.0 x 10 , thereafteraccording to Figure h.l.1, with an interval not Breater than three months.

62Fa

32/4.2

__ _ _ _ _ _ _ - - - .- . ._. - . . . - . _ .- - ..- . . _ ~ _ _ _ . . . , , - . - _.. . ..

.

.

.

i

.

Table 3.2.5 - ContinuedTrip Function and Deviations

__

Trip Function Deviation

Instrumentation That Initiates Emergency low-Iow Reactor Water Level -3 InchesCore Cool ing Systems

Table 3.2.2 Reactor Low Pressure (Pump -10 psiStart) Permissive p

High Drywell Pressure +1 psi |

Low Reactor Pressure (Valve -10 psiPermissive

Instrumentation That Initiates IRM Downscale -2f:25 of Scalei

Rod Block IRH Upscale - +2/125 of Scale4

Table 3.2.3.

APRM Downscale -2/125 ot ScaleAPRM Upscale See Basis 2.3 - Page 24

RBM Downscale -2/125 of Sca?e -

RBM Upscale Same as APRM Upscale

,

i A violation of this specification is assumed to occur only when a device is knowingly set outside of l _)the ILmiting trip settings, cr, wi.en a sufficient number of devices have been affected by any meanssuch that the automatic function is incapable of operating within the allowable deviation while in |a reactor mude in which the specified function must be operable or when actions specified are notinitiated as specified.

3.2 BASES 70REV

i

5,

. - - - . . - - . . . - . . _ - . . . _ - - . - . . . . . . . . . . . - . . . . . _ . . - - - - . . . . . . . - . . ~ - . . . . . . - - - - . - . . . - . . _ - . .

. >

j-

ii:>

+

--I

i

1

1

.'

e

i

i. iiii

imo ;

N=

,

!o.-4 i

i I I i | |- | (3 I*

e.-eOS 24 - _a

4 m :

f M .

le

- t$z i

22 -O

3 21.4% 21.4%. 2895 gal1400 gal . ;4 ,,

5N - - j:e,,

1 F !: x

- :,

w 18 -4

i o ,

ec . t-

w: n-! 16

- -

.

Ni

s ( ~' :!

; 14 -

14.1%,2210 gal- t

' -

: . !4

12- -

i 10.81. 2895 gaf ,

. ,

, 10 - - :; 1 I I I I i-

'

'

1000 2000 3000'

i NET TANK VOLtME (gallons),

:9

f FIGURE 3.h l. Sodium Pentaborate Solution Volume. [! - Concentr tion Requirements 92 |

ac:; 3.h/h.h !i +

;___.__._ _ _ _ _ _- _ _ . _ _ _ _ = _ - _. . , - .

. _ - . . _ - - . . . _- _ _ _ _ . _ . . . _ . - -. -,

,

l'

^( ) ii A '-

1.

''.

,

,

1 l2

4- 1

m a:,

! l# i i ,

!

:1

150 - 7 -

. SOLUTION TEMPERATURE WUST BE

EQUAL TO OR GREATER TH AN TH ATINDICATED BY THE CURVE140 - -

.

! 130 - -

! _ 120 - -

4 &1 C

$110 - -

$t5 100

'

- -

-

! 5: p

| 5 90 - -

a! ;

. 80- -

,J

70 - -

60 - .

.

! 50- -

!

!i i i i

43

! 0 10 20 30 40 50

iWEIGHT PERCENT SODIUM PENT ABORATE IN SCLUTION

(as W/o Na2 10 16 10 11 0)B 0 2

4,.k'

".FIGURE 3.k2 SOCIUM PENTA 80R ATE SOLUTION TEMPERATURE REOUIREMii4TS m

i

:i*

- - ~ .- . - . ,_ , ,- .-, , . - - -, , . - _ - , , . , - . - . _ . .

. . .- . .-. - - - . -- _ . . . . . - . - - ..- . _ ~ - .. ..

*

,

'.

'

;

Bases Continund 3.6 and 4.6: ,

D. Coolant Leakage i

Ihe former 15 gpm limit for leaks from unidentified sources was established assuming such leakage was coming |from the primary system. Tests have bc en conducted which denonstrate that a relationship exists between the size j

of a crack and the probability that the crack will propagate. From the crack size a leakage rate can be determined. '

For a crack size which gives a Icakage of 5 gpm, the probability of rapid propagation is less than 10-5 Thus, an i

unidentified leak of 5 gpm when assumed to be from the primary system had less than one chance in 100,000 of propa- ,

'gating, which provides adequate margin. A leakage of 5 gem is detectable and measurable. The 24 hour periodallowed for determination of Icakage is also based on the low probability of the crack propagating. ,

The capacity of the drywell sump pumps is 100 gpm and the capacity of the dryvell equipment drain tank pumpsis also 100 gpm. Removal of 25 gpn from either of these sumps can be accomplished with considerable margin.

,

An annual report will be prepared and submitted to the AEC summarizing the primary coolant to drywell leakage. ,

measurements. Other techniques for detecting Icaks and the applicability or these techniques to the MonticelloPlant will be the subject of continued study. !

i

E. Safety and Relief Valves ;

Experience in safety valve operation shows that a testing of 50% of the sr.fety' valves per refueling outage is !

adequate to detect failures or deterioration. A tolerance value is specified in Section 111 of the ASME Boiler and *

Pressure Vessel Code as +1% of the set pressure. An analysis has been perfarmed which shows that with all safety.

valver ret 17. higher than the set pressure, the reactor coolant pressure safety Ibnit of 1375 psig is not exceeded.Safety / relief valves are used to minimize activation of the safety valves. The operator will set the pressuresettings ar er below the settings listed. liowever, the actual set points can vary as listed in the basis of |

Specification 2.4.-

, ,

The required safety valve steam flow capacity is determined by analyzing the pressure rise accompanying the -

main steam flow stoppage resulting from a MSIV closure with the reactor at 1670 MWt. The analysis assumes noMSIV closure scram, but a reactor scram from indirect means (high flux). The relief and safety valve capacityis assuned to total 83.9% (477. relief and 36.9% safetv) of the full power steam generatkn rate. This capacitycorresponds to assuming that four safety / relief valves (477.) and four safety valves (36.9%) operated. ;-

4

F

3.6/4.6 BASES ;

134 i

REV

. . . . . . .

.

.

.

3.0 LDilTUM COI;DITICI;S FOR OPEP12I'XI h.O SURVEILIWiCE F2QUIFDIEITIS

C. Secondsry Containment C. Secondary Containment

1. Srconda g containment integrity, shall be 1. Seconlary containnent surveillance shall pma.intained during ill me<!ns of plant be performed as indicated below: ~

oraration except when all of the followingccnditions are met.

The reactor is suberitical and Specifi- a. Seconda f containment capability toa.

cation 3.3.A is net. maintain at least a 1/4 inch of watervactrx under caln vind (<5 mph)conditions with a filter train flowrate of $ 4,000 scfm, shall be dem-onstrated at each refueling outageprior to refueling. This surveillancetesting should be reported in thesemiannual operating reports.

.

b. The reactor water ten erature is below2120 and the reactor coolant system is i

vented. -

c. Ilo activity is being performed whichcan reduce the shutdown margin belowthat specified in Specification 3.3. A.

3 7/4.7 150REV

,

_ ._ _ _. . _ _ . . _ . _ . . _ _ - _ _ _ __ _ .__

1

r

.

.

,

j

*..

ii

!

Bases Continued: I

!

The seceptable values for loesi leak rate tests have been specifiad in terms of ntandardcubic feet per hour (scf/hr) for purposes of clarity. Following is the list of equivalentval tes given in terms of an allowable percentage of the allowable operationil leak rate(L o)-t

17.2 sef/hr = % Lgo O@ hl psig

3h.4 ccf/hr lef, L=tc@ 41 psig '

103.2 scf/hr = 30$ Lto@ 41 psig

where Lto = , T5 Lt (the maxirarn allowable leak rate)and Lg = 1.2 weirbt percent cf the contained air at the test pressure of hl psig.

1

Results of loss of coolar.; accident annlyses indicate that fission products would not be released-

directly to the environs because of leakage throurJi the main line isolation valves due to holdupin the steam system complex. Although this effect shows that an adequate margin exists with

* regard to release of fission products, the results of leak tests on the main steam line isolation !

valves will te closely followed in order to determine the adequacy of these valves to perform g [their intended function. A summary report of the results of main steam line isolation valve s !

leakage tests and closure time measurements will be prepared and subnitted to the AEC following '

completion or periodic main steam line isolation valve leakage tests.!>

Monitoring tly nitrogen makeup require.ments of the inerting system provides a method of otsertingleak rate tree.e and would detect gross leaks in a very short_ time. This equipment must beperiodically remve l from service for test and rnintenance, but this out-of-service time will be

tkept to a pract ; cal minirar,.

;

h.7 EMES

16h

RE7,

__ _.___m __ . _ _ _ _ _ _ _ _ _ _ . _-

.. - . _ - -_ .- . . . . - -

>

.

..

.

(d) lii ghen t , lowest, and the annuni average concentrations or Icvels of -*

,

radiation for the sampling point with the highest average and description.

- of the location of that point with respect to the site. '

- .

(2) If levels of radioactive materials in environmental media as determined by an

^

environmental monitoring program indicate the likelihood of public intakes inexcess of 1% of those that could result from continuous exposure to the i

concentration values listed in Appendix B, Table II, Part 20, estimates of thelikely resultant exposure to individuals and to population groups, and assumptionsupon which estimates are based shall be provided. Ih*

(3) If statistically significant variation of offsite enrironmental concentrationswith time are observed, correlation of these results with effluent release shall

- be provided.

1. Occupational Personnel Radiation Exposure

Tabulate the number of personnel exposures for plant personnel (permanent and temporary)in the following exposure increments for the reporting period:

i

less than 100 mren, 100 - 500 mrem, 500 - 1250 mrem, 1250 - 2500 mrem,above 2500 mrem.

Tabulate the number of personnel receiving more than 500 mrem exposure in the reportingperiod according to duty function, i.e, routine plant surveillance and inspection(regular duty), routine plant maintenance, special plant maintenance (describe main- T

'tenancek routine refueling operations, special refueling operation (describe operation)and other job related exposures. Annually tabulate the number of personnel receivingmore than 2500 mrem and report major cause(s),

i

|6.7 216

REV

i

. _ . . _ _ _ . _ _ _ . __ .J

. . - _ .- - - _ - . .. . - . . . . . - _ _ . . _ --

.

s. .

L.

~.

B. _Ny e Routine Reports,

1. Abnormal Occurrence Feports

Notification shall be made within 24 hours by telephone and telegraph to theDirector of the Regional Regulatory Operations Office (cc to the Director ofLicensing), followed by a written report within 10 days to the Director ofLicensing (cc to the Director of the Regional Regulatory Operations Of fice)'

in the event of the abnormal occurrences as defined in Section 1.0. The writtenreport on these abnormal occurrences, and to the extent possible, the preliminary / 5telephone and telegraph notification, shall: (a) describe, analyze and evaluate

. safety implications, (b) outline the measures taken to assure that the cause ofthe condition is determined, (c) indicate the correcti"e action ' including anychanges made to the procedures and to the quality assurance program) taken toprevent repetiticn of the occurrence and et similar accurrences Involvingsimilar components or systems, and (d) evaluate the safety implications ofthe incident in light of the cumulative experience obtained f rom the , recordof previous failures and ac1 functions of similar systems and components.

.

%

W

6.7 216A_

REV

9

_ _. . . _ _ _ . _ _ _ . . . .- _ _ - .. _ _ _

AEC D'TRIBUTION FOR PART SO DOCKET }" P?tT AL

(TDIPORARY FORM) CONTROL NO: 7931s-

..3

FILE:.

FROM: DATE OF DOC DATE RFC'D LTP. 11D 0 RPT OTHERNorthern States Power CompanyMinneapolis, Minnesota $$401 10-26-73 10-31-73 X

L. O. Mayer

70: ORIG CC OTHER SEliT AEC PDR X

J. F. O' Leary 3 signed SENT LOCAL PDR X

CLASS UNCLA'.'S PROP INFO INTUT NO CiS REC'D DOCKET !;O: |XXXX XX 40 50-263 I

i

DESCRII' TION: ENCLOSURES:Ltr te their 8-20-71 Itr & their 9-22-73 ltr,.. , CHANGE OF TECH SPECS-dtd & notarised

trans the following: 10-26-7J

ACKNOWLEDGED.

DO NOT REMOVEPLAliT NAME: Montice lo ( 3 Orig & 37 cys re'd )

FOR ACTION /INFORMATION 10-31-73 GC

BUTLER (L) SCir4ENCER(L) / ZID1 ANN (L) REGAN(E)W/ Copics W/ Copies W/ 9 Copies W/ CopiesCLARR(L) S1VLZ(L) DICKER (E)W/ Copics W/ Copics W/ Copies W/ CopicsCOLLER(L) VASSALLO(L) KNIGHION(E)W/ Copics W/ Copies W/ Copies W/ Copies}3;IEL(L) SCHQ:EL(L) YOUNGELOOD(E)W/ Copics W/ Copics W/ Copies W/ Copics

INTERNAL DISTRIEUTION-

@ G FILE N TECH REVIL4 .m1 TON A/T INDLIC ASSTy AEC'PDR'' HENDRIE G.11tES BRAITMAN.

/ OGC, R00:1 P-506A SCHROEDER C/2:11LL <DIGGS (L) SALTZMA!!/MUNT;'ING/ STAFF MACCARY KASTNER GEARIN (L) B. HURT

CASE KNIGHT BALLARD GOULBOURNE (L*) ptg);gCIAMBUSSO PAWLICKI SPANGLER LEE (L) MCDONALDBOYD SHA0 MAIGRET (L) /DU B"~HOORE (L)(Ir4R) STELLO ENVIRO SERVICE (L)DEYOUNG(L)(FWR) HOUSTON MULLER SHEPPARD (E) INFO

v SKOVHOLT (L) NOVAK DICKER SMITH (L) C. MILESP. COLLINS ROSS KNIGHTON TEETS (L) 7 A. Cabell

IPPOLITO YOUNGBLOOD WADE (E)

/EGOPR TEDESCO REGAN WILLI /J:S (E)i FILE 6 REGION (3) LONG PROJECT LDR WILSON (L)

MORRIS LAINASSTEELE ED:AROYA HARLESS

'

VOLU'?REXTERNAL DISTRIEUTION #4'

_ _ _

4 - LOCAL FE3 Minneapolis, Minn. j.

| 4 - DTIE(APE"NATHY) (1)(2X 10 bMATIONAL LAB'S 1-PDR-S AN /LA/tFlj V1 - USIC(TUCHA';AN) 1-GERALD LELLOUCHEi I - ASLE(YORE /SAYRE/ 1-U. PENNINGTON, Ih E-201 GT BROOKHAVEN NAT. LAE

WOCGATD/"H" ST. 1-CONSULTANT' S 1-ACMED(Muta Gusscan)16 - CYS ACRS 2nctXXX SENT TO LIC ASST. NENMARE/ BLU!!E/AGBAEI AN RM- B-12 7. GT.

10-31-73 DIGGS 1-GERALD ULEInSON. . 0;U:L 1-la.. MULLER..F-309 G4

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