ML20024G810

Proposed Tech Specs Re Safety Limits

ML20024G810
Person / Time
Site:	Monticello
Issue date:	10/26/1973
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G800	List: ML20024G799 ML20024G810
References
NUDOCS 9104300436
Download: ML20024G810 (14)

Text

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1. RATED IU'4ER In 16, fGt

?. DESIGI! FI,0W IS Y/.6 x 10 '

I" ~ I,1C /iR #D

3. TorAL rEAKIIIG FAf30H I3 f 3 0'3 14 CCE, IPECCIPE IS 2* f 00 I" JIG i W 1" ' ~

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WHERE- Sl =5AFETY LIWli TOR Pf AK!NG > 3 08 PF = PE AMING F AC TOR - 3 08 Sl," 5AFETY LfutT SHOTM ABOVE  %

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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 10 RW

Banes Continued:

The feedvater terperature assumed was the rxtximum design temperature cutput of the feedwater heaters at the given pressures and flows, which is 376"F for rated thermal power. For my hier feedwater temperature, 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 I5 bottom of the steam sep trater skirt (7" on the level instrument is equivalent to 10~6" above the top of the active fue1 nt rated power). As long as the water level is above this point, the safety lirJt cunres 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 available in 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 belnw 600 psig or 5 7. flow. In general, Opecification 2.3.B will only 1 applicau e auring ctortup or shutdown of the plant. A review of all the applicable lov presruve a:ad le1 flor d st (2, 3) laa shown the lowest data point for transition boiling to have a h :at f3ur of 1M,000 ITd/lih/Ft' . To assure applicability to Monticello fuel Cecmetry and provide rme r.arcin, a factor of 1/2 sia us+ a to obtain the criticitl heat flux; i.e., critical heat flux was assunod to occar for thr.se ccnj itions f 3 at 72,000 PTU/HR/Ft 2 . 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~ and no 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 15 FIV

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

l 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 in Shutdown x x x 1 1 A

2. Manual Scram x x x 1 1 A /T i

3 Neutron Flux IRM h120/125 (See Note 2) of full scale x x x(c) h 3 A i 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 scale t 5 High Reactor Pressure 6 1075 psig x , x(f) x(f)- 2 2 A '

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

~

! 7 Reactor Iow i 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 A f

9 Turbine Condenaer Icw Vacutan d 23 in. Hg x(b) x(b, f) x(f) 2 2 A or C 3 1/4.1 3o i.

3

. - -- .n ,.

~

l Table 3.2.1 - Continued . ,

~

Min. I;o. of Operable d

Total No. of Instru- or Operating Instru-ment Channels Per ment Channels Per Trip Required Function Trip Settings, Trip System System (1.2) Conditions i

()

b.  !!igh Drywell Pressure '

(5) 6 2 psig 2 2 D l 3. Reactor Cleanup System (Group 3)  ;

a. Iow Reactor Water 210'6" above Level the top of the 2 ~2 E active fuel i

l 4. IIPCI Steam Lines

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

time delay ,

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

c. IIPCI Steam Line < 2000F-

_. 16(4) 16 F l Area Illgh Temp.

5. RCIC Fteam Lines

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

__2000F 16(4) 16 G Iligh Temp.

J I

3.2/4.2 51 REV

Table h.2.1 - Continued Minimum 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 None

h. Steam Line High Radiation Once/veek (5) Note 6 Once/chift i HPCI ISOIATION

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

2. Steam Line Iligh Terperature Note 1 . Once/3 months None RCIC ISOIATION

1. Steam Line High Flov Note 1 Once/3 months None

2. Steam Lino High Temperature Note 1 Once/3 months Ncne REAC'IOR BUILDIIC VENTIIATION

l. Radiation Monitors (Flenum) Note 1 Once/3 months Once/ shift

2. Radiation Monitors (Reite11ng Floor) Note 1 Once/3 months (h)

OFF GAS ISOIATION

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

NOTES:

(1) Initially once per month until exposure hours (M as defined on Figures h.l.1) is 2.0 x 105 , thereafter according to Figure h.l.1, with an interval not Breater than three months.

62 Fa 32/4.2

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

i Table 3.2.5 - Continued Trip Function and Deviations Trip Function Deviation Instrumentation That Initiates Emergency low-Iow Reactor Water Level -3 Inches Core Cool ing Systems Table 3.2.2 Reactor Low Pressure (Pump -10 psi Start) Permissive p High Drywell Pressure +1 psi l Low Reactor Pressure (Valve -10 psi Permissive i Instrumentation That Initiates IRM Downscale -2f:25 of Scale 4 Rod Block IRH Upscale -

+2/125 of Scale

. Table 3.2.3 APRM Downscale -2/125 ot Scale APRM 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 means such 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 not initiated as specified.

3.2 BASES 70 REV i

5

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10

1 I I I I i-1000 2000 3000 i NET TANK VOLtME (gallons) f FIGURE 3.h l. Sodium Pentaborate Solution Volume.

! - Concentr tion Requirements 92 ac:

i 3.h/h.h +

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150 SOLUTION TEMPERATURE WUST BE EQUAL TO OR GREATER TH AN TH AT 140

- INDICATED BY THE CURVE -

! 130 - -

! _ 120 - -

4 &

1 C

$110 - -

t 5

100 - -

! 5 p

l 5 90 - -

a

. 80

,J 70 -

60 - .

! 50

! i i i i 43

! 0 10 20 30 40 50 i

WEIGHT PERCENT SODIUM PENT ABORATE IN SCLUTION (as W/o Na2B100 16 10 112 0) 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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed 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 pumps is 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 Monticello Plant 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 pressure settings ar er below the settings listed. liowever, the actual set points can vary as listed in the basis of l 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 no MSIV closure scram, but a reactor scram from indirect means (high flux). The relief and safety valve capacity is assuned to total 83.9% (477. relief and 36.9% safetv) of the full power steam generatkn rate. This capacity

- corresponds 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 p ma.intained during ill me<!ns of plant be performed as indicated below: ~

oraration except when all of the following ccnditions are met.

a. The reactor is suberitical and Specifi- a. Seconda f containment capability to cation 3.3.A is net. maintain at least a 1/4 inch of water vactrx under caln vind (<5 mph) conditions with a filter train flow rate of $ 4,000 scfm, shall be dem-onstrated at each refueling outage prior to refueling. This surveillance testing should be reported in the semiannual operating reports.

b. The reactor water ten erature is below 2120 and the reactor coolant system is i vented. -

c. Ilo activity is being performed which can reduce the shutdown margin below that specified in Specification 3.3. A.

3 7/4.7 150 REV

r j

i i

Bases Continued: I The seceptable values for loesi leak rate tests have been specifiad in terms of ntandard cubic feet per hour (scf/hr) for purposes of clarity. Following is the list of equivalent val tes given in terms of an allowable percentage of the allowable operationil leak rate (Lt o)-

17.2 sef/hr = % Lgo O

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

@ 41 psig '

103.2 scf/hr = 30$ L to

@ 41 psig where Lto = , T5 Lt (the maxirarn allowable leak rate) 1 and Lg = 1.2 weirbt percent cf the contained air at the test pressure of hl psig.

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 holdup in 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 otserting leak rate tree.e and would detect gross leaks in a very short_ time. This equipment must be periodically remve l from service for test and rnintenance, but this out-of-service time will be t kept 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 in excess 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 the likely resultant exposure to individuals and to population groups, and assumptions upon which estimates are based shall be provided. Ih (3) If statistically significant variation of offsite enrironmental concentrations with 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 reporting period 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 receiving more than 2500 mrem and report major cause(s),

6.7 216 REV

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

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~

B. _Ny e Routine Reports ,

1. Abnormal Occurrence Feports Notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the Director of the Regional Regulatory Operations Office (cc to the Director of Licensing), followed by a written report within 10 days to the Director of Licensing (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 written report on these abnormal occurrences, and to the extent possible, the preliminary / 5 telephone and telegraph notification, shall: (a) describe, analyze and evaluate safety implications, (b) outline the measures taken to assure that the cause of the condition is determined, (c) indicate the correcti"e action ' including any changes made to the procedures and to the quality assurance program) taken to prevent repetiticn of the occurrence and et similar accurrences Involving similar components or systems, and (d) evaluate the safety implications of the incident in light of the cumulative experience obtained f rom the , record of 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 s (TDIPORARY FORM) CONTROL NO: 7931 3 ..

. FILE:

FROM: DATE OF DOC DATE RFC'D LTP. 11D 0 RPT OTHER Northern States Power Company Minneapolis, 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:

l 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 REMOVE PLAliT 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/ Copies CLARR(L) S1VLZ(L) DICKER (E)

W/ Copics W/ Copics W/ Copies W/ Copics COLLER(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 IND y AEC'PDR HENDRIE G.11tES LIC ASST 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);g CIAMBUSSO PAWLICKI SPANGLER LEE (L) MCDONALD BOYD 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. MILES P. 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 LAINAS STEELE ED:AROYA HARLESS VOLU'?R EXTERNAL DISTRIEUTION #4

. 4 - LOCAL FE3 Minneapolis, Minn. j l 4 - DTIE(APE"NATHY) (1)(2X 10 bMATIONAL LAB'S 1-PDR-S AN /LA/tFl j V1 - USIC(TUCHA';AN) 1-GERALD LELLOUCHE i

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.

4 10-31-73 DIGGS 1-GERALD ULEInSON. . 0;U:L 1-la.. MULLER..F-309 G

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