Text

Proposed Tech Specs Increasing Max Power Level by 6.3% to 1,775 from 1,670 Mwt.Summary of Plant Mods for Power Rerate Implementation & Power Rerate Environ Evaluation Encl
	ML20116G895
Person / Time
Site:	Monticello
Issue date:	07/26/1996
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML19311C180	List: ML20116G875; ML20116G891; ML20116G895;
References
	NUDOCS 9608090087
	Download: ML20116G895 (102)
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EXHIBIT B

LICENSE AMENDMENT REQUEST DATED JULY 26,1996 l

l l PROPOSED CHANGES MARKED UP ON EXISTING i

l OPERATING LICENSE AND TECHNICAL SPECIFICATION PAGES l

(

l N !

l NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA LICENSE NO. DPR 22 DOCKET NO. 50-263 L

9600090087 960726 PDR ADOCK 05000263 p PDR

e v i .

l l

Exhibit B Monticello Nuclear. Generating Plant Ucense Amendment Reauest dated July 26.1996 Proposed Changes Marked Up on Existing Operating License and Technical Specification Pages Exhibit B consists of the existing Operating License and Technical Specification pages with the -

proposed changes marked up on those pages. Existing pages affected by this change are listed below.1 Affected Pace Associated Chance Affected Pace Associated Chance License pace 3 A 30" E 4 A 19 E 14 -A 38 E 15* 'A 50" F 6 B 53 " G 56 B M G 16 B 104 H 114* B 101 H 23 C 112 H,i 24 - C 176 H

150 C ii H,i 28 D iii I 37 D-

. Pages reflect proposed changes submitted with amendment request entitled

" implementation of BWR Thermal Hydraulic Stability Solution," dated June 22,1995, TAC M92947.

Pages reflect proposed changes submitted with amendment request entitled " Surveillance

' Test Interval / Allowed Outage Extension Program - Part 2," dated July 5,1995, TAC M92948.

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~ . - . ~ . - - . - - . . - - - .- - - - - .- - .- - .

L 3

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4. Pursuant-to the Act and 10 C7R Parts 30, 40 and 70, to receive, !

possess and use in amounts as required any byproduct source or- {

special nuclear material without restriction to chemical or j physical form, for sample analysis or instrument calibration' j or associated with radioactive apparatus or components; and j i

5. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, '

but,not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations. in 10L CFR Chapter I and is subject to all pplicable provisions of the Act and to the rules, regulations, and orders of the Cornission, nowL or hereafter in effect; and is subject to the additional conditions j specified or incorporated below:

1. Maximum Power Level q The license's is authorized to operate the facility at steady

~

b state reactor core power levels not'in excess of g mega- k watts (thermal)

37 7 S .

]e .

2. Technical Soecifications :E' c

The Technical Specifications contained in Appendix A as i revised through Amendment No. 96 are hereby incorporated in the license. The licensee shall operate the facility in accordance l1 with the Technical Specifications.

E

i lJ 3. Physical Protection The licensee shall fully implement and maintain in effect all l

provisions of the Commission-approved piiysi~al~ c security, guard '

training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Monticello Nuclear Generating Plant Physical ao Security Plan," with revisions submitted through November 30, 1987; * ;

"Monticello Nuclear Generating Plant Guard Training and Qualification d i Plan," with revisions submitted through-February 26, 1986; and *

"Monticello Nuclear Generating Plant Safeguards Contingency Plan," $

with revisions submitted through Augu.st 20, 1980. Changes made in- E l accordance with 10 CFR 73.55 shall be implemented in accordance with 1- !

the schedule set forth therein. j 5

i

4. Protective the channels monitoring a particular plant condition.

Function - A system protective action which results from the protective acti on of

<C R.-1476-Ratedthermal Neutron Flux - Rated flux is the neutron flux that corresponds to a steady-state megawatts. er level of pow d 17757 #

S. Rated Thermal Power - Rated thermal power means a steady-state power level of 4676- thermal megawatts. ;s 1775 t3 T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise indicated ,

reactor ves'sel pressures listed in the Technical Specifications are those existing in the vessel steam space .

U. Refueline Operation and Refueline Outare temperature is less than 212*F and movement of fuel or core components . -For the is in progress- Refueli purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly ue sched l d refueling.

refueling outage, outage; however, where such outages occur within 8 months of the completion of the previous scheduled outage. the required surveillance testing need not be performed until the next regularly V. Safety system integrity Limit - The aresafety limits are limits below which the maintenance of the cladding and assured. primary Commission before resumption of plant operation. Exceeding such a limit is cause for plant shutdown and review by the Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review .

W. Secondary Containment Interrity closed and the following conditions are met:- Secondary Containment Integrity means that the reactor building is

1. At' least one door in each access opening is closed.

2. The standby gas treatment system is operable. ,

3. All in the reactor closed building ventilation system automatic isolation valves are operable or are secured position.

. X. Sensor operation. Check This determination - A qualitativeshall determination include, where of operability possible, by observation of sensor behavior during sensors measuring the same variable. comparison with other independent 1.0 4

REV -

s -

MASES:

2.3 The abnormal operational transients applicable 17 S g to oporation of the Monticello Unit. have bee analyzed g throughout the spectrum of planned operating conditions up to the thermal power level of HWt.

The analyses were based upon plant operation in accordance with the operating map. The licensed maximum n power level -life-MWt l'7 7 5 represents the maximum steady-state power which shall not knowingly be exceeded.

3 Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as 0 void powerreactivity shapes. coefficient, control rod scram worth, scram delay time, peaking factors, and axial These factors are selected conservatively with respect to their effect on the applicable tranisent results as determined by the current analysis model.

2.3 BASES 14 .

REV

Bases Continued:

For analyses of the thermal consequences of the transients, the Operating MCPR Limit (T.S.3.11.C) is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of -

control parameters and analyzing at higher power levels.

Deviations setting error, fromdrift as-left of the settings setpoint, of setpoints etc. are expected due to inherent instrument error,. operator Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discussed above. Although the operator will set the setpoints the specified within tripthe. setting trip settings by the allowable specified,deviation. the actual values of the various setpoints can vary from A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power b 17 7g /TI470 MWt). Because, fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from lI0 the fuel (reactor thermal power) is less than the instantaneous neutron flux ; due to the time constant of the fuel.

fuel will be less thanTherefore, during abnormal operational transients, the thermal power of the that indicated by . the neutron flux at the scram aetting. Analyses demonstrate that, with a 120% scram trip. setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin form fuel damage. Also, the flow biased neutron flux scram (spe'cification 2.3.A.1) provides protection to the fuel safety .f

-limit in the unlikely event of a thermal-hydraulic instability.

2.3 BASES 15 REV g gp -

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2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CIADDING INTEGRITY 2.3 RIEL CIADDING INTECRITY Anolicability Applicabilltv Applies to the interrelated variables associated Applies to trip settings of the instruments and with fuel thermal behavior devices which are provided to prevent the reactor system safety limits from being exceeded.

Objective:

To define the level of the process variables at To establish limits below which the integrity which automatic protective action is initiated to of the fuel cladding is preserved. prevent the safety limits from being exceeded.

Spec i fi ca tion:

Specification:

The Limiting safety system settings shall be as specified below:

A. Core Thermal Power Limit (Reactor Pressure >800 psia and Core Flow is >10% of Rated) A. Neutron Flux Scram When the reactor pressure is >800 psia and core 1. APRM - The APRM flux scram trip setting flow is >10% of rated, the existence of a shall be:

minimum critical power ratio (MCPR) less than a. For two recirculation loop operation 1.07, for two recirculation loop operation, or (TID): 65.6*/o less than 1.08 for single loop operation, shall where, constitute violation of the fuel cladding S s 0.66W + J8TL S - Setting in percent of rated integrity safety limit.

thermal power rated power d

beine MWT l'7'15 d re.c icwbd.m 'W - Percent 2-the- drive flow O required to produce a voted I

b. For single recirculation loop operation core flow of 57.6 x 10' lb/hr (SLD): O. (o6 g

GS.foYo S $ M (W - 5.4) + 4t%

c. No greater than 120s.

2.1/2.3 6

REV

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.I

. TABLE J.Z.J .

Instrumentation That Initiates Rod Block Reactor Modes Which Function Must be Operable Total No. of Min. No. of Oper-or Operating and Allow- Instrument able or Operating able Bynass conditions ** Channels per Function Trin Settines Re fuel Startun Run Trin System Instrument Channels Required I ner Trin System Conditions *

1. SRti

a. Upscale s5x105 cps X X(d) 2 1(Note 1, 3, 6) A or B or C !

b. Detector X(a) not fully X(a) 2 1(Note 1, 3,.6) A or B or C inserted

2. TRM i

a. Downscale 23/125 X(b) X(b) full scale 4 2(Note 1, 4, 6) A or B or C '

b. Upscale s108/125 X X 4 2(Note 1, 4, 6) full scale A or B or C !

3. APRM

a. U (1)pscale TLD 53.b o/o X 3 1(Note 1, 6, 7) D or E '

Flow < 0. 66W + )M ;

Biased - (Note 2) d 67 (2) SLO O, O 63 b Flow g 0,4e(V - 5.4) +)M d Biased (Note 2) g- ,

(3) 111gh s 108% d Flow Clamp

b. Downscale 2 3/125 full scale X 3 1(Note 1, 6, 7) D or E 3.2/4.2 56 REV

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Evm.6 & ons. M se wsma ^ HE oc.-325%P, Mr I ff6 6%edn A g . gc p b.Q\ h of t4E'LLLA ena %c.F 4o7 ysh c o4 h'0% .

2A e.Md %g %c, cA/oduchion cke. mons 4roke.d. -t-he, act.9hbelh Bases Continued: - o4 tAELLL A 4 o r s w0 ie. \eap o fcMM* -

i D (tAELLL A) j f Maximum Extended Lead Line Limit Analyses ave been performed to allow operation at higher powers at flows p below 87%.

y for two loop operation (See Core Operating Limits Report).The '"*.,... flow referenced scram1L. ^ (and rod block g ;gr ri;r f;r ;i-;1 1;; ;p; rc. tis-..

..1,_._"..-...,,m m.. ....a of... .;

The supporting analyses are discussed in CE NEDC-31849P report p (Reference! Letter from NSP to NRC dated September 16, 1992),

f., 'I t F)

@ Increased Core Flow nalyses have been performed to allow operating at flows above 100% for powers equal to or less than 100t (see core Operating Limit Report). The supporting analyses are discussed in General 5 Electric NEDC-31778P report (

Reference:

Letter from NSP to UFC dated September 16, 1992).

For operation in the startup mode while the reactor is at low pressure, the IRM scram setting of 20% of rated rated.

power provides adequate thermal margin between the setpoint and the safety limit, 25% of startup. The margin is adequate to accommodate anticipated maneuvers associated with power plant Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures.

Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.

Because peaks, the flux distribution associated with uniform rod withdrawals does not involve high local and because several rods must be moved to change power by a significant percentage of rated l

power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the IRM system would be more than adequate to -

assure a scram before the power could exceed the safety limit. The IRM scram remains active untti the mode switch is placed in the run position and the associated APRM is not downscale. This switch occurs when reactor pressure is greater than 850 psig.

'lhe operator will set the APRM neutron flux trip setting no greater than that stated in Specifica-tion 2.3.A.l. However, the actual setpoint can be as much as 31 greater than that stated in Specification 2.3.A.1 for recirculation driving flows less than 50% of design and 24 greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on page 39.

B. Deleted 2.3 BASES 16 REV i Next Page is 18 i

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

_ . ____u -- _,___.-m._._ -

Bases 3.5/4.5 cont M esli F. Recirculation System The reactor is designed such that thermal ~ hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal-hydraulic instability, a power-flow exclusion region, to be avoided during normal operation, is calculated using the approved methodology as stated in specification 6.7.A.7.

each fuel cycle the limits are contained in the Core Operating Limits Report.Since the exclusion Specific region may directions arechange g.

provided to avoid operation in this region and to-immediately exit upon an entry. Entries into the %-

A g, exclusion region are not part of normal operation. An entry may occur as the result of an abnormal event such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed 3 s to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though . g operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow biased { L scram function will suppress oscillations prior to exceeding the fuel safety limit. O g

Power distribution controls are established to ensure the reactor is operated within the bounds of the n e-b=

stability analysis. With these controls in place, there is confidence that an oscillation will not occur outside of the stability exclusion region. Without these controls, it is theoretically possible to operate G ;

the reactor in such a manner as to cause an oscillation outside of the exclusion region. A nominal 51 (Q power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the O4

[

analytically defined exclusion region. The buffer region may be entered only when the power distribution controls are in place. *j 5I '

Continu 3.5.F us operation with one recirculation loop was analyzed and the adjustments specified in specification [

g I

i Operatio were determined by NEDO-24271 June 1980, "Monticello Nuclear Generating Plant Single L p ;

loop. A Specification 3.6. A.2 governs the restart of a recirculation pump in an idle rceirculation $

stre s.

erence to this specification limits the probability of excessive flux transients and/or thermal t

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i NEXT PAGE IS 121 '

3.5/4.5 114 i REV t

Bases Continued:

2.2 The normal

"'e~ir =?*h operating pressure h of the reactor coolant system is approximately railurfafdLypass--syst.amesaprasants tha ==a=e s g, 1010 h pi pressure =evera_primasy gystem-valves (S/RV's) areresulting s from an abnormal operational transtent., The-safetf/ relief sa in assumed is from an indirect means no direct h-El scram during.MSIV closure. The only scram the eight S/RV's are o erable-and-~th'at they open aThe~ analysis e assiuses that only seven of second

-.;Et E .aa dalay-Re a1. e or pressure remains below the 1375 psig ASMtheir setpoint with a 0.4 ,

t for the *-

bVAlaak4^5 h n v c. cle.+e (mi oe.d kk 4he, %f gge g . 4seg 7 f c,.as;e A is bom A A by +h <- c!os.ac of c li m s Ws,4 glo w by ~ <<.a.eko e s.ei w a+ k: h na.h<os 4t" v. L 4 a;l u a

C 4h-d-i / ec. S C fo. # ALLoci g.d wi 'b MSlY pos:ike D A II ASS.w M ck). .

%.asa abcas -t+ 4cs. a4 +us. e +. % mtsc.s y

ve.wW s. at.% +<A, +lu scAA7 he. lie 4 .*tve. cafac;47 is g

cuydle. c4 una;Aa;dn3 r <cu.wre w.los +k A sm e ce d. ti d e4- 11 oc/o of vesse{ desga p <<. ss o. , c. (lI o % x taso pay : y> .r gg g ),

The sedeAy I:n ek c. a s <>. < c s +kek she c~ce y % s \ab i 04 d * ' "d 4 he. Aesicyn b c~s is eve r+

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2.2 BASES 23

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Bases: 1 g 0 %4c,c n o. A dt.5 @ ^ 9 #'S W C~J /

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r 2.4 The settings on the reactor high pressure scram, reactor coolant system safety / relief valves, turbine control valve fast closure scram, and turbine stop valve closure scram have been established to assure never reaching the reactor coolant system pressure ;

safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit. The APRM neutron flux scram and the turbine '

bypass system also provide protection for these safety limits. In addition to preven- '

ting power operation above 1075 psig, the pressure scram backs up the APRM neutron flux

scram for steam line isolation type transients. ,

t The reactor coolant system safety / relief valves assure that the reactor coolant system pressure safety limit is never reached.

In compliance with Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, the safety / relief valves must be set to open at a pressure no higher than 105 percent of design pressure,*and- they must limit ,

the reactor pressure to no more than 110 percent of design pressure. The safety / relief l775 va ves are sized according to the Code for a condition of MSIV closure while operating '

at w?e- Mut, fo11oved by no MSIv elosure scrs= but'scrs= from an indirect (high flux) means.

With the safety / relief valves set as specified h ein, the maximum vessel pres-

' sure remains below the 1375 psig ASME Code limit. Only of the eight valves assumed to be operable in this analysis and the valves are assumed to open at Y3 O

0 ges their setpoint*with a 0.4 second delay. The upper limit on safety / relief valve setpoint ove is established by the design pressure of the HPCI and RCIC systems af il2C A . ' y The operator will set the reactor coolant high pressure scram trip setting ac 1075 psig d or lower. e However, the actual setpoint can be as much as 10 psi above the 1075 psig T indicated set point due to the deviations discussed in the basis of Specification 3.1. I In a like manner, the operator will set the reactor coolant system safety / relief valve -1 initiat on trip setting at 1120 psig,or-lover. However, the ::p;l set point can be as p ,3 ""inuch as psi above the 1120 psig indicated set point due tol the deviations discussed d in the basis of Specification 3.6.

Ql1 Oy pig + t*/ ) L ah-So M -

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting, or when a sufficient number of devices have been affected by any means, T k M6Q^ cap bI +k UC'I O^d R4IC-

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a+ pc.ss uc s. 3Y / rcli 4 24 .

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Bases Continued 3.6 and 4.6: ..

D. Coolant Leakare

Ihe and experimentally allowable leakage observed ratesbehavior of coolant of cracksfrominthe reactor coolant system have been based one the predict d pipes.

to equipment design and the detection capability of the instrumentation efor wasdetermining lea

. also considered. The evidence obtained from experiments suggests that for leakage somewhat greater then-that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed

'the values specified or the leakage is located and known to be Pressure Boundary Leakage and they ca I reduced within the allowed times, the reactor will be shutdown to allow further. Investigation and corrective action.

Two leakage collection sumps are provided inside primary containment.

from the recirculation pump seals, valva stem leak-offs, reactor vessel flange leak-off, bulkheadIdentified leakage is and bellows drains, and vent cooler drains.to the drywell equipment drain aump. All other leakage is collected in the drywell floor drain sump.

mitters connected to recorders in the control room.Both sumps are equipped with level and flow trans-vided in the control room to alert operators when allowable leak rates are approached.An Drywell annunciator and compu airborne perature and pressure. particulate radioactivity is continuously monitored as well as drywell atmospheric ten-for leakage by the Proce'ss Liquid Radiation Monitoring System. Systems connected to the reactor coo The than sensitivity one gpm inofa the onesump leakage iMeetion systems for detection of leak rate changes is better hour period.

and are not directly calibrated to provide leak rate measurements.Other leakage detection methods provide war E. Safety /Re lef Valves

( sc.c.o&cL "i W 5t/d5M E ' OP4-l* If8 I Mt[ ASP (E OM-l-19 81 d hTesting tolerance of +14 value safetyof elief valves each refueling outage ensures that any valve deteriorat for safety / relief valve setpoints is specified in S::ti:r !!! :f is detected. O

d I'ressur '? :::1 0:d2 in the Section 2.2 Bases, the 1375 psig Code limit is not ex eeded in any t'_ .2"I 5:11: 2

'dg'.. 'As discussed case, Analyses have bee

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+

TABLE 3.L.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Modes in which func- Total No. of Min. No. ot Operable tion must be Oper- Instrument or Operating Instru-Limiting able or Operating ** Channels per ment Channels Per Required Trip Function Trip Settings Refuel (3) S tartup Run Trip System Trip System (1) Condition *

1. Mode Switch in Shutdown X X X 1

2. Manual Scram X X X 1 1 A

3. Neutron Flux IRM s 120/125 1 A (See Note 2) of full scale X X 4 3

a. liigh-lligh A

b. Inoperative

4. Flow Referenced See Specifi-Neutron Flux APRM cations (See Note 5) 2.3A.1 X 3 2

a. Iligh-liigh A or B

b. Inoperative

c. liigh Flow Clamp $ 120 1

5. Iligh Reactor Pressure s 1075 psig X X(f) X(f) 2 2 (See Note 9) A

6. liigh Drywelf Pressure s 2 psig X X(e.f) X(e,f) 2~ 2 A (See Note 4)

7. Reactor Low Water Level 2 7 in.(6) X X(f) X(f) 2 2 A

8. Scram Discharge Volume liigh Level

a. East s 56 gal.(8) X(a) X(f) X(f) 2 2

b. West A O

s 56 gal.(8) X(a) X(f) X(f) 2 2 A y 9. Turbine Condenser g Low vacuum 2,,yf in. lig X(b) X(b.f) X(f) 2 2 A or C o 22.

D 3.1/4.1 g 28 REV

_ ,m '- - - - - - '

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~

Bases Continued:

3.1 condenser eliminates the vacuum heat input initiates to the a closure condenser.of the turbine stop valves and turbine bypass valv es which:

a pressure transient, Closure of the turbine stop and by neutron flux rise, and an increase in surface heat flux. pass valves To prevent causes the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine closure. op valve st limit from Section 14.5.1.2.2 being exceeded in the event of a turbine trip transient Reference FSAR withouto bypas safety and supplemental information submitted February 13 1973 .

vacuum scram is a back-up to the stop valve closure scram and causes a, scram. The condenser low - N are closed and thus the resulting transient is less severe. before the sto C valve closure occurs at 20" lig vacuum, and bypass closure at 7" lig vacuum. Scramstop occurs at FjHg vacu (y//

g The s10% main closed steamline from fullisolation open. valve closure scram is set to scram when the occur when the valves close. This scram anticipates the pressure and flux transient, isolation which valveswould are Reference Section 14.5.1.3.1 By scramming at this setting the resultant transient FSAR and supplemental information submitted February 13is , 1973. insignificant.

Atoreactor the mode switch is provided which actuates or hypasses the various scram functions appropriate particular planc operating status. Reference Section 7.7.1 FSAR.

The manual scram function is active in all modes, i

inserting control rods during all modes of reactor operation.thus providing for a manual means of rapidly

'1he IRM systein provides protection against excessive power levels and short reactor periods in the 1

e 3.1 BASES 37

. REV c._~ _ - _ - _ . _ . . ..__.___m _ _ _ _2t_ _ _ _ _ _ _

Table 3.1.1 - Continued 6.

Seven inches on the water level instrumentation is 10'6" above the top of the active fuel at rated power.

7. Trips upon loss of oil pressure to the acceleration relay.

8. Limited trip setting refers to the volume of water in the discharge volume receiver tank and does not include the volume in the lines to the level switches.

9. liigh reactor pressure is not required to be operable when the reactor vessel head is unbolted.-

Reautred Conditions when minimum conditions for operation are not satisfied. (ref. 3.1.B) 1 A. All operable control rods fully inserted. %d B. Power on IRM range or below and reactor in Startup, Refuel, or Shutdown mode.

O C. Reactor in Startup or Refuel mode and pressure below 600 psig.

fl D. Reactor power less than 45% (764-+ MWt.).

??g,75 v.9 D

- Allowable Evnass Conditions FI It is permissible to bypass:

a.

The scram discharge volume fligh Water Level scram function in the refuel mode to allow reactor protection system reset. A rod block shall be applied while the bypass is in effect.

b.

The Low Condenser

^

+

is below 600 psig. vacuum and MSIV closure scram functions in the Refuel and Startup modes if reactor pressure p c. Deleted.

Il d.

The turbine stop valve closure and fast control valve closure scram functions when the reactor thermal power

{ is s 45% (M1.+ MWt) .

778.7s M

3.1/4.1 %M NASt-5 fopod bs/ My6,1M W@ kb' 30

- _ . _ - _ _ _ _ _ _ . - _ _ _ _ - _ _ . _ _ _ . - - _ . _ _ . - - - _ - - - _ . _ _ _ _ . _ - - - _ - _ . _ _ _ - . . - _ . _ _ _ . - - - - - _ . - . _ - - - _ _ _ . _ _ _ . _ _ _ _ - - - _ - _ _ . _ _ . , _ _ _ _ - - _ _ - _ - - -_____e - - -a

Bases Continued: ~

meeting their criterion.

To raise the ECCS' initiation setpoint would be'in a safe direction, but it would duringteduce normally theexpected margin established transients. to prevent actuation of the ECCS during normal operation or The top of the active fuel.

operator will set the low low water level ECCS initiation trip setting >,6'6" <6'10" above the _

Ilowever, the actual setpoint can be.as much as 3 inches lower than the 6'6" setpoint and 3 inches greater than the 6'10" setpoint due to the deviations discussed on page 39.

E. Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided d to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass. This transient y is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin 4 exists. p 45% ratedSpecific thermal power. analyses have generated specific limits which allow this scram to be bypassed below

. In order to ensure the availability of this scrare above 45% rated thermal C-power, this scram is only bypassed below 30% turbine first stage pressure. This takes'into account the possibility of 464 IW, power being passe rectly to the condenser through the bypass valves, d ne.c %l F. Turbine Ston Valve Scram The turbine-stop valve closurepower a s. M&aM b y scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 10% of valve closure from full' open, the resultant increase in surface heat flux is limited such that MCPR remains above the Safety Limit (T.S.2.1. A) even during the worst case transient that assumes the turbine bypass is closed. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability u.)

,)

of this scram above 45% rated thermal power, this scram is only bypassed belowAturbine first stage p pressure. This takes into account the possibility of .M1 power bein ssea directly to the condenser through the bypass valves. I'@

%ecut po%c e LA;Me.A by C. Main Steam Line Isolation Valve Closure Scram The main steam line isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation closure. With the scram set at 10% valve closure there is no increase in neutron flux.

11. Main Steam Line Low Pressure Initiates Main Steam Tsolation Valve Closure The low pressure isolation of the main, steam lines at 825 psig was provided to give protection against rapid reactor derressurization and the resulting rapid cooldown of-the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel 2

cladding integrity safety limit. Operation at steamline pressures lower than 825 psig requires 4

2.3 BASES '

19 REV

Bases Continued:

3.1 start-up and intermediate power ranges.

Ref. Section 7.4.4 FSAR. A source range monitor (SRM) system no scramisfunctions.

also provided to supply additional neutron . level information during start-up but has "Startup" modes. Ref. Section 7.4.3 FSAR. Thus, the IRM is required in the " Refuel" and 7.4.5.2 FSAR. Thus, thepower In the IRM range the APRM system provides required protection. Ref. Section system is not required in the "Run* mode. The APRM's cover only the power range, fore, the IRM's provide adequate coverage in the start-up and intermediate' range and there-the APRM's are not required for the " Refuel" or "Startup'" modes. ,

The high reactor pressure, high drywell pressure, and reactor low water level scrams are required for all modes of plant operation unless the reactor is suberitical and depressurized. They are, therefore, required to be operational for all modes of reactor operation except in the " Refuel" mode with the reactor suberitical and reactor temperature less than 212*F as allowed by Note 3.

The scram discharge volume high level trip function is required for all modes with the exception that it may be bypassed in the " Refuel Mode".under the provisions Table 3.1.1, allowable by-pass condition (a).

In order discharge to reset volume thethis to clear safety system scram inputafter a scram condition, it is necessary to drain the scran condition.

matter what the initial cause might have been. This condition usually follows any scram, no Since all of the control rods are completely inserted following a scram it is permissible to bypass thia condition because a control rod block prevents withdrawal as long as the switch is in the bypass condition for this function.

To permit plant operation to generate adequate steam and pressure to establish ~ turbine seals and condenser 600 psig. vacuum at relatively low reactor power, the main condensor vacuum trip is bypassed until Section 7.7.1.2ThisFSAR bypass also applies to the main steam isolation valves for the same reason. Ref.

,4, g g .

ggg gg An automatic bypass!

f the turbine control valve fast closure scram and turbine stop closure scram is effective L.; .;r;r em6

n: -fiest-etage-pressure-le-1.as-than M= -f- i+- --tv --lue This insures that reactor thermal power is less than 45% of its rated value. .

l Closure of these valves from such a low initial power level does not constitute a threat to the integrity'of any barrier to 4 the release of radioactive material. gy a

3.1 BASES 38

REV

-a e

i TABLE 3.2.1 - Continued Hin. No. of Operable Total No. of Instru-' or Operating Instru-ment Channels Per ment Channels Per Trip Required Function Trio Settines Trio System System (1.2) Conditions

b. liigh Drywell Pressure $2 psig 2 2

D-(5)-

3. Reactor Cleanup System (Group 3)

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

b. High Drywell Pressure 52 psig 2 2 E I

4. IIPCI Steam Lines (Croup 4)

a. IIPCI High Steam Flow $150,000 lb/hr 2(4) 2 y with 560 second time delay

b. IIPCI Illgh Steam Flow 5300,000 lb/hr 2(4)~ 2 F

c. IIPCI Steam Line s200*F 16(4) 16 Area High Temp. F

5. RCIC Steam Lines (Group 5)

a. RCIC High Steam Flow 545,000 lb/hr 2(4) - g WI with 5 1 2 see time delay

b. RCIC Steam Line Area' s200*F. 16(4) 16 C

6. Shutdown Cooling Sunolv Isolation

a. Reactor Pressure 575 psig 2(4)' 2 C Interlock at pump +be. $ -

,% vtar h e c

$iCh 6 O ets L. d 4

Wn C.hage 5 ptopo54d hf_ T y 6 jl 9 pg AM trP le.lf cc, d

Table 3.2.2 Instrumentation That Initiates Emergency, Core Cooling Systems Minimum No. of Oper-Minimum No. of able or Operating Operable or Operating Trip Total No. of Instru- Instrument Channels ment Channels Per Per Trip System Required Function Trin Settine Systems (3.6) Trio System B. HPCI System (3.61 Conditions

Ml

1. High Drywell s2 psig 1 4 4 Pressure (1) A.

2. Low-Low Reactor 26'6"s6'10" 1 4 Water Level 4 A.

C. Automatic Depres-surization

1. Low-Low Reactor 26'6"s6'10" 2 2 Water Level 2 B.

and

2. Auto Blowdown s120 seconds 2 1 Timer 1 B.

and =

4

3. Low Pressure Core $100 7:'; 2 d 12(4) 12(4)

Cooling Pumps Dis-Charge Pressure 2 bopd y B. {

Interlock I

.5 ITo y s*1 I

. d 3.2/4.2 .:f. C.Wgs-S. f'Of & b[ K \[ S>\117 HdirP \53* #"-

s Bases Continued:

3.2 This trip setting level was chosen to be low enough to prevent spurious operation butnough hightoe ,

initiate ECCS operation and primary system isolation so that no melting of the fuel cladding occur will '

and so. that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will enot b violated.

For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, above cr'.teria. ECCS initiation and primary system isolation are initiated in time to meet the !

Reference Section 6.2.7 and 14.6.3 FSAR.

range or spectrum of breaks and meets the above criteria. The instrumentation also covers the full' :

Reference Section 6.2.7 FSAR. -

The addition highto drywell pressure initiating ECCSinstrumentation it causes is a back-up to the water level instrumentation and in i complete circumferential break discussed above, isolation of Croup 2 and Croup 3 isolation valves. For the about the same this instrumentation will initiate.ECCS operation at !

applicable here time also.as the low low water level instrumentation; thus the results given above are i

i isolation, and RUCU system valves. Group 2 and Croup 3 isolation valves include the drywell vent, purge, sump i I

Two the four pressure switches are provided on the discharge of each of the two core spray pumps and each of RHR pumps. I can permit automatic depressurization.Two trip systems are provided in the control logic such that either trip system both trip logic channels are required to permit a system trip.Each trip system consists of two trip logic channej I

Division IIIcore Division spray and permissives willRllRinterlock pump the discharge pressure other trip system. permissives will interlock one trip system and j i One pressure switch on each pump will j

.within interlock theirone of the trip respective tripchannels system. and the other pressure switch will interlock the other trip channel

[

i l

The pump pressure permissive control. logic is designed such that no single failure (short or open I circuit) will prevent auto-blowdown or allow auto-blowdown when not required. !

Venturis are provided in the main steamlines as a means of measuring steam flow and also limiting the of mass inventory from the vessel during a steamline break accident. -

ss lo

/ In addition to monitoring steam flow, ;

T he. M yo se.h 3 4oe +be low p'e.ssu.ce. E'4 5 P t

- Y loe Aos ;s 54 + sac.k %e4 1Y i s. \a,ss -S aa 64. r P*' a '55iva-pup disdafge p /s ssar e. . v/ he.n. cs p.- p is opedig in a b il Nos wdit. on o.ad j 3.2 BASES 5' 9 "[ "^ # '

65 iA a d isc1A.r34 . pMs*xAfa- petel5SIV4- . W ben Md. f '^Y

REV uc. w o+ opewt. ,

I

3.0 LIMITING CONDITION FOR OPERATION 4.0 S'JRVEILIANCE REQUIREMENTS C. Containment C ,._j/ Cooling System C. Containment C;.. ,/ Cooling System

1. Except as specified in 3.5.C.2 -n A ^_below, 1. Demonstrate the RilR Service Water pumps both Containment Spr;7/ Cooling Subsystems develop 3,500 gpm flow rate against a 500 shall be operable whenever irradiated fuel is ft head when tested pursuant to in the reactor vessel and reactor water Specification 4.15.B.

temperature is greater than 212*F. A containment /rpr:7 cooling subsystem consists 2. Test the valves in accordance with

, d tr1 of the following equipment powered from one Specification 4.15.B.

division: f

2. ":::nctr :: :L; _,. . Lili;., J 0. A ,. 11 g

.14 RilR Service Water Pumpf :pr:y 5 :d: : and.~; ale 10. .. L .n 1 6 Exchan during . d S , c... i_d.

19- RilR Pumpp* geURHN Valves and piping necessary fot:

Torus Cooling L, 11 C,.r , -

2. O n e "d'" C e rri: : " ster Pump y be i- eperehle-1 far 20 d=yz.

y 3. On: "" Ecrrier "eter ? p i-. r e a- - bey-*--

( :7 5: in:p:r:bl f: r d , .

2.-+. one Containment pr y/ Cooling Subsystem may be J inoperable for 7 days. oc 3 -9. If the requirements of 3.5.C. . end '-

cannot be met, an orderly shutdown of the reactor will be initiated and the reactor water temperature shall be reduced to less than 212*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

For allowed out of service times for the RilR pumps see Section 3.5.A.

3.5/4.5 104 REV

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS JE 4

3.5 CORE AND CONTAINMENT 1rn/3/ COOLING SYSTEMS e 4.5 CORE AND CONTAINMENT "PT'J/ COOLING SYSTEMS Anolicability: (

Annlienbility: s g)

Applies to the operational status of the emergency cooling systems, Applies to the periodic testing of the emergency cooling systems.

Obiective:

Obiective: .

To insure adequate cooling capability for heat removal in the event of a loss of coolant accident or To verify the operability of the emergency cooling systems.

isolation from the normal reactor heat sink. ,

Speci fica tion:

Specification:

A. ECCS Systems A. ECCS Systems

1. Except as specified in section 3.5.A.3, both 1.

Core Spray subsystems and the Low Pressure Demonstrate the Core Spray Pumps develop a [

Coolant Injection (LPCI) Subsystem (LPCI Mode 2,800 gpm flow rate against a system head of RllR System) shall be operable whenever corresponding to a reactor pressure of 130

[

irradiated fuel is in the reactor vessel and psi greater than containment pressure, when i the reactor water temperature is greater than tested pursuant to Specification 4.15.B.

212*F. 2. Demonstrate the LPCI Pumps develop a 3,870

2. Except as specified in section 3.5.A.3, the gpm flow rate against a system head liigh Pressure Coolant Injection (IIPCI) System corresponding to two pumps delivering 7,740 and the Automatic Depressurization System gpm at a reactor pressure of 20 psi greater (ADS) shall be operable whenever the reactor than containment pressure, when tested pursuant to Specification 4.15.B.

pressure is greater than 150 psig and irradiated fuel is in the reactor vessel 3. Demonstrate the llPCI Pump develops a 2700 except during reactor vessel hydrostatic or gpm flow rate against a reactor pressure i

leakage tests, range of 1120 psig to 150 psig, when tested pursuant to Specification 4.15.B.

3.5/4.5 101 Rev

i .

%i D e. ~

l so l o.k.o 3 Ved s/cS 0.fe Of4Aod MA-k * -

c,ooldow 40 e d o. bits.k (cc, i re.sd d'8 ^ b '# N0 54[

j Bases 3.5/4.5 continued: R M R- sacA* u liac a.M. /c4u.ca t,ta a.s, O e/s.by a.as.4.c ;

LA M 4cr e c,.ol do a e4 hh p ipi ng .

.-,4

.h automatically 4 for two valves.controls three selected safety-relief valves although the safety analysis only takes credit o It reducing is therefore appropriate to permit one valve to be out-of-service for up tog days without materially system reliability. p/

lk 4

B. RllR Intertie Line An intertie line is provided to connect the RIIR sucti.n line with the two RilR loop return lines. This e The four-inch for water hammer line isinequipped the recirculation with and three RllRisolation system, _.. valveg:r r ; purpose uired t: :::ld:r.ofrith this :- line i :1 is tedto::reduce idi: the potential

-.irculation cy t.x.

"'"' inj ::tlor. 7i ping.

5. i::1stien ::1::: awe-opened-dur-ing-a-oeolder t: :::ure : unifer :::!d:= cf th:

I-f-ene-we e i r cui n t l en4oop4s4 sol a te d-o r-idler-these--valves-an ' seeeelatod-piping g 9e eper ti ; leep t: :::1 the 12:leted-or-141: 1::p. The RllR loop return line isolation valves s.)

receive a closure signal on LPC1 initiation.

In the event of an inoperable return line isolation valve, there is a potential for some accident. Surveillance requirements have been established to periodically cycle the RilR intertie line of the LPCI flow to be diverted to the broken loop during a loss of coolant isolation valves.

valve is closed or the In the event of an inoperable RilR loop return line isolation valve, either the inoperable other two isolation valves are closed to prevent diversion of LPCI flow. The RilR intertie line flow is not permitted in the Run Mode to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a reduction in core flow during a loss of coolant accident.

C. Containment "p. ,/ Cooling Systems i an EHE i Two containment :pr .j/ cooling subsystems of th RilR system are provided a remove heat ener from the containment and control torus and drywell press e in the event of a loss f coolant acciden . A containment .r. ,/ cooling subsystem consists of RIIR Service Water Pump /, Ileat Exchanger, RilR Pumpg, and valves and piping necessary for Torus Cooling nd Sr7 211 ?p- 7 a containment ,ym.j/ cooling subsystem. For the flow specified, th: ::nt:ix;nt TorusSprayp;notconsideredpartof

'n _ .

limit.2 :, 1::: th:n 5 prig 2nd, ther:fer , (s zu tic-- _.i.1. to provide the r quired heat removal capability. Reference Section 6.2.3.2.3. USAR.

y q,se.k S A sysk m K o.de. b ud . O<pc.ti spey are.

co M osO t.ooling}

Either subsystem 5 i= < ap'able of performing the containment spuay/ cooling function. Loss of 'one RilR service water pump 'does not seriously jeopardize the containment prey / cooling capability as :_ cf t'.: _..i. -

three perp: can satisfy the cooling requirements. Sin:: th:r: i: r u r:^end:::7 1:ft, 1 ?? fry reprir p:ried i; :d ;u;t:. Loss of (

<:ontainment ..p:y/ cooling subsystem leaves one remaining system to perform

[ the containment rprey/ cooling unction. % b o g g, 5J A, f

.I 3.5/4.5 Bases e et, 112 REV t

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

~

s i

Bases Continued:

Vent watersystem level. downcomer submergence is three feet below the minimum specified suppression pool-This length has been shown to result in reduced postulated accident loading of the torus whileatthesametimeassuringthedowncomersremainsubmegdunderallseisat.candaccident conditions and possess adequate condensation effectiveness.

The mar.inum temperr.ture at the end of blowdown tested during the Humboldt Bay (1) and Bodega Bay (2) tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170*F. -

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions at the discharge exit. Specifications have been placed on the envelope of reactor operating of potentiallyconditions so that thechamber high suppression reactor can be depressurized in a timely manner to avoid the regime loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open.. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers. (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve tot Rassure y g. p e p in 5 arP<cssion Paal c oo t modaoA1ME mixing and uniformity of energy insertiob,to hadiethe pool.For the neTael com-plement of containment cooling pumps (2 1."Ci y ,: - ' " - -'-8-- '

l'- cervice water pump /)

nteir_::nt pr ::rr: in n:t r:;uir:d t: ::intri cf:; ret: :t p::itir; ;;_ ti;..

r; apr:7, ' ?! :nd ""! r t__2 C:."r::) f._ ;:.

heur efter : 1 :: ef :::1:nt r; -

I!!L;h;nid

id;nt, CI. d:ri 5Pd'" :: OFFrt:150t17 10 p i; ir.e

307 F ri;d ;t rtins ; f;W 1 x:nt pr:::xr: 5: r:du :d t: st- _;;hcric prc:;;;; thr;;;h _.., n:;;;p;ratic sad sh;uld the c;..tain-2512 Sin : : :ntr:n;17 degended ;;,ndition aust _ist, the peried

, ;d;;;;t; ""C" ns id n;; he ;c il-of vulse.;;Lili;. Le thi. .....i is y

-re trict:d by Sp::ific ti;n 2.'.."..l.5 O y,. ret .:. 2nd th; p;ried of p .rati n nith ca; ir.;perchi; ty P"",1;;p.

liniting the. gfi4 ;;k;;r; ;icr p;;l initici ; k hea,A we , ovie.quodc. na posthe_ s a c- ti gg ofw e d(.g ns [.}-; o M is e. h m 4 in me. s i+% Aid

-frar k h a ( o re. 5 p 4 y RHR G.nd H P/- I P*PS MW uc beme.M e 4 c odadaswee4 g (1) Robbins, C.H. " Tests of Full Scale 1/48 Segment of the Humboldt Bay Pressure d Suppression Containment," CEAP-3596, November 17, 1960. (" "

4 '^ ' " *

(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(3) Ceneral Electric NEDE-21885-P, " Mark 1 Containment Program Downcomer Reduced Submergence Functional Assessment Report", June, 1978.

3.7 BASES '

176 REV

s Page 3.4 and 4.4 Standby Liquid Control System 93 H

g Sys4L" s,1,4;o, h a;,repeh5 [

A.

B.

L 21 Operation

[

c-- +- n :==;.. u ; c. vu .. .u 3 93 7 95 s q

C. '.';ir C:n:: ::::i;n ".;quirese..ts jJP% d 3.4 and 4.4 Bases 99 7 3.5 and 4.5 Core and containment /r;rg Cooling Systems t) 101 .p A. ECCS Sy544 ^S f 101 l j< d, B. RHR Intertie Return Line Isolation Valves 103 l C.

D.

Containment C RCIC r . ,/0ooling System 104 gl SQ 105 E. Cold Shutdown and Refueling Requirements 106 F. Recirculation System 107 H n

3.5/4.5 Bases llo 4 G9- IE 4

3.6 and 4.6 0 Primary System Boundary 121 A. Reactor Coolant Heacup and Cooldown 121 B. Reactor Vessel Temperature and Pressure 122 C. Coolant Chemistry 123 D. Coolant Leakage 126 E. Safety / Relief Valves 127 F. Deleted G. Jet Pumps lab H i 128 g ;

M. Snubbers 129 cn C

3.6 and 4.6 Bases l'/I% 44 ]

3.7 and 4.7 O Containment Systems 156 A. Primary Containment 156 B. Standby Cas Treatment System 166 ti C. Secondary Centainment 169 D. Primary Containment Isolation Valves J 170 0 E. Combustible Gas Control System I'72. 4.4Me- (

3.7 Bases 175 d 4.7 Bases 183 11 REV

\

i l

I

2151 3.8 and 4.8 Radioactive Effluents 192 A. Liquid Effluents 192

B. Gaseots Effluents 197 l

l C. Solid Radioactive Wasta 198e l Dose from All Uranium Fuel Cycle Sources D. 198f '

'3~.8 and 4.8 Bases 198u ,

i. 3.9 and 4.9 Auxilliary Electrical Systems '199 A. Operational Requirements for Startup 199 {

B. Operational Requirements for Continued Operation 200

^

1. Transmission Lines 200 ,

2. Reserve Transformers 201 .H

3. Standby Diesel Generators 201 $, . j '

4 Station Battery System 2 03'292- C l l F. AVV S atic y sys,te,e Ao3 j ,

3.9 Bases 204

~

O l 4~.9 Bases 205 t

, 3.10 and 4.10 Refueling 206. I A. Refueling Interlocks 206 B. Core Monitoring 207 j C. Fuel Storage Pool Water Lavel 207 D. Movement of Fuel 207 w E. Extended Core and Control Rod Drive Maintenance 208 l 3.10 and 4.10 Bases 209 '

3.11 and 4.11 Reactor Fuel Assemblies 211 A. Average Planar Linear Heat Generation Rate 211 ;

B. Linear Heat Generation Rate 212. i C. . Minimum Critical Power Ratio 213 H y j v2 -

3.11 Bases 216 C ,

4.11 Bases a1S ^ tit I f' ]

0 '

3.12 and 4.12 Sealed-Source Contamination '

219 A. Contamination 219 B. Records 221 l

3.12 and 4.12 Bases . 222 l h

1 I

l iii REV

EXHIBIT C LICENSE AMENDMENT REQUEST DATED JULY 26,1996 REVISED OPERATING LICENSE AND TECHNICAL SPECIFICATION PAGES i

i NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT ;

MONTICELLO, MINNESOTA i

LICENSE NO. DPR-22 i DOCKET NO. 50-263

Exhibit C Monticello Nuclear Generating Plant License Amendment Reauest dated July 26.1996 Revised Operating License and Technical Specification Pages Exhibit C consists of the Operating License and Technical Specification pages with the proposed changes incorporated. Existing pages affected by this change are listed below.

Affected Page Associated Change Affected Page Associated Chance Ucense page 3 A 30" E 4 A 19 E 14 A 38 E 15* A 50" F 6 B 53 " G 56 B M G 16 B 104 H 114* B 101 H 23 C- 112 H,1 24 C 176 H 150 C ii H,i 28 D iii I 37 D Pages reflect proposed changes submitted with amendment request entitled

" Implementation of BWR Thermal Hydraulic Stability Solution," dated June 22,1995, TAC M92947.

Pages reflect proposed changes submitted with amendment request entitled " Surveillance Test Interval / Allowed Outage Extension Program - Part 2," dated July 5,1995, TAC M92948.

I

1 l

l

)

I

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This license shall be deemed to contain and is subject to the l conditions specified in the Commission's regulations in 10 CFR '

' Chapter I and is subject to all. applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level The licensee is authorized to operate the facility at steady g state reactor core power levels not in excess of 1775 mega- .

watts (thermal) j u

i 2. Technical Snecifications @

4 The Technical Specifications contained in Appendix A as- g revised through Amendment No. 96'are hereby incorporated in the $

license. The licensee shall operate the facility in accordance with the Technical Specifications.

4

3. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments ,

and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and e 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). g The plans, which contain Safeguards Information protected under 10 CFR 2

, 73.21, are entitled: "Monticello Nuclear Generating Plant Physical y 4 Security Plan," with revisions submitted through November 30, 1987; g "Monticello Nuclear Generating Plant Guard Training and Qualification j Plan," with revisions submitted through February 26, 1986; and *

"Monticello Nuclear Generating Plant Safeguards Contingency Plan," 4 with revisions submitted through August 20, 1980. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with !

l the schedule set forth therein.

I i

1

i l' !

4.

Protective Mmetion - A system protective ' action which results from the protective' action of - i

the channels monitoring a particular plant condition.

i R. Rated Neutron Flux - Rated flux is the neutron flux that corresponds to a steady-state power level of.1775 thermal megawatts..

S. Rated Thermal Power - Rated thermal power means a steady-state power level of 1775 thermal megawatts.

T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures. listed in the Technical Specifications are those existing in the vessel steam space.

U. Refueline Operation and Refuelinz Outare - Refueling Operation is any operation when the reactor water temperature is less than 212*F and movement of fuel or core components is-in progress. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the regra. red surveillance testing need not be performed until'the next regularly scheduled outage.

V. Safety Limit - The safety limits are limits below which the maintenance of the' cladding and primary system integrity are assured. Exceeding such a limit is cause for plant shutdown and review by the Commission before resumption of plant' operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

W. Secondary Containment Interrity - Secondary Containment Integrity means that the reactor building is

, closed and the following conditions are met:

1. At least one door in each access opening is closed.

2. The standby gas treatment system is operable.

3. All reactor building ventilation system automatic isolation valves are operable or are secured in the closed position.

X. Sensor Check - A qualitative determination of operability by observation of sensor behavior during operation. This determination shall include, where possible, comparison with other independent.

sensors measuring the same variable.

1.0 4 4 REV -

_ _ _ _ _ - - = _ _ , __- . _ - - . __. _. -- _ - . - . . . .

BASES:

2.3 The abnormal operational transients applicable to operation of the Monticello Unit have been analyzed throughout the spectrum of planned operating conditions up to the thermal power level of 1775 MWt. l The analyses were based upon plant operation in accordance with the operating map. The licensed ..

maximum power level 1775 MWt represents the maximum steady-state power which shall not knowingly be exceeded. l Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power' shapes. These factors are selected conservatively with respect to their effect on the applicable tranisent results as determined by the current analysis model.

2.3 Bases 14 REV

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

a Bases Continued: -

For analyses of the thermal consequences of the transients, the Operating MCPR Limit (T.S.3.11.C) is i conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Deviations from as-left settings of setpoints are expected due to inherent instrument error. operator setting error, drift of the setpoint, etc. Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discussed above. Although the operator will set the setpoints within the trip settings specified, the actual values of the various setpoints can vary from the specified trip setting by the allowable deviation.

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power

, (1775 MWt). Because fission chambers provide the basic input signals, the APRM system responds l directly to average neutron flux. During transients, the instantaneous rate of heat transfer from ,

the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time l constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the '

fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 1201 scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage. Also, the flow biased neutron flux scram (specification 2.3.A.1) provides protection to the fuel safety i limit in the unlikely event of a thermal-hydraulic instability.

i 2.3 BASES 15 REV

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1 WEL CIADDING INTEGRITY 2.3 FUEL CIADDING INTEGRITY Aeolicability Aeolicability Applies to the interrelated variables associated Applies to trip settings of the instruments and

with fuel thermal behavior, devices which are provided to prevent the reactor system safety liaits from being exc'eeded.

Obiective: Obiective:

To establish limits below which the integrity To define the level of the process variables at of the fuel cladding is preserved, which automatic protective action is initiated to prevent the safety limits from being exceeded.

Specification: Specification:

A. Core Thermal Power Limit (Reactor Pressure >800- The Limiting safety system settings shall be as psia and Core Flow is >10% of Rated) specified below:

When the reactor pressure is >800 psia and core A. Neutron Flux Scram flow is >10% of rated, the existence of a minimum critical power ratio (MCPR) less than 1. APRM - The APRM flux scram trip setting 1.07, for two recirculation loop operation, or shall be:

less than 1.08 for single loop operation, shall constitute violation of the fuel cladding a. For two recirculation loop operation integrity safety limit. (TID):

S $ 0.66W + 65.6% where l S - Setting in percent of rated thermal power, rated power being 1775 MWT W - Percent of recirculation drive flow required to produce a core flow of -57.6 x 10' lb/hr i

b. For single recirculation loop operation ,

(SID): !

SS 0.66(W - 5.4) + 65.6%. l

c. No greater than 120%.

2.1/2.3 6 ,

REV l

ma ... ____ u m._:m -._ __._.-.-..__.-a__- . - . - m.,._:_ _.a._u._..--.-_.___-- _ . _ _ _ _ _-_..-mm :. ___A-__c ___--mm_-_m _ _ _ _ -_ _ _ _ . _ . _ _ . 'as a____w-

m.:'s _ a _ __._m m _e __ A .._ +m- ' ' _ _ _ a. A_* _

o-l TABLE 3.2.3

, Instrumentation That Initiates Rod Block Reactor-Modes Which Function Must be Operable Total No. of Min. No. of Oper-or Operating and Allow-i Instrument .able or Operating able Bvoass Conditions ** Channels per- Instrument Channels Required Function Trio Settinas Refuel Startuo Run Trio System 'oer Trio System Conditions *

1. s.Eli

a. Upscale 55x10s cpa X- X(d) 2 1(Note-1, 3, 6) A or B or C

b. Detector X(a) X(a) 2 1(Note 1, 3, 6) A or B.or C-not fully inserted.

2 1 811

a. Downscale- h3/125 X(b) .X(b) 4" 2(Note 1, 4, 6) A or B or C full scale

b. Upscale 5108/125 X X' 4 2(Note 1, 4, 6) A or B or C full scale

3. AI'E!!

a. Upscale X 3 1(Note 1, 6, 7) D or E '

(1) TID Flow $ 0.66W + 53.6%

Biased (Note 2)

(2) SIA Flow $ 0.66(W - 5.4) + 53.6%

Biased (Note 2)

(3) High $ 1081 Flow Clamp

b. Downscale k 3/125 full scale X .3 1(Note 1, 6, 7) D or E 4

3.2/4.2 56 i REV

Bases Continued:

Maximum Extended Load Line Limit Analyses (MELLLA) have been performed to allow operation at higher ;

powere at flows below 87%. The flow referenced scram (and rod block line) have increased (higher slope '

and y-intercept) for two loop operation (See Core Operating Limits Report). The supporting analyses are discussed in GE NEDC-31849P report (

Reference:

Letter from NSP to NRC dated July 7,1996). l Increased Core Flow (ICF) analyses have been performed to allow operating at flows above 100% for powers equal to oc less than 100% (See Core Operating Limit Report). The supporting analyses are discussed in General Electric NEDC-31778P report (

Reference:

Letter from NSP to NRC dated July 7,1996).

Evaluations discussed in NEDC-32546P, July 1996, demonstrated the acceptability of MELLLA and ICF for ,

rerate conditions. In addition, the evaluation demonstrated the acceptability of MELLLA for single loop '

operation.

For operation in the startup mode while the reactor is at low pressure, the IRM scram setting of 20% of rated power provides adequate thermal margin between the setpoint and the safety limit, 25% of rated. l The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects ,

of increasing pressure at zero or low void content are minor, cold water from sources available during !

startup is not much colder than that already in the system, temperature coefficients are small, and I control rod patterns are constrained to be uniform by operating procedures. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow.

Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per '

minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The IRM scram remains' active until the mode switch is placed in the run position and the associated APRM is not downscale. This switch occurs when reactor pressure is greater than 850 psig.

The operator will set the APRM neutron flux trip setting no greater than that stated in Specification 2.3.A.1. However, the actual setpoint can be as much as 3% greater than that stated in Specification 2.3.A.1 for recirculation driving flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on page 39.

B. Deleted 2.3 BASES 16 REV Next Page is 18

i l

~!

t

! . l Bases 3.5/4.5 continued: ,

F. Recirculation System j The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected t and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal *

[ hydraulic instability, a power-flow exclusion region, to be avoided during normal operation, is calculated [

using the approved methodology as stated in specification 6.7.A.7. Since the exclusion region may change i each fuel cycle the limits are contained in the Core Operating Limits Report. Specific directions are [

provided to avoid operation in this region and to immediately exit upon an entry. Entries into the i exclusion region are not part of normal operation. An entry may occur as the result of an abnormal event l such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed [

to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though ;

operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow biased -

scram function will suppress oscillations prior to exceeding the fuel safety limit.

Fower distribution controls are established to ensure the reactor is operated within the bounds of the stability analysis. With these controls in place, there is confidence that an oscillation will not occur [

outside of the stability exclusion region. Without these controls, it is theoretically possible to operate ;

the reactor in such a manner as to cause an oscillation outside of the exclusion region. A nominal 5%- 1 power-flow buffer region outside of the exclusion region is provided to establish a stability margin to' the l analytically defined exclusion region. The buffer region may be entered only when the power distribution t

. controls are in place.

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specification ;

3.5.F.3 were determined by NEDO-24271, June 1980, "Monticello Nuclear Generating Plant Single Imop l Operation;"NEDC-30492, April 1984, " Average Power Range Monitor, Rod Block Monitor and Technical l Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant;" and NEDC-32456P, July l 1996. Specification 3.6.A.2 governs the restart of a recirculation pump in an idle recirculation loop. ,

Adherence to this specification limits the probability of excessive flux transients and/or thermal stresses. [

I i

l

'NEXT PAGE IS 121 3.5/4.5 114 i REV ;

i t

Bases Continued:

2.2 The normal operating pressure of the reactor coolant system is approximately 1010 psig.

Evaluations have determined that the most severe pressure transient is bounded by the closure of all MSIVs, followed by a reactor scram on high neutron flux (failure of the direct scram associated with MSIV position is assumed). The USAR discusses the analysis of this event. .The analysis results demonstrate the safety / relief valve capacity is capable of maintaining pressure below the ASME Code limit of 110% of vessel' design pressure (1101 X 1250 psig - 1375 psig). The safety limit ensures that the acceptance limit of 1375 psig is met during the design basis event at the vessel' location with the highest pressure.

1 b

1 f

t 2.2 BASES 23 REV

._ -- - _ . - . . _ _ . . . - . . - _ _ . . _ . , , . - . _ - _ . . . - __.....____._..__----.--_.-._---.___-_-._.,-__-----_w-_________.__.-___ _.____n-__ . - - _ - _ - _ _- - - _ _ - . _ _ _ _ _ _ _ _ - . _ _ - _ _ - - . = - . , . _ _ _ _ _ . - - - _ .

i I

Bases:

2.4 The settings on the reactor high pressure scram, reactor coolant system safety / relief valves, turbine control valve fast closure scram, and turbine stop valve closure scram have been established to assure never reaching the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit.

The APRM neutron flux scram and the turbine bypass system also provide protection for these safety limits. In addition to preventing power operation above 1075 psig, the pressure scram backs up the APRM neutron flux scram for steam line isolation type transients.

I The reactor coolant system safety / relief valves assure that the reactor coolant system pressure

, safety limit is never reached. In compliance with Section III of the ASME Boiler and Pressure Vessel Code,1965 Edition, the safety / relief valves must be set to open at a pressure no higher than 105 percent of design pressure, with at least one safety / relief valve set to open at a pressure no greater than design pressure, and they must limit the. reactor pressure to no more than 110 percent of design pressure. .The safety / relief valves are sized according to the Code for a condition of MSIV closure while operating at 1775 MWt, followed by no MSIV closure scram but scram from an indirect (high flux) means. With the safety / relief valves set as specified herein, the maximum vessel pressure remains below the 1375 psig ASME Code limit. Only five of the eight valves are assumed to be operable in this analysis and the valves are assumed to open at 31 above their setpoint of 1109 psig with a 0.4 second delay. The upper limit on safety / relief valve setpoint is established by the design pressure of the HPCI and RCIC systems. The design capability of the HPCI and RCIC systems has been demonstrated to be acceptable at pressures 3% greater than the !

safety / relief valve setpoint of 1109 psig. l l

The operator will set the. reactor coolant high pressure scram' trip setting at 1075 psig or lower.

However, the actual setpoint can be as much as 10 psi above the 1075 psig indicated set point due- I

, to the deviations discussed in the basis of Specification 3.1. In a like manner, the operator will !

set the reactor coolant system safety / relief valve initiation trip setting at 1120 psig (1109 psig !

+ 11) or lower. However, the as-found set point can be as much as 22.3 psi above the 1120 psig indicated set point due to the deviations discussed in the basis of Specification 3.6.

A violation of this specification is assumed to occur only when a device is knowingly set outside of l the limiting trip setting, or when a sufficient number of devices.have been affected by any means. l l

i i

2.4 BASES 24 ;

REV i

i f

Bases Continued 3.6 and 4.6:

D. Coolant Imakare The allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining leakage was also considered. The evidence obtained from experiments suggests that for_ leakage somewhat greater than that specified for unidentified leakage, the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be Pressure Boundary Leakage and they cannot be reduced within the allowed times, the reactor will be shutdown to allow further investigation and corrective action.

Two leakage collection sumps are provided inside primary containment. Identified leakage is piped from the recirculation pump seals, valve stem leak-offs, reactor vessel flange leak-off, bulkhead and bellows drains, and vent cooler drains to the drywell equipment drain sump. All other leakage is collected in the drywell floor drain sump. Both sumps are equipped with level and flow transmitters connected to recorders in the control room. An annunciator and computer alarm are provided in the control room to alert operators when allowable leak rates are approached. Drywell airborne particulate radioactivity is continuously monitored as well as drywell atmospheric temperature and pressure. Systems connected to the reactor coolant systems boundary are also monitored for leakage by the Process Liquid Radiation Monitoring System.

The sensitivity of the sump leakage detection systems for detection of leak rate changes is better than one gpa in a one hour period. Other leakage detection methods provide warning of abnormal leakage and are not directly calibrated to provide leak rate measurements.

E. Safetv/ Relief Valves Testing of the safety / relief valves in accordance with ANSI /ASME OM-1-1981 each refueling outage ensures that any valve deterioration is detected. An as-found tolerance value of 3% for safety / relief valve ,

setpoints is specified in ANSI /AEME OM-1-1981. Analyses have been performed with the valves assumed to open at 3% above their setpoint of 1109 psig. As discussed in the Section 2.2 Bases, the 1375 psig Code limit is not exceeded in any case. i 3.6/4.6 BASES 150 REV 4

_ _ . _ . _ - _ . _ _ _ - - _ . _ _ ._ ___-___._._.__...___.__.__________.___.___.._._.__.___-__._____.-_m _ __.__________.__m____ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _.__m_________

! .-l 2

I TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRIMENT REQUIREMENTS Modes in which tunc. Total No. ot Min. No. ot Operable

. tion must be Oper- Instrument- or Operating Instru-Limiting able or Operating ** Channels per ment Channels Per Required Trip Function Trip Settings Refuel (3) Startup Run Trip System- Trip System (1) Condition *

1. Mode Switch in I Shutdown X X X 1 1 A

2. Manual Scram X X X 1 1 A

3. Neutron Flux IRM s 120/125 ,

(See Note 2) of full scale X X 4 3 A

a. High-High

b. Inoperative -

4. Flow Referenced See Specifi-Neutron Flux APRM cations (See Note 5) 2.3A.1 X 3 2 A or 8

a. High-High

b. Inoperative

c. High Flow Clamp < 120 1

5. High Reactor Pressure s 1075 psig X X(f) X(f) 2 2 A

.(See Note 9)

6. High Drywell Pressure s 2 psig X X(e,f) X(e.f) 2 2 A (See Note 4)

7. Reactor low Water Level 2 7 in.(6) X X(f) .X(f) 2 2 A

8. Scram Discharge Volume High Imvel

a. East s 56 gal.(8) X(a) X(f) X(f) 2 2 A'

b. West s 56 gal.(8) X(a) X(f) X(f) 2 2 A

9. Turbine Condenser Iow Vacuum E 22 in. Hg X(b) X(b,f) X(f) 2 2 A or C 3.1/4.1 28 REv

_ _ _ _ _ _ . . _ - . _ . . . . _ _ _ _ . . - o _-

l I

I l

l Bases Continued:

i 3.1 condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which j eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a I

pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient without bypass. Reference FSAR Section 14.5.1.2.2 and supplemental information submitted February 13, 1973. The condenser low vacuum scram is a back-up to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe. Scram occurs at 22" Hg vacuum, stop l valve closure occurs at 20" Hg vacuum, and bypass closure at 7" Hg vacuum.

The main steamline isolation valve closure scram is set to scram when the isolation valves are $10%

closed from full open. This scram anticipates the pressure and flux transient, which would occur when the valves close. By scramming at this setting the resultant transient is insignificant.

Reference Sectica 14.5.1.3.1 FSAR and supplemental information submitted February 13, 1973.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status. Reference Section 7.7.1 FSAR.

The manual scram function is active in all modes, thus providin6 for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system provides protection against excessive power levels and short reactor periods in the 3.1 BASES 37 REV

Table 3.1.1 - Continued i

6. Seven inches on the water level instrumentation is 10'6" above the top of the active fuel at rated power.

7. Trips upon loss of oil pressure to the acceleration relay.

8. Limited trip setting refers to the volume of water in the discharge volume receiver tank and does not include the volume in the lines to the level switches.

9. High reactor pressure is not required to be operable when the reactor vessel head is unbolted.

Reauired Conditions when minimum conditions for coeration are not satisfied. (ref. 3.1.B)

A. All operable control rods fully inserted.

B. Power on IRM range or below and reactor in Startup, Refuel, or Shutdown mode.

C. Reactor in Startup or Refuel mode and pressure below 600 psig.

D. Reactor power less than 45% (798.75 MWt.).

- Allowable Bvoass Conditions It is permissible to bypass:

a. The scram discharge volume High Water 14 vel scram function in the refuel mode to allow reactor protection system reset. A rod block shall be applied while the bypass is in effect.

b. The low Condenser vacuum and MSIV closure scram functions in the Refuel and Startup modes if reactor pressure is below 600 psig.

c. Deleted.

d. The turbine stcp valve closure and fast control valve closure scram functions when the reactor thermal power is s45% (798.75 MWt).

l 3.1/4.1 30 REV t

Bases Continued:

meeting their criterion. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during -

normally expected transients.

The operator will set the low low water level ECCS initiation trip setting 2:6'6" s6'10" above the top of the active fuel. However, the actual setpoint can be as much as 3 inches lower than the 6'6" setpoint and 3 inches greater than the 6'10" setpoint due to the deviations discussed on page 39. . ~

E. Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass. This transient is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin exists.

Specific analyses have generated specific limits which allow this scram to be bypassed below 451 rated thermal power. In order to ensure the availability of this scram above 45% rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14% power being passed directly to the condenser through the bypass valves.

F. Turbine Stoo valve Scram Tha turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With'a scram trip setting of 10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the Safety Limit (T.S.2.1.A) even during the worst case transient that assumes the turbine bypass is closed. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power. In order to ensure the availability of this scram above 451 rated thermal power, this scram is only bypassed below 30% thermal power as indicated by turbine first stage pressure. This takes into account the possibility of 14% power being passed directly to the condenser.through the bypass valves.

G. Main Steam Line Isolation Valve closure Scram The main steam line isolation valve closure scram anticipates the press.ure and flux transients which occur during norac1 or inadvertent _ isolation closure.

With the scram set at.10% valve closure there is no increase in neutron flux.

H. Main Steam Line low Pressure Initiates Main Se== Isolation Valve Closure The low pressure isolation of the main steam lines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation at steamline pressures. lower than 825 psig requires 2.3 BASES 19

. g t

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

Bases Continued:

3.1 start-up and intermediate power ranges. Ref. Section 7.4.4 FSAR. A source range monitor (SRM) system is also provided to supply additional neutron level information during start-up but has no scram functions. Ref. Section 7.4.3 FSAR. Thus, the IRM is required in the " Refuel

and "Startup" '

modes. 'In the power range the APRM system provides required protection. Ref. Section 7.4.5.2 FSAR.

Thus, the IRM system is not required in the "Run" mode. The APRM's cover only the power range, the IRM's provide adequate coverage in the start-up and intermediate range, and therefore, the APRM's are '

not required for the " Refuel" or "Startup" modes.

The high reactor pressure, high drywell pressure, and reactor low water level scrans are required for all modes of plant operation unless the reactor is suberitical and depressurized. They are, therefore,' required to be operational for all modes of reactor operation except in the " Refuel" mode . ,

with the reactor suberitical and reactor temperature less than 212*F as allowed by Note 3.

The scram discharge volume high level trip function is required for all modes with the exception that it may be bypassed in the " Refuel Mode" under the provisions Table 3.1.1, allowable by-pass condition (a). In order to reset the safety system after a scram condition, it is necessary to drain the scram discharge volume to clear this scram input condition. This condition usually follows any scram, no matter what the initial cause might have been. Since all of the control rods are completely inserted following a scram it is permissible to bypass this condition because a control rod block prevents withdrawal as long as the switch is in the bypass condition for this function.

To permit plant operation to generate adequate steam and pressure to establish turbine seals and condenser vacuum at relatively low reactor power, the main condenser vacuum trip is bypassed until i 600 psig. This bypass also applies to the main steam isolation valves for the same reason. Ref.

Section 7.7.1.2 FSAR.

An automatic bypass of the turbine control valve fast closure scram and turbine stop closure scram is effective below 30% thermal power as indicated by turbine first stage pressure. This insures that reactor thermal power is less than 45% of its rated value. Closure of these valves from such a low initial power level does not constitute a threat to the integrity of any barrier to the release of radioactive material. I 3.1 BASES 38 !

'REV

_ _ . _ . _ . .__.__.m. _ _ . . _ _ . . _ _. 2 _..- _. .___-___ mmm______ ._____ __._.___ _ _ . ___.___ _ _ . _ . _ _ _ _ - . _ _ . _ _ _ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ . - .__.________.__._._.__.___2*u_____---w ees.*4 ere 9-T-=_ e*- q*-e *-- --> - ^-

TABLE 3.2.1 - Continued Min. No. of Operable Total No. .of Instru- or Operating Instru-ment Channels Per ment Channels Per Trip Required Function Trio Settines Trio System System (1.2) Conditions

b. High Drywell Pressure s2 psig 2 2 D (5)

3. Reactor Cleanup System (Groun 3)

a. Low Reactor Water 2:10'6" above 2 2 E Level the top of the active fuel

b. High Drywell Pressure s2 psig 2 2 E

4. HPCI Steam Lines (Grouc 4)

a. HPCI High Steam Flow s150,000 lb/hr 2(4) 2 F with 560 second time delay

b. HPCI High Steam Flow 5300,000 lb/hr 2(4) 2 F

c. HPCI Steam Line s200*F 16(4) 16 F Area High Temp.

5. RCIC Steam Lines (Groun Si

a. RCIC High Steam Flow 545,000 lb/hr 2(4) 2 G with 5 1 2 see time delay

b. RCIC Steam Line Area s200*F 16(4) 16 C

6. Shutdown Cooling Sunolv Isolation

a. Reactor Pressure 575 psig 2(4) 2 C Interlocic at the reactor steam done 3.2/4.2 50 REV

- Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems Minimum No. of Oper-Minimum No. of able or Operating Operable or < Total No. of Instru- Instrument Channels Operating Trip ment Channels Per Per Trip System Required Function Trio Settine Systems (3.6) Trio System (3.6) Conditions

B. HPCI System

1. High Drywell 52 psig 1 4 4 A.

Pressure (1)

2. Low-Low Reactor 26'6"s6'10" 1 4 4 A.

Water Level C. Automatic Deores-surization

1. Low-law Reactor 26'6"s6'10" 2 2 2 B.

Water Level and

2. Auto Blowdown s120 seconds 2. I 1 B.

Timer and

3. Low Pressure Core h60 psig 2 12(4) 12(4) B.

Cooling Pumps Dis- s150 psig Charge Pressure Interlock p

I 3.2/4.2 53 REV

__-_--.-__---_w ----n _-----_e -_-.-- _ _ _ _ -. - - - ___ - _ ~ -- * ,. < ---- - - - _ _ - r - _ . . - - ~ -

_ _ < __-_ew-_

- - _ - - . - - - - - n , --- = + - , e ---e-, v-

Bases Continued:

3.2 This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. Reference Section 6.2.7 and 14.6.3 FSAR. The instrumentation also covers the full range or spectrum of breaks and meets the above criteria. Reference section 6.2.7 FSAR.

The high drywell pressure instrumentation is a back-up to the water level instrumentation and in addition to initiating ECCS it causes isolation of Group 2 and Group 3 isolation valves.- For the-complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about.the same time as the low low water level instrumentation; thus the results given above are applicable here also. Group 2 and Group 3 isolation valves include the drywell vent, purge, sump isolation, and RUCU system valves.

Two pressure switches are provided on the discharge of each of the two core spray pumps and each of i

the four RHR pumps. Two trip systems are provided in the control logic such that either trip system can permit automatic depressurization. Each trip system consists of two. trip logic channels such that both trip logic channels are required to permit a system trip.

Division I core spray and RHR pump discharge pressure permissives will interlock one trip system and Division II permissives will interlock the.other trip system. One pressure switch on each pump will interlock one of the trip channels and the other pressure switch will interlock the other trip channel within their respective trip system.

The pump pressure permissive control logic is designed such that no single failure (short or open circuit) will prevent auto-blowdown or allow auto-blowdown when not required. The trip setting for the-low pressure ECCS pump permissive for ADS.is set such that it is less than the pump discharge pressure when a pump is operating in a full flow condition and also high enough to avoid any condition that results in a discharge pressure permissive when the pumps are not operating.

Venturis are provided in the main steamlines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steamline break accident. In addition to monitoring steam flow, 3.2 BASES 65 REV

- -__-____ - ___ __ __ __- ___ _____ __ _ _ _ __.__- - .--_-_-___ ______-_. ___ _____-..___ - . - ~ _ -

3.0 LIMITING CONDITIONS FtR OPERATION 4.0 SURVEII.IANCE REQUIREMENTS C. Containment Cooling System C. Containment Cooling System l-

1. Except as specified in 3.5.C.2 below, both 1.' Demonstrate the RHR Service Water pumps

~

Containment Cooling Subsystems shall be develop 3,500 gpa flow rate against a 500 operable whenever irradiated fuel is in the ft head when tested pursuant to reactor vessel and reactor water Specification 4.15.B.

temperature is greater than 212'F. A containment cooling subsystem consists of 2. Test the valves in accordance with the following equipment powered from one Specification 4.15.B.

division:

1 RHR Service Water Pump 1 RHR Heat Exchanger 1 RHR Pump

Valves and piping necessary for:

Torus Cooling

2. One Containment Cooling Subsystem may be inoperable for 7 days.

3. If the requirements of 3.5.C.1 or 2 cannot be met, an orderly shutdown of the reactor will be initiated and the reactor water-temperature shall be reduced to less than 212*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

For allowed out of service times for the RHR pumps see Section 3.5.A.

3.5/4.5 104.

REV

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

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

}

3.0 LIMITING CONDITIONS FT)R OPERATION 4.0 SURVEILIANCE REQUIREMENTS

. . - 3.5 CORE AND CONTAINMENT COOLINC sisIEnS 4.5 CORE AND CONTAINMENT COOLING sistm5

'Apolicability: Aeolicability:

-Applies to the operational status of the emergency Applies to the periodic testing of the emergency _

cooling systems. cooling systems.

Obiective: Obiectiver 4

To insure adequate cooling capability for heat To verify the operability of the emergency cooling removal in the event of a loss of coolant accident systems, or isolation from the normal reactor heat sink.

Snecification: Specification:

A. ECCS Systems A. ECCS Systems

1. Except as specified in section 3.5.A.3, -

1. Demonstrate the Core Spray-Pumps develop a both Core Spray subsystems and the low- 2,800 gpa flow rate against a system head Pressure Coolant-Injection (LPCI) Subsystem corresponding to a reactor pressure of 130 (LPCI Mode of RHR System) shall be operable psi greater than containment pressure, when whenever irradiated fuel is in the reactor tested pursuant to Specification 4.15.B.

vessel and the reactor water temperature is greater than 212*F. 2. Demonstrate the LPCI Pumps develop a 3,870 gpm flow rate against a system head

2. Except as specified in section 3.5.A.3, the corresponding to two pumps delivering 7,740 High Pressure Coolant Injection (HPCI) gpa at a reactor pressure of 20 psi greater System and the Automatic Depressurization than containment pressure, when tested System (ADS) shall be operable whenever the pursuant to Specification 4.15.B.

reactor pressure is greater than 150 psig and irradiated fuel is in the reactor 3. Demonstrate the HPCI Pump develops a 2700 vessel except during reactor vessel gpa flow rate against a reactor pressure hydrostatic or leakage tests. range of--1120 psig to 150 psig, when tested pursuant to Specification 4.15.B.

3.5/4.5 101 Rev:

Bases 3.5/4.5 Continued:

automatically controls three selected safety-relief valves although the safety analysis only takes credit for two valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability. l B. RHR Intertie Line An intertie line is provided to connect the RHR suction line with the two RHR loop return lines. This four-inch line is equipped with three isolation valves. The purpose of this line is to reduce the potential for water hammer in the recirculation and RHR system. The isolation valves are opened during a cooldown to establish recirculation flow through the RHR suction line and return lines, thereby ensuring a uniform cooldown of this piping. The RHR loop return line isolation valves receive a closure signal on LPCI initiation. In the event of an inoperable return line isolation valve, there is a potential for some of the LPCI flow to be diverted to the broken loop during a loss of coolant accident. Surveillance requirements have been established to periodically cycle the RHR intertie line isolation valves. In the event of an inoperable RHR loop return line isolation valve, either the inoperable valve is closed or the other two isolation valves are closed to prevent diversion of LPCI flow. The RHR intertie line flow is not permitted in the Run Mode to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a reduction in core flow during a loss of coolant accident.

C. Containment Cooling Systems Two containment cooling subsystems of the RHR system are provided to remove heat energy from the containment and control torus and drywell pressure in the event of a loss of coolant accident. A containment cooling subsystem consists of 1 RHR Service Water Pump, an RHR Heat Exchanger,'1 RHR Pump, and valves and piping necessary for Torus Cooling. Torus Spray and Drywell Spray are not considered part of a containment cooling subsystem. For the flow specified, containment cooling is adequate to provide the required heat removal capability. Reference Section 6.2.3.2.3 USAR.

t Either subsystem is capable of performing the containment cooling function. Imss of one RHR service water pump in each subsystes does not seriously jeopardize the containment cooling capability as one pump in one subsystem can satisfy the cooling requirements. Loss of one containment cooling subsystem leaves one remaining system to perform the containment cooling function. ,

3.5/4.5 112 Rev i

i

, _ _ _ - _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - ~ ' - "

i Bases Continued:

Vent system downcomer submergence is three feet below the minimum specified suppression pool water level. This length has been shown to result in reduced postulated accident loading of the torus while at theand conditions same time assuring possess adequatethe downcomerseffectiveness.(

condensation remain submerged under all seismic and accident The maximum temperature at the end of blowdown tested during the Humboldt Baym and Bodega Bay m tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170*F.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160'F during any period of relief valve operation with sonic conditions at the discharge exit. Specifications have been placed on the e of reactor o envelogd to avo the regime ogerating conditions potentially so that the chamber high suppression reactor can be depressurized in a timely manner loadings In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently o ens or sticks open. This action would include: (1) use of all available means to close the valve, initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4 2)f i other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

For an initial maximum suppression chamber water temperature of 90*F and assuming the minimum complement of containment cooling pumps (1 RHR pump in suppression pool cooling mode and 1 RHR service water pump), with 90*F ultimate heat sink temperature; adequate net positive suction head (NPSH) is maintained for the core spray, RHR and HPCI pumps with the benefit of containment pressure.

(1) Robbins, C.H. " Tests of Full Scale 1/48 Seament of the Humboldt Bay Pressure Suppression Containment," GEAP-3596 Novem5er 17, 1960.

(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(3) General Electric NEDE-21885-P, " Mark I Containment Program Downconer Reduced Submergence Functional Assessment Report", June,1978.

t 3.7 BASES 176 REV

O M IAER 3.4 and 4.4 Standby Liquid Control System 93 A '. System Operation 93 ,

B. Boron Solution Requirements 95 C. 96 4

3.4 and 4.4 Bases 99 f

3.5 and 4.5 Core and Containment Cooling System 101 l A. ECCS Systems 101 l l B. RHR Intertie Return Line Isolation Valves 103 ;

C. Containment Cooling System 104- l D. RCIC 105 :

E. Cold Shutdown and Refueling Requiremetns 106 F. Recirculation Systa.m 107 3.'5/4.5 Bases 110 l 3.6 and 4.6 Primary System Boundary 121 !

A. Reactor Coolant Heatup and Cooldown 121 B. Reactor Vessel Temperature and Pressure 122 i C. Coolant Chemistry 123 ,

D. Coolant Leakage 126 E. Safety / Relief Valves 127 F. Deleted 128 l G. Jet Pumps 128 H. Snubbers 129 3.6 and 4.6 Bases- 145 l 3.7 and 4.7 Containment Systems 156 A. Primary Containment 156 1 B. Standby Cas Treatment System 166 C. Secondary Containment 169 ,

D. Primary Containment Isolation Valves 170 !

E. Combustible cas Control System 172 1 3.7 Bases 175 l 4.7 Bases 183 '

l 11 1 l

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l

r ZAER 3.8 and 4.8 Radioactive Effluents 192 A. Liquid Effluents 192 7 B. Caseous Effluents 197 C. Solid Radioactive Wasta 198e D. Dose from All Uranium Fuel Cycle Sources 198f 3.8 and 4.8 Bases 198u' 3.9 and 4.9 Auxilliary Electrical Systems 199 A. Operational Requirements for Startup 199 B. Operational Requirements for Continued Operation 200

1. Transmission Lines 200

2. Reserve Transformers 201

3. Standby Diesel Generators 201

4. Station Battery System 203

5. 24V Battery Systems 203 3.9 Bases 204 ,

4.9 Bases 205 3.10 and 4.10 Refueling 206 i

A. Refueling Interlocks 206 !

B. Core Monitoring 207 l C. Fuel Storage Pool Water Lavel 207 j D. Movement of Fuel 207 i E. Extended Core and Control Rod Drive Maintenance 208 l

3.10 and 4.10 Bases 209 I 3.11 and 4.11 Reactor Fuel Assemblies 211 A. Average Planar Linear Heat Generation Rate 211 B. Linear Heat Generation Rate 212 C. Minimum Critical Power Ratio 213 1 3.11 Bases 216 I 4.11 Bases 218 l 1

3.12 and 4.12 Sealed Source Contamination 219 A. Contamination 219 B. Records 221 3.12 and 4.12 Bases 222 ,

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l EXHIBIT D LICENSE AMENDMENT REQUEST DATED JULY 26,1996

)

i

SUMMARY

OF PLANT MODIFICATIONS FOR POWER RERATE IMPLEMENTATION l

l 1

l NORTHERN STATES POWER COMPANY l MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA l

I LICENSE NO. DPR-22 DOCKET NO. 50-263

l l

I Exhibit D Summary of Plant Modifications for Power Rerate implementation Hardware Charmes High Pressure Turbine rotor replacement to support power rerate steam flow path.

Modification to piping or equipment supports for some plant systems due to load changes .

Involves approximately twelve (12) pipe supports.

Modification to Control Room Emergency Filtration Train (EFT) system to reduce control room l ventilation filter bypass leakage to establish consistency with control room dose calculation inputs.

Modification to the main steam drain and condenser attached piping to satisfy seismic criteria for dedication of piping in accordance with BWROG methodology.

Modification of the four Moisture Separator Drain valves to enhance the drain capacity.

Modification of the main generator isophase bus cooling system to enhance the cooling capacity.

Modification of the Main Generator Stator Water Cooling system to enhance cooling capacity.

Modification of the Feedwater Regulating Valve stroke length to maintain feedwater flow runout capacity.

Non-Hardware Channes !

Perform adjustments to installed plant instrumentation as necessary to support changes to system process conditions. Examples of affected systems: !

Condensate Demineralizer Flow Controllers, Main Steam Line High Flow, Main Steam Line High Radiation, PeveAss for Turbine Stop Valve Closure and Turbine Control Valve Fast Closure, Main Generator Protective Devices, !

APRM flow biased neutron flux scram, and Condenserlow vacuum scram.

Process computer software and computer data point changes to reflect increased rated thermal power.

Setpoint change for two of the four High Pressure to Low Pressure Turbine cross-around relief '

valves.

i i

Q 9%

i Recertify the ASME Section Vill Feedwater Heaters for the slightly higher temperatures and pressures necessary to support the power rerate.

Upgraded the HPCI System piping analysis to 170 F.

Upgraded the RCIC System piping analysis to 140*F.

Revised the EFT system filtration efficiency to 98%.

Revised the SBGT system filtration efficiency to 85%.

1 Revise operating limits to ensure the 1R transformer design ratings are not exceeded. !

Environmentally qualified equipment maintenance interval changes to address the power l rerate effects on service life for a limited scope of equipment. !

l Erosion / Corrosion program changes to address power rerate effects on inspection frequency. j i

l l

l D-2 l

EXHIBIT F LICENSE AMENDMENT REQUEST DATED JULY 26,1996 MNGP POWER RERATE ENVIRONMENTAL EVALUATION l

l NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA l

l LICENSE NO. DPR-22 DOCKET NO. 50-263

. - . __ . -- -. .- ~ =- ~. . - - ..

Table of Contents ;

Paae Ro.

Executive Summary il

. 1.0 Introduction F-1 2.0 Overview of Operational and Equipment Changes F-2 3.0 Proposed Action and Need . F-2 3.1 Proposed Action F-2 3.2 Need for Proposed Action F-2 4

4.0 Socioeconomic Effects F-3 4.1 Economic Structure F-4 )

4.2 Economic Benefits for Power Rerate Equipment and Service Suppliers F-4 l 4.3 . Tax Benefits of Power Rerate F-5 4.4 Economic Competitiveness of MNGP Under Power Rerate Conditions F-6 1 5.0 Cost- Benefit Analysis F-8 6.0 Non-Radiological Environmental impact F-7 6.1 Terrestrial Effects F-7 6.2 Hydrology F-11 7.0 Radiological EnvironmentalImpact F-18 1

7.1 Radioactive Waste Streams F-18 7.2 Radiation Levels and Offsite Dose F-25 7.3 Radiological Consequences of Accidents F-27 7.4 Other Potential Environmental Accidents F-36 8.0. Environmental Effects of Uranium Fuel Cycle Activities and Fuel and Radioactive F-36 Waste Transportation 8.1 Compliance With 10 CFR 51.51, Uranium Fuel Cycle Environmental Data F-36 (Table S-3) 8.2 Compliance With 10 CFR 51.52, Environmental Effects of Transportation F-37 and Waste (Table S-4) 9.0 ' Decommissioning Effects F-38 10.0 Conclusions F-38 11.0 References F-40 i

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. t EXECUTIVE

SUMMARY

l-l This exhibit presents an evaluation of the environmentalimpacts of the proposed Monticello thermal power rerate from 1670 MWt to 1775 MWt. The intent of this exhibit is to provide !

sufficient information for the Staff to evaluate the environmental impacts of power rerate in j accordance with the requirements of 10 CFR Part 51, l

The environmental impacts of power rerate are identified and compared against the l environmental impacts associated with the present power level which have been previously evaluated by the Staff in the Final Environmental Statement associated with the issuance of I the Monticello full term operating license and in other related docketed correspondences.

The environmental impacts identified by the Staff in the Final Environmental Statement are based on conservative assumptions for source terms and other environmental parameters.

i Since initial operation, a variety of systematic environmentalimprovements have been '

I implemented at Monticello that have further increased the margin of conservatism associated with these assumptions. By adjusting actual plant operating parameters for power rarate effects, it can readily be demonstrated that the previous assumptions and conclusions I concoming the environmental impact of Monticello operation at present power levels continue l to bound plant operation at power rerate conditions with significant margin.

In a few cases, the Final Environmental Statement and its associated documentation does not !

contain sufficient information necessary for a detailed comparison of the power rerate

environmental impacts with previously evaluated impacts. In these instances, comparisons 1 and conclusions are made using other appropriate environmental criteria established by the l

NRC. Where other environmental authorities govem Monticello operation such as in the '

matter of state water appropriation limits, comparisons and conclusions are made using the appropriate environmental permits and regulations.

The Monticello power rarate is being implemented without consequential changes to the plant systems that directly or indirectly interface with the environment. This evaluation demonstrates that the environmental impacts of plant operation at power rerate conditions are insignificant.

The environmental impacts of power rerate are either well bounded or encompassed by previously evaluated environmentalimpacts and criteria established by the Staff in the F*mal Environmental Statement or well bounded by other appropriate regulatory criteria.

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l il l

l l

1.0 INTRODUCTION

The Northem States Power Company (NSP) is committed to operating the Monticello Nuclear Generating Plant (MNGP) in an environmentally sound manner. All plant activities, including design, construction, maintenance, and operation, are conducted in a manner that involves strict compliance with environmental regulations and deliberate consideration !

of environmental practices and consequences. Numerous controls and modifications have l been implemented to prevent and reduce impacts to the environment, and extensive ;

- environmental monitoring programs have been instituted at MNGP. In keeping with this important obligation and in accordance with regulatory requirements, NSP has conducted a .

comprehensive environmental evaluation of the proposed MNGP power rarate from 1670 MWt to 1775 MWt.

This environmental evaluation is provided pursuant to 10 CFR 51.41 and is intended to fully support the Commission in complying with the requirements of Section 102(2) of the National l Environmental Policy Act (NEPA), as amended, for the proposed change to the MNGP ;

operating powerlevel. The scope of the evaluation is limited to that information necessary .

and sufficient to determine the environmentalimpact of those particular changes l associated with the proposed power rerate at MNGP from 1670 MWt to 1775 MWt. This evaluation is not specifically intended to reestablish the current environmental licensing basis or to justify the environmental impacts of operating at the present power level.-

The environmental impact of operation at the present power level has been reviewed and determined to be acceptable by the Staff. In 1971, NSP provided an Environmental Report (Ref. 2) to the Atomic Energy Commission (AEC) as part of NSP's application for a full term operating license. The Environmental Report addressed the environmental impacts of -

construction and operation of MNGP. The report was utilized by the AEC in preparing a Final Environmental Statement or FES (Ref. 3) in fulfillment of the requirements of the National Environmental Policy Act of 1969. The NRC subsequently issued a full term operating license to MNGP (Ref. 22).- This license authorized a maximum power level of 1670 MWt. By the Notice of issuance included as Enclosure 2 to Ref. 22, the Commission stated that "... issuance of this license will not result in any environmental impacts other than those evaluated in the Final Environmental Statement since the activity authorized by the license is encompassed by the overall action evaluated in the Final Environmental Statement."

l This evaluation demonstrates that the environmentalimpacts of power rerate are either well bounded or encompassed by previously evaluated criteria established by the Staff in the FES or well bounded cy other appropriate regulatory criteria.

Since power rerate involves no significant environmentalimpacts as delineated by 10 CFR l Sections 51.22(a), (c)(9) and as further described and evaluated herein, NSP believes that l sufficient evidence exists tu justify application for a categorical exclusion as provided by 10 i' CFR 51.21. Accordingly, NSP is hereby requesting that the Staff consider the proposed change in power level eligible for a categorical exclusion.

F- 1

i 2.0 OVERVIEW OF OPERATIONAL AND EQUIPMENT CHANGES Monticello is a Boiling Water Reactor (BWR) that operates in a direct thermodynamic cycle between the reactor and the turbine. At power rerate conditions, thermodynamic processes are changed to extract additional work from the turbine. Simply put, power

rarate involves an increase in the heat output of the reactor to support increased turbine inlet steam flow requirements and an increase in the heat dissipated by the condenser to 4 support increased turbine exhaust steam flow requirements in order to support a power rerate to 1775 MWt, the reactor core operating range will be expanded by increasing -

reactor power within existing rod and core flow control lines. No changes in operating pressure, core flow, or turbine throttle pressure are necessary to support power rerate. In l the turbine portion of the heat cycle, increases in steam flow will result in a slight increase In the heat rejected to the Mississippi River. The environmentalimpacts of these operational changes are discussed herein. l Due to design and safety margins inherent in plant equipment, the proposed power rerate i can be accomplished with relatively few modifications. The most significant change - 1 involved replacement of portions of the High Pressure (HP) turbine. Other minor I modifications to support power rerate, such as improvements to the phase bus duct cooling system, are routine in nature and are being conducted within the plant boundary. These ;

modifications are being accomplished by standard maintenance and modification '

processes that are similar to those performed during normal outages. The majority of plant systems will not require any significant modifications. ;

I 3.0 PROPOSED ACTION AND NEED 3.1 Proposed Action With the operational goal of increasing electrical generating capacity, NSP, in conjunction with the plant designer, General Electric, has comprehensively evaluated i the effects of a power rerate at Monticello. This evaluation concluded that sufficient safety and design margins exist such that a prudent increase in the rated core thermal power from 1670 to 1775 MWt can be accomplished without any adverse impact on the health and safety of the public and without any significant impact on the environment.

Accordingly, NSP is proposing an amendment to the MNGP Operating License to allow for an increase in the licensed core thermal power level to 1775 MWt.

NSP does not intend to raise power in increments at MNGP. The maximum power level proposed by this action and evaluated for environmental impact herein is 1775 MWt.

3.2 Need for Proposed Action NSP has filed a fifteen year resource plan for the period 1996-2010 with the State of l Minnesota (Ref. 9). This resource plan includes a forecasted increase in expected customer peak demand, without additional conservation and load management, of approximately 1.8% - 2.8% per year through the 1996-2010 planning period. To meet this projected demand, generating capacity must increase by 3,800 to 6,000 MW by 2010.

l t

F- 2 1

NSP postulates three scenarios (median, semi-low, and semi-high) to forecast future electric energy and demand. The base median forecast includes the impacts of past and futum Demand Side Management (DSM) programs. This forecast includes s'

" business as usual" assumption in which there is no basic change in the relationship between the regional and national economies. The semi-low and semi-high forecasts account for growth uncertainties using different assumptions. The forecast range . .

bounded by the semi-low and seml-high forecasts is used as the basis for determining NSP's long-term resource needs.

NSP has determined the need for additional generation resources through a comparison of the projected resource needs (Obligations) to the resources available to NSP (Committed Resources). NSP's resource obligations include forecasted summer.

peak demand, Mid-Continent Area Power Pool (MAPP) minimum reserve requirements, and other contracted obligations. Committed resources include existing capacity, !

, committed capacity additions, and committed capacity purchases.- The results of this comparison are shown in Table 1 below. ;

Table 1. NSP Total Resource Needs (MWe) 1998 2002 2006 2010 Forecasted 8,969 - 9,889- 10,871 - 11,811 - I Obligations 9,516 10,912 12,440 14,022 i (semi-low to semi high)

Committed 8,746 8,646 8,071 8,021 Resources Resource 223-771 1,243 - 2,800- 3,790 -

Need 2,267 4,369 6,001 Before DSM The proposed increases in electrical generating capacity due to the Monticello power rerate are already included in the Committed Resource values displayed above. As shown in Table 1, NSP expects to require additional capacity by the late 1990s, and a significant shortfall in electrical generating capacity is predicted for the back end of the planning period.

The shortfall will require additional generating capacity. The capability of DSM .

programs to fulfill capacity needs is limited (Section E, Executive Summary and Section ;

IV.D of Ref. 9). In addition, the MAPP system is projected to experience a 100 MWe '

committed capacity deficit in the year 2001 (Section IV.H of Ref. 9). The MAPP deficit ;

is projected to grow by almost 500 MWe per year. This capacity deficit is not confined !

to the MAPP region, and surrounding regions are reporting similar deficits.

i 4.0 SOClOECONOMIC EFFECTS l

This section addresses the effect of power rerate on the social and economic conditions of ,

communities affected by MNGP operation. NSP, as a matter of policy, does not presently ;

recommend selection of generation resources on the basis of quantified environmental F- 3 1

l extemalities and socioeconomic effects (Section IV.C of Ref. 9). Therefore, the following discussions do not include these types of comparisons. .

4.1 Economic Structure l l

Power rerste does not significantly affect the size of the MNGP work force and does not I have a material effect on the labor force required for future plant outages.

In 1995, NSP employed approximately 350 full-time workers at MNGP. These workers have a disproportionate influence on the economics of the region because of higher -

incomes. Estimated per capita and median household income in 1990 for Monticello, St.

' Cloud, Short>urne County, and Wright County are presented in Table 2. The 1990 estimated average annual wage of MNGP employees was $45,000. The 1995 estimate is i l

$52,000.

Table 2.1990 Census Per Capita Personal and Median Household income Jurisdiction Per Capita Personalincome Median Household income Communities Monticello $11,907 $29,583 .

St. Cloud $11,736 $24,004 Counties Sherbume $13,147 $35,585 Wright $12,687 $33,456 Two-County Average $12,917 $34,521 ,

Minnesota $14,389 $30,909 Sources: Ref. 7, Minnesota State Planning Department- Office of Demography 4.2 Economic Benefits for Power Rarate Equipment and Service Suppliers Although the amount of plant modification and new equipment required to implement power i rerate is relatively small, there is a significant positive economic benefit to local and national )

businesses derived from power rerate at MNGP. The General Electric Company was j awarded the contract for the major engineering services associated with power rerste.

Other local engineering firms, equipment suppliers, and service industries are receiving ,

payments for power rerate related activities. This direct revenue from NSP will be reduced i and will eventually cease within a few years of power rerate implementation. Successful implementation of power rerate at MNGP will likely entail a significant amount of follow-on revenue for equipment and service suppliers as other nuclear plant owners recognize the l viability of power rarate and commence similar projects. l 1

F- 4 l

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1 4.3 Tax Benefits of Pover Rorate According to tax revenue classifications used by the Staff (Section 3.7.3 of Ref. 8), the '

. MNGP contribution to local taxing judsdictions is classified as significant. This contribubon to the local tax base will increase under power rerate conditions. It is expected that the HP -

j turbine modification will contribute approximately $2.67 million in property taxes above that l associated with the present HP_ turbine over the life of the power rarate project. This i modification will result in a payment of over $155,000 in sales taxes. In addition, the HP turbine modification resuKs in a project life increase of $7 million in state and federalincome i taxes. Moreover, because of increased revenues, power rarate is expected to result in an 4

increase of $28 million in state and federalincome taxes over the project life.

The ability of the local community to provide public services at a reasonable tax rata is j largely due to NSP payments to local taxing jurisdictions. Public services, including law

enforcement, fire protection, public education, and health serv'ces, receive a substantial amount of economic support through tax revenues generated by MNGP. NSP paid a total

{ of $14,110,644 in local taxes to the City of Monticello, Wright County, School Distdct 882, and Monticello/ Big Lalw Community Hospital in 1995. The majority of local taxes (over 50 j percent) are paid to School District 882. The school district is actively considering a bond 4

issue to finance the construction of new facilities. This action, if approved, willincrease the i- district's dependence on revenues from plant operation. A significant reduction in the NSP j

j. contribution from MNGP operations will result in economic penalties and/or loss of services !

i to businesses, farmers, and homeowners as the NSP tax contribution differentialis l apportioned to the remaining tax revenue sources.-

I Market values and tax disbursements for 1980,1985,1990, and 1995 can be found in

Table 3 and Table 4 below.

t Table 3. Assessed Market Values of MNGP L

! 1980 1985 1990 1995 Assessed Value $81,424,761 $167,935,000 $251,098,300 ' $257,557,200

{

) Table 4. Northem States PowerTax Disbursements I

{ Taxing i Taxes Paid Jurisdiction '

1980 1985 1990_ 1995 City $650,174 $1,151,243 $2,057,032 $2,277,902 l

County $728,333 $1,202,785 $2,792,685 $3,843,695 l

School $1,324,780 $2,422,405 $5,191,879 $7,666,590 l District l Hospital $95,020 $182,867 $357,027 $322,457 L

) . Total $2,798,307 $4,959,308 $10,398,623 $14,110,644 i

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< 4.4 Economic Competitiveness of MNGP Under Power Rerate Conditions ;

i l The socioeconomic effects of power rarate are, in part, dependent on whether power rarate I

is necessary for MNGP to remain economically competitive in a deregulated electrical -

generation market. Although implementation of power rerate is not the sole factor affecting the future economic competitiveness of MNGP, it is a real and material factor. MNGP is not the least cost provider among NSP's generation assets, and reductions in operating costs .

are necessary to make MNGP more competitive. Notwithstanding the uncertainty and i

economic viability concems associated with deregulation, power rerste will make MNGP l more competitive by reducing MNGP's operating costs. Power rerste represents an investment that will help to manage risk by improving the cash flow and enhancing the value of Monticello as a generating asset. Power rerate will certainly be a contributing factorin l any future dec'sion on the continued operation of MNGP under deregulated market conditions.

The impact of not implementing power rerate on the probability of early termination of

' MNGP operation is difficult to quantify. Since this probability is greater than zero, it is prudent to outline the socioeconomic impacts of early termination, albeit briefly, herein.

NSP is a major employer in the community (second behind the school district) and the i

largest single contributor, by far, to the local tax base. MNGP personnel also contribute to the tax base by payment of sales taxes and property taxes. Many MNGP personnel are i involved in volunteer work within the community. An early termination of plant operation would have a significant negative impact on the local economy and the community. The ability of the local economy to provide substitute tax revenues and similar employment opportunities for MNGP employees is severely limited. Serious reductions in public '

services, employment, income, business revenues, tax revenues, and property values would result. These reductions may be mitigated somewhat by decommissioning activities i in the short term.

l 5.0 COST - BENEFIT ANALYSIS ,

i The direct benefit of power rarate to NSP's residential and commercial customers is that power rerate will supply an additional 26 MW of reliable electrical energy.

i in addition to the indirect benefits from increased tax revenues, direct economic benefits are derived from the savings associated with avoided costs. The incremental expenditures for power rerate on a $ / MWh basis are substantially lower than those associated with l altemative generating units. NSP conducted an economic analysis of current avoided generating unit costs in keeping with its Resource Recovery Plan filed with the Minnesota

Public Utilities Commission (Ref. 9). The avoided unit costs are on file with the l Commission. These costs include energy (fuel) and operational costs. When avolded i costs for deferral of peaking, intermediate, and baseline costs are accounted for, the future

deferred energy and capacity savings from implementation of power rarate realized over l the life of the project are as follows.

Peaking (1998-1999) $9.6 million l Intermediate (2000-2004) $34.84 million

[ Baseload (2005-2010) $50.56 million l Total $95 million I

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i In NSP's analysis the peaking reference plant is a natural gas combustion turbine plant, I and the reference plant for the intermediate and baseload cases is a natural ges combined !

cycle plant. For the peaking and baseload cases, the reference plant is an existing NSP unit. The intermediate case involves competitive market analycis with existing non-NSP l capacity (i.e. purchased capacity). Spot market comparisons were not performed, due to ;

long term unreliability, lack of security, and increased risk associated with power purchases '

over the power rarate projectlife. !

l Although a quantitative study of environmental costs of altamatives was not pe Tormed, l (see Section 4.0 above), it is apparent that significant environmental benefits can be ;

derived from power rerate when compared to the reference plant options of adding i capacity described above. As demonstrated herein, there are no significant environmental costs associated with power rerate. Unlike fossil fuel plants, MNGP does not routinely emit SO , NOx, CO2, or other atmospheric pollutants during normal operation. Routine j l operation of MNGP at power rerate conditions will not contribute to greenhouse gases or l I

acid rain. By Section 9.3 of Ref. 28, the Staff evaluated the potential for renewable energy resources to replace nuclear capacity. Although this evaluation pertains to license renewal, the conclusions of Section 9.3.11 are relevant to power rarate at MNGP. That is, I because of technical, geographical, and availability factors, renewable energy sources do i not have the economic potential to replace the generating capacity associated with power )

rerate in the near term. Moreover, the economic competitiveness of renewable energy j l resources is likely to decrease as a competitive market emerges for utility services.

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l Given the arguments above, it is reasonable to conclude that the MNGP power rarate l project is economically superior to all generation altamatives. The power rerate option is also environmentally superior to fossil fuel generation. In summary, power rerate involves more effective utilization of an existing asset with no significant additional environmental impacts which is preferable to the most reasonable replacement option which involves ;

i inefficient use of resources with adverse environmentalimpact.

( 6.0 NON. RADIOLOGICAL EPT!!RONMENTAL IMPACT f

6.1 Terrestrial Effects 6.1.1 Land Use The MNGP power rarate does not result in any activity which will change or otherwise modify the present requirements forland use at the plant site. There are no plans to build facilities or modify access roads, parking areas, laydown areas or onsite transmission / distribution equipment to support power rarate activities at MNGP. Except for transportation of equipment and routine disposal of waste, power rerate maintenance activities are confined to the inner-plant security fenced area. Power rerate does not affect the storage requirements for above ground or below ground tanks. Other lands located outside the inner security fence will not be modified or changed to support power rerate activities. Power rerate does not involve changes to any aesthetic resources and does not involve any impacts to l

lands with historical or archaeological significance.

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NSP does not anticipate the need to construct additional or new low-level radioactive waste storage buildings to support present or power rerate activities.

The replaced turbine components are being shipped to a private salvage firm. The components will be decontaminated, as necessary, and recycled to the extent possible.

6.1.2 Transmission Facilities A. Transmission Design and Equipment No changes in operating transmission voltages, onsite transmission equipment or power line right of way are required to implement or support power forate.

The stability of the offsite power system at MNGP is especially robust due to MNGP's relative grid location and its proximity to the large base-loaded Sheroo Generating Station. There are no new requirements or modifications necessary for the local offsite power system at MNGP to maintain grid stability. In part because of MNGP generator power factor changes at power rerate conditions, j certain modifications to offsite substations are being made to enhance stability !

at remote grid sites. These upgrades are in keeping with NSP's continuing program of systematic improvements in grid stability and are in accordance with NSP's commitments to the Mid-Atlantic Power Pool.

B. ShockHazards Power rarate does not increase the probability of shock from primary or secondary currents. Transmission lines are designed in acedise with the applicable shock prevention provisions of the National Electric Safety Code (NESC). !

C. Electromagnetic Fields (EMF) l 1

According to the Staff, the chronic effects of EMF on humans are unquantified at this time, and no significant impacts to terrestrial biota have been identified (Sections 4.5.4.2.3 and 4.5.6.3.4 of Ref. 8). According to the National Institute of Environmental Hsalth Sciences, there is no scientific consensus regarding I the health effects,if any of EMF (Ref.15). The chronic effects from EMF exposure have not bem conclusively established. Notwithstanding the above, the following informatkvi ls presented to show that power rarate does not involve ,

significant increases in exposure to electromagnetic magnetic fields from transmissionlines.

)

Power rarate will not result in any changes to the magnetic fields associated with the offsite transmission system. The MAPP Design Review Committee has previously approved the combined net output of the Sherco and Monticello generating stations. This net output is being maintained by re-allocating 30 MW of previously approved Sherco capacity to MNGP. That is, the additional power was scheduled for generation and transmission irrespective of Monticello's power rerate. See introduction Section of Ref.1 for supplementary information.

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The increased generator output at MNGP wiu cause a corresponding current (and thus magnetic field) rise in the onsite transmission line between the MNGP l main generator and the plant substation. This line, however, is located within the outer fenced boundary of the plant where public access is prohibited. The elevated line is primarily located over gravel-covered areas without foliage that are not typically inhabited by humans or wildlife.

l 6.1.3 MisceHaneous Wastes l

Sanitary wastes from MNGP are discharged directly to the MonticeHo Wastewater Treatment Plant in accordance with a pemiit issued by the City of MonticeHo. Acid drains are processed in a retention basin in accordance with NPDES permit l requirements. ~ Other waste sources include hazardous waste generation from

routine plant operations and air emissions from the plant heating boiler and diesel

! generators. Effluents from these pathways are controHed as required by state and federal permits. Power rerate does not have any significant impact on the quality or -

quantity of effluents from these sources, and operation under power rerate conditions wiu not reduce the margin to the limits established by the apprepf.ie permits. See Section 6.2.5 herein for additionalinformation on water quauty.

6.1.4 Cooling Tower Drift, icing, and Fog Estimates of ground fog frequency and drift and the associated enviiceTental impacts were provided to the Staff in Supplement 1 to the MNGP Environmental Report (Questions 4.b and 8 of Ref. 4). These impacts bound operation at power rerate conditions. <

Since initial operation, the coohng towers at MNGP have been upgraded with high-efficiency polyvinyl chloride drift eliminators. These polyvinyl chloride drift eliminators replaced wooden drift eliminators. Loss of water by evaporation and drift previously totaled about 3 percent of the design circulation rate when water is cooled through a 30*F range. The polyvinyl chloride drift eliminators reduced this loss to approximately 2.7 percent of total flow. Unlike the previous drift eliminators, which allowed water droplets to retum to the cooling tower air stream (and ultimately to exhaust air), the new drift eliminators channel water to hoHow polyvinyl chloride spacers that empty into the coohng towers cold water basin. These new drift eliminators have reduced icing in the immediate vicinity of the cooling towers in winter, Drift, icing, and fog from the MNGP cooling towers have been negHgible and have had no discemible impacts on vegetation, agriculture, recreational activities, highway safety, air traffic, or river traffic. The Mississippi River does not contain the l

salt content of other water sources, and sufficient rainfaH is available to prevent undesirable chemical concentrations in the soll from trace chemicals in the drift.

Noglechng increased turbine efficiency and conservatively assuming that the fogging rate is proportional to the proposed powerincrease, the amount of coohng tower fogging due to power rarate could increase by approximately 6.3% over the normal summer operating period of four months, in addition, power rerate may l

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Involve an extra week of cooling tower operation. (See Section 6.2.5 herein.)

These changes will not have a significant effect on the environment. Assuming cooling tower operation from April to October (seven months), the Staff conservatively estimated a total fogging time of 45 hr/yr. The fogging rate at power ;

rerate conditions is bounded by this estimate. The Staff concluded that no

- biological impacts are attributable to cooling tower fogging at MNGP (Sections V.C.1, Xill, and Question 42 of Ref. 3).

6.1.5 Noise

.. Power rarate does not result in any significant changes to the character, sources, or energy of noise generated at MNGP. The new equipment necessary to implement power rarate will be installed within existing plant buildings. No significant increases in ambient noise levels are expected within the plant.' This includes the upgraded HP turbine which will operate at the same speed as the original equipment. The additional week of cooling tower operation on ambient noise levels is not significant.

No new significant noise-generating equipment will be installed outside the plant. l The Staffs conclusions for noise levels (Part Xill, Question 28 of _Ref. 3) remain I bounding for power rerate conditions.

6.1.6 Terrestrial Biota Power rerate will not change the previously evaluated land use at MNGP and will not disturb the habitat of any terrestrial plant or animal species. There are no - i significant increases in previously evaluated environmental impacts from cooling tower operation at power rerate conditions. Water quality and contaminant levels are not affected by power rerate. No changes to aquatic prey will occur.

There are no known rare or endangered plant species that exist within the area of the inner security fence. Certain threatened and endangered animal species have been sighted or are likely to exist in the vicinity of MNGP outside the confines of the j plant inner security fence. These species are identified below.

The state listed species include the following.

- Loggerhead Shrike (Lanius ludovicianus) - Federal Category 2 1

- Blanding's Turtle (Emydoidea blandingli) - Federal Category 2 l l (no actual sightings) ;

- Trumpeter Swan (Cygnus buccinator) - proposed for threatened designation i

-1 The federal and state listed species include the following, i

- Bald Eagle (Haliaeetus leucocephalus) ,

- Peregrine Falcon (Falco peregrinus) i J

The effect of power rerate on these species was assessed using information from 1 i the Minnesota Department of Natural Resources and the U.S. Fish and Wildlife

! Service. The biological assessment did not identify any impacts from power rarate l j on these species or on their critical habitat.

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l The trumpeter swan is sensitive to the availability of open water during late fall and winter. These swans winter in the Mississippi River downstream of MNGP. The I availability of open water in winter could be affected by an unplanned plant outage.

The frequency of unplanned outages and the subsequent recovery from these outages are independent of power rerate. A peregrine falcon nest box was installed on the MNGP offgas stack in 1992. In 1995, a female falcon successfully fledged three young. Power rerate will not affect this nest box.

6.2 Hydrology 6.2.1 Groundwater .

Power rerate does not affect groundwater resources and does not involve significant increases in the consumptive use of these resources at MNGP. Station groundwater use is govemed by water appropriation limits of the Minnesota Departrnent of Natural Resources (MDNR). The domestic water supply is obtained i from four wells located on the plant property. No dewatering or collector-type wells (Ranney wells) are used at MNGP. The Domestic Water System, which is serviced by two 50 gpm wells, provides domestic water to lavatories, showers, and laundries '

and provides raw water to the reverse-osmosis system and seal water to certain pumps located at the plant intake structure. Groundwater appropriation permit ,

number 670083 establishes limits associated with these 50 gpm wells. Power rerate does not affect compliance with these limits. The annual appropriation limit is 15 million gallons and annual usage over the last five years is less than 13 million gallons. Any increases in makeup to plant systems under power rerate from these-sources are expected to be minor, and operation within the allowable limit will continue. Two 45-gpm we!!s provide water to office and warehouse facilities not serviced by the Domestic Water System. The wells are of standard vertical construction. Power rerate has no effect on these sources.

MNGP monitors groundwater as part of the Radiological Environmental Monitoring Program. Since 1976, four wells have been sampled quarterly which includes sampling for tritium and gamma-emitting radionuclides. No radioactive or chemical contamination has been detected in any of the wells. See Section 4.3 of Ref.19 for recent gamma isotopic results. Power rerate has no effect on the contamination levels, radioactive or non-radioactive, of these wells.

6.2.2 Surface Water Appropriation Surface water use at MNGP is in accordance with the water appropriation limits of the Minnesota Department of Natural Resources (MDNR). Under surface water appropriation permit number PA 661172-S, NSP may withdraw a maximum of 645 cubic feet per second (cfs) of water from the Mississippi River at MNGP. Special l operating restrictions apply at lower than average river flows of 860 cfs and 240 cfs.

! Power rerate does not introduce any significant changes to the screen wash, service j water, or circulating water flow requirements. Power rerate does not involve any ,

changes to the water appropriation requirements of this permit. '

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Using conservative assumptions, the present cooling tower evaporation rate at MNGP has been calculated to be approximately 4100 acro-ft/ year. Assuming that

! the evaporation rate is linearty proportional to the proposed power increase and including the effects of an additional week of cooling tower operation, power rerste could potentially cause an increase to 4400 acre-ftlyr in evaporative losses from the '

j river. This increase in evaporation is well below that value for evaporative losses t

(5000 acro-ftlyr) previously evaluated by the Staff (Part 3.- Summary and Conclusions and Section V.B of Ref. 3).

Power rerate may involve an additional week of cooling tower operation. The combined effects of evaporation and cooling tower operation at power re ate conditions have an insignificant impact on water consumption. The Final Environmental Statement conservatively assumed that cooling tower operaten would occur from April to October (approximately 210 days). Annual cooling tower operation at power rerate conditions is estimated to be less than 130 days. The Staff's conclusion concoming the insignificant consumptive use of water at MNGP

made in Section V.B of the FES (Ref. 3) will remain valid at power rerate operating l conditions.

6.2.3 Discharges Surface water and wastewater discharges are regulated by the State of Minnesota.

The National Pollutant Discharge Elimination System (NPDES) permit is penodically

! reviewed and re-issued by the Minnesota Pollution Control Agency (MPCA). The present NPDES permit for MNGP, permit number MN0000868, which expired December 31,1995, authorizes discharges from five outfalls. The outfalls and their effluent limits are listed in Table 5 herein. None of the limits listed in this table will require modification to implement power rerate. No changes to the permit requirements, other than administrative and descriptive changes, are necessary to i implement power rerate.

l An application for renewal of the existing NPDES permit was initiated in June 1995 l as required (Ref.12). NSP identified the relevant power rerste changes and j requested eariy consideration of these changes to support engineering efforts. The ,

l identified changes include the slight increase in circulating water discharge !

temperature (outfall 010). The MPCA reviewed the power rerate changes and !

has informed NSP that, as of October 1995, there were no specific concems i l associated with the expected impacts from power rerate (Ref.13). A copy of this power rerate environmental evaluation will be transmitted to the MPCA.

l l A new permit is expected to be issued in 1996. NSP will submit for NRC review any l changes to the permit that may affect the environmental evaluations herein.

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Table 5. NPDES Discharge Umit Summary Outfall Number Description Parameter Limit 010 Plant Cooling Flow (mgd) N/A Water intake Temperature 'F N/A Discharge Temperature *F Seasonal

Plant Capacity Factor N/A l Total Residual Chlorine and 0.2 mg/l(instantaneous Bromine maximum) 012 Holdup Pond Flow (mgd) N/A Effluent Total Suspended Solids 30 mg/l monthly average 45 mg/l 7 day average 100 mg/l daily maximum pH pH (6.0 to 9.0) 020 Turbine Building Flow (mgd) N/A Sump Total Suspended Solids 30 mg/l monthly average .

45 mg/l 7 day average l 100 mg/l daily maximum i Oil and Grease 10 mg/l monthly average 15 mg/l daily maximum g pH pH (6.0 to 9.0) 030 Screen Flow (mgd) N/A Backwash

~

040 RoofNard Flow (mgd) N/A Drains and Screen Backwash

In no case shall the maximum daily average temperature at the end of the discharge canal exceed the following limits:

(i) During the months of April through October: 95 'F (ii) During the months of November and March: 85 'F (iii) During the months of Decemberthrough February: 80 'F F- 13

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6.2.4 Increase in Circulating Water Discharge Temperature At power rerate conditions, the heat rejected by the condenserincreases. This . !

results in a corresponding increase in the circulating water outlet temperature for a given system flow rate. The steam cycle heat dissipation is provided by the Circulating Water System and the Cooling Tower System. See Sections 11.5 and 11.6 of the MNGP Updated Safety Analysis Report ( Ref.11) forinformation about the design and operation of these systems. The heat dissipation system at MNGP is the source of thermal discharges from the plant. No physical modifications or operational changes are required for these systems to implement power rerate.

The NPDES permit issued by the MPCA limits maximum average daily discharge temperatures at the end of the discharge canal (Note a to Table 5 above). Power rerate wiH not involve any changes to the MPCA discharge temperature limits. The slight discharge canal temperature increase will not result in one half of the surface width of the river temperature exceeding the 90*F maximum as delineated in the FES. Extensive field studies have been performed to confirm that the litnits imposed by the NPDES permit are conservative and assure no significant adverse impact on the environment (Ref. 5). These temperature studies ended in 1988 when the MPCA determined that 20 years of temperature monitoring had ,

adequately characterized the thermalimpacts of MNGP operation. Based on studies that evaluate the MNGP impact on the river ecosystem, cooling tower operation during the summer months has adequately prevented detrimental environmental effects, and water temperatures downstream are not high enough to harm aquatic species or impede fish migration even in summer months.

l Temperature monitoring of outfan 010 (discharge canal) is continuous, and NSP has l consistently operated MNGP in conformance with the permit's thermal discharge requirements.

The temperature increase across the intake and plant discharge is highest in faH -

and winter, when once-through cooling is employed. The temperature increase is :

lowest in summer and during periods of low river flow, when NPDES permit limits I associated with upstream average river temperature necessitate cooling tower use.

During open cycle operation at rated circulating water system flow, it is conservatively estimated that power rarate will result in an increase in temperature of water entering the discharge canal by approximately 1.7'F. During other modes of operatum, the water temperature increase wiH be less due to tempering from partial or full cooling tower operation. The calculated temperature increase of 1.7'F at the discharge canal inlet would be experienced during those months where cooling tower operation is not required to meet NPDES permit temperature i

requirements. This resultant discharge canal temperature increase is well bounded by seasonal variations. During combinations of low river flow and high atmospheric temperatures, discharge canal temperatures have approached the NPDES permit limits with cooling tower operation. During such periods NSP has reduced power at MNGP to maintain compliance with the NPDES permit. This practice wiu continue under power rerate conditions.

A 1.7'F inlet temperature increase would not involve any significant increase in harmful thermophylic organisms in the discharge canal. MNGP daily average l

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1 a l discharge canal temperatures range from 66 to 95'F when the plant is operating ,

and rarely average more than 90*F over a month. Thermophilic bacteria generally i

occur at temperatures of 25 to 80*C (77-176*F), with maximum growth at 50 to 60'C l

. (122-140*F). Pathogenic forms have evolved to survive in the digestive tract of '

mammals and, accordingly, have optimum temperatures of around 37'C (99'F).

l Similarly, pathogenic protozoans such as Naegleda tbwledhave maximum growth and reproduction at temperatures ranging from 35 to 45'C (95-113*F) and are rarely found in water cooler than 35'C (95*F).

. Because of NPDES permit requirements, MNGP discharge canal temperatures are below those optimal for growth and reproduction of pathogenic microorganisms but could permit limited survival of these organisms in summer months. The heated effluent flows over a weir at the end of the discharge canalwhich psTetes atmosphenc mixing and cooling before entry into the Mississippi River.

Temperatures in the Mississippi Riverimmediately downstream of MNGP are consistently several degrees coolei than those in the discharge canal and under normal power rarate conditions would not accelerate the propagation of these pathogenic organisms. Another factorlimiting concentrations of pathogenic microorganisms in the MNGP discharge is the absence of a seed source or

. Inoculant. Wastewater, whether municipal sewage, industrial wastewater, or agricultural runoff, is usually the source of pathogens in natural waters. Since October 1983, MNGP has pumped its sanitary wastes to the City of Monticello's wastewater treatment plant. Consequently, the power rarate does not inve!ve

, significant discharges of pathogenic microorganisms to the discharge canal aid the Mississippi River. Pathogenic organisms in the Mississippi River downstream of MNGP would typically come from upstream anthropogenic sources or animal wastes.

1 MNGP operation at the rerate power level is not expected to stimulate growth and 1 reproduction of pathogenic microorganisms in the Mississippi River downstream of the plant. Under certain circumstances these organisms may be present in the discharge canal but not in sufficient concentrations to pose a threat to downstream water users. it should be noted that many of these pathogenic microorganisms -

(e.g., Pseudomonas, Salmone#a, and Shige#a) are ubiquitous in nature, occuning in the digestive tracts of wild mammals and birds, but are usually only a problem when the host is immunologically compromised.

Given the above, the slight increases in circulating water outlet temperature due to !

power rerate will not involve any changes to NSP's compliance with the present discharge temperature limits established by the Minnesota Pollution Control Agency (MPCA) and will not result in any significant impacts to the environment.

6.2.5 Water Quality The Mississippi River at the point of discharge for MNGP is classified as Class 2Bd i by the State of Minnesota. Class 2Bd water quality is sufficient to allow for water i sports, fishing, and aquatic recreation.

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l Water quality upstream and downstream of the plant has been addressed in l considerable detail in MNGP Annual Environmental Monitoring Reports (e.g.1981 -

l 1987). Based on 20 years of water quality monitoring at MNGP, NSP submitted a l report for review by the MPCA (Ref. 5) in 1987. In 1988, the MPCA determined that MNGP operation had not adversely affected the water quality of the Mississippi River downstream of the plant and allowed NSP to reduce the monitoring program.

There is no indication that chemical discharges from MNGP have caused any j detrimental effects to the aquatic biota. The MPCA determined that water

temperature was the only physicochemical parameter significantly affected by plant operation.

Water quality monitoring programs are detailed in the MNGP Monitoring Plan in accordance with the NPDES permit. Effluent limitations and monitoring i requirements for the discharges are an integral part of the NPDES permit. Each

! outfall identified in the permit requires continuous flowrate monitoring. Chemical l discharges from MNGP have been nominally less than with those predicted in the t

1971 Environmental Report (Ref. 2). Modifications of the non-radiological drain systems or the retention basin system are not required due to power rerate, and -- 1 l biocide / chemical discharges will be consistent with existing permi'!!mits. Power 1 rerate will not introduce any new contaminants or pollutants and will not significantly 1 l . increase the amount of any those potential contaminants presently allowed for re! ease by the MPCA.

NSP has determined that an additional week of cooling tower operation may be required to support power rerate operation. This is due to the present MPCA permit requirement to place the cooling towers in operation at 68'F inlet river temperature. ;

! Based on an examination of operating temperatures, NSP has determined that the i 68'F requirement would preempt the 95'F requirement in all but a few cases. , ;

Bromine and sodium hypochlorite are injected into plant water systems at various '

concentrations to minimize microbiological fouling. The additional week of operation may require a very slight increase in normal bromine and sodium hypochlorite ,

l injection. The discharge of any additional residual halogens attributable to the extra !

week of cooling tower operation is expected to be insignificant, and effluent !

concentrations would continue to be well below the NPDES daily discharge limits. !

l 6.2.6 Mississippi RiverThermal Plume '

The results of the Seebon 316(a) demonstration (Ref.10) for MNGP determined that '

MNGP operation has had subtle alterations in the structure of some aquatic

. communities, but these impacts have been lirnsted to a small area drectly ;

downstream of the plant. Biological diversity has not suffered and may have been l enhanced by thermalinputs during certain times of the year. Based on available !

information, the minor increase in thermal output to the river due to power rarate is .

l not expected to result in any impacts on aquatic biota that are different in kind or greater in magnitude than those identified over the past 25 years and will not alter the previous 316(a) demonstration.

L; in addition to the 316(a) demonstration, NSP conducted thermal plume studies j following the construction of the discharge canalweir. These studies showed that F- 16 '

i l even in the worst case year the thermal plume disperses rapidly, is largely restricted

, to the near side of the river and is not a banier to fish movement. In addition, j depending on the ambient conditions and the distance downstream from the plant, )

L roughly 30 to 70 percent of the river is unaffected by the heated discharge. Power j rerate does not alter the water volume requirements for the heat dissipation system,

, the physical construction of the discharge canal terminus, or the temperature limits j established by the NPDES permit. Therefc,re, power rerate conditions do not j change the findings of the thermal gradient and plume studies.

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6.2.7 Cold Shock I i l

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. . Cold shock is caused by an unplanned shutdown, and the probability of an j unplanned shutdown is independent of power rerate. The projected increase in i discharge canal inlet temperature of 1.7'F at power rerate conditions will not result in a significant increase in the overall discharge canal temperature, and the ,

magnitude of the temperature decrease in a cold shock situation is not sigr.T,widiy - :

i changed. The cold shock concems of aquatic river species have been reduced by the construction of a weir at the end of the discharge canal, and by backwashing of

the traveling screens above 50*F. The weir and the backwashing controls limit the

amount of aquatic species in the discharge canal and reduce the effects of cold u 4 shock on aquatic species in the river, in addition, administrative procedures for

controlled temperature reduction of the discharge canal are in place to minimize -

thermal shock to the aquatic biota. The consequences of a cold shock event have '

i been reduced at present and power rarate operating conditions, and power rerste )

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\_ does not increase the cold shock mortality previously evaluated by the Staff in the ;

FES (Summary and Conclusions Ref. 3). I i

i 6.2.8 Impingement and Entrainment '

The 316(b) demonstration for MNGP (Ref. 6) is considered to be conservative. The study was conducted from April 1976 to April 1977 when the river level was historically low. This demonstraton concluded that" . operation of the MNGP intake t does not appear to have damaged the fish community of the Mississippi River near j

' MNGP since it began operating in 1971," and " Continued operation of the MNGP j intake should not affect the propagation of the balanced indigenous aquatic ]

1 l community of the Mississippi River." l 3 Power rerate does not effect the impingement and entrainment of organisms and l l will not cause effects that have not been previously evaluated. The circulating water i and service water system flow rates are not changed significantly. No significant

increase in entrainment of organisms or impingement of fish is anticipated at power rerate conditions above that for present operating conditions. Since initial operation, NSP has modified the MNGP intake structure to reduce impingement and ;

entrainment impacts. These modifications include a dedicated sluiceway for the ,

l traveling screen backwash system to allow wildlife species impinged on the screens ;

i to be returned to the river during backwash cycles and a log boom and bar rack at !

j the intake to minimize impingement and entertainment of drift organisms. The j practice of backwashing of the traveling screens to the river when river temperature I

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l Is above 50*F have also reduced the potential for organism impingement and entertainmentimpacts.

l The Staff estimated that operation of MNGP at average river flows and intake flows of 640 cfs may entail a possible mortality rate of up to 15% of passing drift organisms through entrainment (Summary and Conclusions, Ref. 3). Studies at

, MNGP, conducted during low flow conditions before the modifications above were l implemented, indicate an entrainment rate of 19% of the total drift orgrinisms.

I Because of the modifications and administrative controls described above and the study year bias due to low flow conditions, the Staff's estimate on mortality is consistent with plant operating data. Power rerate has no significant effect on the entrainment rate associated with present operating conditions.

Power rerate does not affect circulating water system operation or intake operating conditions. Cooling tower operation may increase an additional week due to power rerate. Power rerate will not involve significant increases in the mortality of drift organisms above present levels, and the entrainment and impingement rates of organisms have been reduced by plant modifications and these rates are consistent with that previously evaluated by the Staff. in addition, power rerate does not affect the conclusions made in the previous 316(b) demonstration.

7.0 RADIOLOGICAL ENVIRONMENTAL IMPACT 7.1 Radioactive Waste Streams The radioactive waste systems at MNGP are designed to collect, process, an'd dispose ,

of radioactive wastes in a controlled and safe manner. The design bases for these !

systems during normal operation is to limit discharges in accordance with 10 CFR 20 q and to satisfy the design objectives of Appendix l to 10 CFR 50 (Section 9 of Ref.11). '

These limits and objectives will continue to be adhered to under power rerate.

In addition, operation at power rerate conditions does not result in any changes in the operation or design of equipment in the solid waste, liquid waste, or gaseous waste systems. The safety and reliability of these systems is unaffected by power rarate.

Power rerate does not affect the environmental monitoring of any of these waste I i

streams, and the radiological monitoring requirements of the MNGP Technical ;

Specifications will not be affected. Power rerate does not introduce any new or l different radiological release pathways and does not increase the probability of an operator error or equipment malfunction that would result in an uncontrolled 'i' radioactive release. The specific effects of power rerate on each of the radioactive waste systems are evaluated below.

7.1.1 Solid Waste R

i NSP continually tracks the volume of solid radwaste generated at MNGP.

i Significant volume reductions have occurred in recent years which has made

, MNGP a recognized industry leader in wasts reduction. For calendar years

!- 1994 and 1995, the low level solid radwaste volume at MNGP was 48 and 49 cubic meters respectively. This is well below the U. S. BWR Industry Median F- 18

r 'ii i

Volume of Low-Level Solid Radwaste of 178 cubic meters in 1994 and 107 I cubic meters in 1995 (Ref.14). I The largest volume contribution to radioactive solid wastes is due to spent resin and filter sludges from process wastes. Equipment wastes from operational and maintenance activities, chemical wastes, and reactor system wastes also contribute to solid waste generation. Power rerate does not significantly affect the production or type of equipment and chemical wastes. The effect of power rerate on process wastes and reactor system wastes is evaluated below.

l A. Process Wastes Power rerate conditions may involve slight increases in the process wastes generated from operation of the Reactor Water Cleanup (RWCU)

( filter /demineralizers and the condensate demineralizers.

I ,

l The changeout limits for the RWCU filter /demineralizers are based on l

! differential pressure and effluent chemistry. It is expected that more l frequent RWCU backwashes will occur at power rerate conditions due to L chemistry limits. Power rerate will not involve changes in RWCU flow rate or l filter performance.1 NSP determined that the increase in backwashes for l

RWCU would likely be less than or equal to 4 total backwashes per year.

l The changeout limits for condensate domineralizer operation are based on l differential pressure and conductivity. The principal power rerste effect on the Condensate Demineralizer System is increased condensate flow. A consequent result of increased condensate flow is that the vessel differential pressure changeout limit will be reached more frequently. Without modification, it is expected that the spent resin generation from condensate demineralizers willincrease.

At present power levels, the percentage of solid waste generation from

. RWCU backwashes, relative to other process waste sources, is low. The assumed increase in RWCU filter backwashes at power rerate conditions will result in less than 1 cubic meter of additional resin waste per year. The condensate domineralizers, without modification, will generate approximately 4.3 cubic meters of additional resin waste per year. The worst case expected increase in spent resin volume, including both RWCU and the condensate domineralizers, is therefore less than 6 cubic meters per year.

This would result in total generation rates of approximately 55 cubic meters per year assuming 1995 generation levels. This rate is substantially below industry median values and is also well below historical radwaste generation values at MNGP.

The slight increases in solid wastes from the processes above will not result in waste volumes substantially above present levels. Moreover, in light of NSP's successful and ongoing efforts to reduce radioactive wastes at ,

MNGP, the projected increase in solid waste generation from process wastes under power rerate conditions described above is not significant and is insufficient to reverse the steady decline and the continuing downward F- 19 l

l trend in the production and activity of activated corrosion products. These efforts include the Zinc injection System, the Cobalt Reduction Program, turbine material improvements, condensate demineralizer performance improvements, and CRD water quality changes.

' Along with cobalt reduction efforts, the injection of depleted zine has reduced the amount of activated corrosion products. Cobalt 60 buildup L rates and recirculation pipe contact dose rates have declined steadily.

Recent replacement of the low pressure turbine should result in lower iron -

concentrations in the condensate being supplied to the domineralizers. The new casings are constructed of more erosion resistant materials. In 1 addition, portions of the original turbine's last stage blades were lined with stellite. The new turbine uses flame-hardened steel. This should further .

reduce the amount of cobalt in the primary system. The source of CRD water has been changed to a higher quality source which will also reduce the inventory of corrosion products. Several manufacturers have developed improved filter elements to increase the condensate demineralizer surface area in order to reduce element flux (flow / surface area). NSP has instituted a program to replace the condensate demineralizer elements at MNGP with the improved type. A set of elements are being tested in one of the five condensate domineralizer filters at MNGP, and it is expected that more elements will be replaced prior to implementation of power rerate. These ,

l filters have produced significant reductions in iron concentration. l l' Preliminary results show sustained iron effluent as low as 0.5 ppb. This modification, of itself, should compensate for the projected increases in condensate demineralizer spent resin generation described above.

NSP's successful and ongoing solid waste reduction efforts will obviate the projected slight increases in RWCU and condensate domineralizer process wastes, and spent resin generation from these inputs is therefore j insignificant at power rerate operating conditions.

i

. B. Reactor System Wastes Reactor system wastes will not increase significantly due to operation at power rerate conditions. These wastes are currently stored in the spent fuel pool and are not shipped offsite. Because of the mitigating effects of-extended bumup and increased U-235 enrichment on fuel throughput under power rerate operating conditions, the number of irradiated fuel assemblies discharged from the reactor will not increase substantially. The need for onsite dry cask storage is not expected for present operating conditions, and power rerate has no material impact on this expectation.

The volume and activity of waste generated from spent control blades and in-core ion chambers may increase slightly under the higher flux conditions

. associated with power rerate conditions. This increase, however, is i expected to be mitigated by improved longer-lived LPRM strings, improved l lower-cobalt content control rod blades, and longer fuel cycles.

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i

$. The annual environmental impact of low and high level solid wastes has i been generically evaluated by the Staff for a 1000 MWe reference reactor.

! The estimated activity content of these wastes is given by Table S-3 in.10 CFR 51.52 and is bounding for MNGP at power rerate operating conditions.

1 See Section 8.1 herein for additionalinformation.

Given the arguments above, the environment impact due to generation of i solid radwaste from power rerate conditions is insignificant.

i j 7.1.2 Uquid Radwaste I

i Although NSP is authorized to discharge liquid radwaste at MNGP per the FES ,

{ and the Technical Specifications, NSP has administratively operated Monticello i 1 as a zero radioactive liquid release plant since 1972. By Section V.D.3 of the 1 j FES, the Staff concluded that no adverse effects were expected for aquatic 4

organisms due to radionuclide releases. This was based on allowed liquid

radwaste releases. This conclusion remains bounding for power rarate i conditions as no change is expected in the zero release policy as a result of power rerate.

l

.. Filter backwashing provides input to the Uquid Radwaste System from i dewatering of sludges. The potential impact of increasing reactor thermal power on the system will be a slight increase in decanted inputs from small increases

in backwash frequency expected to occur in the RWCU System and possibly i ' from the Condensate Demineralizer System. Because of planned physical j changes to the demineralizer filter elements power rerate is not expected to 4 increase the liquid radwaste generation from the condensate domineralizers.

' See Section 7.1.1.A above. In addition, because of the zero liquid radwaste discharge at MNGP, this slight increase in input to the liquid radweste system j will be recycled instead of discharged, and therefore will not involve any j environmental concems.

j i Power rerate conditions will not result in significant increases in the volume of fluid from other sources to the Uquid Radwaste System. The reactor will .

! continue to be operated within its present pressure control band. Valve packing ,

leakage volume into the liquid radwaste system is not expected to increase. !

!_ There will be no changes in reactor recirculation pump seal flow or any other -

i normal equipment drain path. In addition, there will be no impact to the Dirty i

Radwaste, Chemical Waste, or Laundry Waste subsystems of the Uquid ;

4 Radwaste System as a result of power rerate since the operating modes and the inputs to these subsystems are independent of power rarate.

With the current low waste generation rate at MNGP and the insignificant effect of power rarate on liquid radwaste generation, it is reasonable to conclude that

! power rarate will not increase liquid radwastes above presently allowed limits.

3 in addition, power rerate will not affect compliance with the limits of 10 CFR 20 L or the guidelines of Appendix l to 10 CFR 50 for liquid effluents at MNGP. In j the Safety Evaluation for the Monticello full term operating license, the Staff j assumed annualliquid releases of 5 Cl of gross radioactivity and 20 Cl of tritium.

4 i

i F- 21 i

a 8_-_ . - - -.. - - _ , , _ -- . ,, , _ _

1 i

')

s J

Assuming these release rates, the Staff concluded that the resultant doses j j~ would be a small percentage of 10 CFR Part 20 limits and were deemed to be j as low as practicable for the Monticello Plant (Section 2.1.2 of Ref. 23).

l . i i Power rerate will not involve liquid releases in excess of those predicted by the .

j Staff in the SER for the MNGP full term operating license. If, for comparison

i. purposes, it is assumed that a hypothetical radioactive discharge at power -

F rerate conditions would be from a tank containing an activity level similar to that

normally contained in the condensate storage tank, approximately 1.5 million gallons of liquid radwaste would have to be discharged to release 5 Cl.

Moreover, no changes to Monticello's zero liquid radwaste discharge policy are i 3 expected at power rerate conditions. Given the above, the environmental I effects of liquid radwaste generation at power rerate conditions are insignificant .

and well bounded by previous Staff evaluations.

l 7.1.3 Gaseous Wastes i 1

During normal operation, radioactive gaseous effluents are released through the i l

~

- Reactor Building Ventilation System and the Offgas System pathways. These effluents include small quantities of noble gases, halogens, particulates, and

tritium. The dose to individuals from normal gaseous effluent releases at MNGP '

i. are well within the guidelines of 10 CFR 50 Appendix I and the limits of 10 CFR

! 20 for all airbome radioactive nuclides. The effluent radioactivity, in curies, of

noble gases, iodine, and particulates discharged from MNGP has been reduced

! steadily and is significantly below discharges during initial operating conditions.

4 The Staff estimated gaseous noble gas releases at MNGP at 110,376 Ci/yr and iodine and particulate releases at 0.75 Cl/yr (Section ll!.D.2.a of the FES). The .

vtual three year gaseous effluent average at MNGP (1992-1994) is 688 Ci/yr (noble gases) and 0.022 Ci/yr (iodines and particulates). Power rerate has an

i. Insignificant effect on the present production and activity of gaseous effluents, I and this effect is insufficient to reverse the ongoing steady decline in radioactive '

i- gaseous effluents from MNGP.

l The gaseous radioactivity of the reactor coolant system is, in part, a function of

the extent of fuel defects; the causes of which are independent of power rerste.

Because of changes in the core flux profile at power rarate conditions, the 4

consequences of a fuel defect for a bundle in a non-peak location at present power conditions are theoretically increased. This increase, however, continues to be bounded by the consequences for the peak bundle, since peak bundle power limits are not changed for power rerate.

]

Fuel design improvements have more than compensated (by orders of 1 magnitude) for the differential between power rerate effluents and previous staff i evaluations for gaseous effluents. New fuel designs have much better performance than the 7x7 fuel design used during initial power operations. Fuel

< reliability has increased steadily since initial operation. In comparison to 7x7 i fuel, studies in the nuclear industry have verified that overall releases due to fuel

! cladding defects have lessened from the use of 8x8 fuel assemblies. All of the fuel

} presently loaded in the MNGP core is barrier fuel. The reliability of the 8x8 barrier l

F- 22

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l l

design fuel has been reported to be better than 99.999% (Ref. 25). Reliability of the replaced 7x7 fuel was typically 99.0. In addition, because the bundle power is distributed over more pins, the fission gas release from 8x8 is theoretically less than that produced from 7x7 fuel (Section 14.7.6.2 of Ref.11). Studies conducted by the Staff have confirmed lower fission gas release rates from 8x8 fuel (Ref. 26).

{

in 1976, NSP began converting the MNGP reactor to 8x8 matrix fuel assemblies. There are no 7x7 fuel assemblies remaining in the MNGP core.

MNGP began utilizing 9x9 fuel assemblies in 1996 as a portion of the nonnal fuel reload. The 9x9 fuel is expected to be as reliable as 8x8 fuel and will have even L less fuel duty due to a lower Kw/ft. Lower fuel duty will further suppress fission gas releases. The historical fuel reliability at MNGP has increased and gaseous effluent activity has steadily declined, and these trends demonstrate the site specific validity of the above studies. These trends are expected to continue under power rerate operating conditions.

r

The effect on the environment from gaseous effluents under power rerate l conditions is further analyzed below for each environmental pathway.

I A. Offgas Stack The offgas stack receives gaseous effluent from the Steam Jet Air Ejector (SJAE) Offgas System and the Steam Packing Exhauster System. The effect of power rerate on the gaseous wastes processed by the offgas stack is not significant.

i The setpoints forisolation of the Stack release are determined by the Offsite l Dose Calculation Manual (ODCM) as stated by Technical Specification 3.8.

The current setpoint is 90,000 pCl/see which is below the 10 CFR 20 limit of 260,000 pCi/sec. Typical stack release rates are 10 pCi/sec (noble gas) and 3 E-5 pCi/sec (1-131). Assuming that power rerate will result in an increase in the release rate that is linearly proportional to power, the resultant effluent

j. increase in noble gas activity is 0.6 pCi/sec. The predicted power rerate i stack release rate is well below that assumed by the Staff in the FES and in the safety evaluation for MNGP's full term operating license (Sections 2.1.1 and 2.1.3 of Ref. 23). See Table 6 below.

The stack gas effluents for noble gases, halogens, and particulates at power rerate conditions are well below that previously evaluated by the Staff in 1-Section Ill.D of the FES. In accordance with Section Xill.A.24 of the FES and MNGP Technical Specifications, the tritium released from MNGP will remain well within regulatory limits. Although tritium release levels through the offgas stack were not quantified in the FES, tritium is an insignificant contribution to the dose rate (Section 9.2.3.2 of Ref.11), and the significant margin in the other gaseous effluents, halogens, lodines, and particulates, i

' bounds any effect that a small increase in tritium at power rerate conditions j may have on dose conclusions made by the Staff in the FES.

l F- 23

l I

r i

Assuming the noble gas generation rates above and assuming that I radioactivity from halogens, particulates, and tritium is approximately ]

proportional to the power increase, it is reasonable to conclude that the effect of power rerate on radioactive gaseous effluents from the offgas stack l pathway is negligible, and that compliance with the release limits of 10 CFR 20 and the guidelines of Appendix ! to 10 CFR 50 is maintained with i significant margin. )

I The Standby Gas Treatment System (SBGT) System is used to exhaust the reactor building during SBGT testing or the primary containment during containment venting. During startup, the mechanical vacuum pump discharges to the stack. From plant operating experience, there is no l significant increase in gaseous effluent levels during primary containment I venting or mechanical vacuum pump operation when compared to nominal I stack releases. This is consistent with Section 111.0 of the FES. I Consequently, operation of the offgas stack in these modes under power rerate conditions will not result in a significant increase in guceous effluent release levels. i i

B. Reactor Building Ventilation ;

1 Gaseous releases through this pathway are dependent on the radioactivity.

and concentration of airbome particles and gases from leakage of i contaminated systems. Leakage is independent of power rerate.

The radioactivity of the coolant from other sources will increase in linear proportion to the power increase. For most systems with leakage that can ,

reasonably be expected to contribute to reactor and turbine building airbome ~l radioactivity (e.g. secondary systems), the increase in radioactive species is theoretically neutralized by a relative decrease in the concentration. This results from a proportional increase in the steam flow rate at power rerate conditions. No credit is assumed for this mitigating effect in determining the gaseous release rate at power rerate conditions.

The design basis of the reactor building ventilation system is based on compliance with the 10 CFR 20 release limits and the guidelines of Appendix I to 10 CFR 50. The MNGP reactor building vent effluent release rate is limited in av.vidence with the MNGP Technical Specifications to 4500 pCi/sec. The nominal effluent release rate during normal plant operation is 2 pCi/sec. Typical vent release rates are 2 pCi/sec (noble gas) and 1.5 E-4 pCl/sec (1-131) Conservatively neglecting neutralization effects and also conservatively assuming an increase in activity that is proportional to power, the effect of power rarate is insignificant for this pathway. The predicted power rerate release rate is well below that assumed for the reactor building vent gaseous waste stream previously evaluated by the Staff in Section ill of the FES and in Section 2.1.1 of the safety evaluation for MNGP's full term operating license (Ref. 23). See Table 6 below.

r i

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1

-l Table 6. MNGP Gaseous Activity Releases (pCl/sec)

Plant Offgas Reactor Building Stack Effluent Vent Effluent Noble Gas 1-131 Noble Gas 1-131 ODCM Limit

90,000 4500 FES 3500 5.6 E-3 53 1.8 E 2 Assumption SER, Full Term 1430 8 E-5 1.5 E-3 Operating License Nominal 10 3 E-5 2 1.5 E-4 Operating Values Adjusted Power 10.6 3.2 E-5 2.13 1.6 E-4 Rerate Values

For sampling accuracy, the continuous ODCM Umit is based on the noble gas mixture.

" The Staff used various values of vent noble gas activity in calculating j radiological releases. See Section 2.1.1 of Ref. 23. !

Given the above, it is reasonable to conclude that the effect of power rerate on gaseous radioactive effluents from the offgas tack and reactor building vent pathways is negligible, and that continued compliance with the release limits of 10 CFR 20 and the design objectives of Appendix 1 to 10 CFR 50 is maintained with significant margin.

7.2 Radiation Levels and Offsite Dose j 7.2.1 Operating And Shutdown in-Plant Radiation The cycle average dose at Monticello has decreased at an average annual rate of 13% from cycle 11 to cycle 17. Power rerate will involve potential increases in radiation levels. These potential increases, however, are more than compensated for by physical plant improvements and administrative controls and are insufficient to reverse the significant steady decline and continuing downward trend in occupational exposures at MNGP.

MNGP was conservatively designed with respect to shielding and radiation sources. In the shleiding analysis, the analytical assumptions for reactor water fission product concentrations and corrosion products are 8 pCl/cc and 0.07 l pCi/cc respectively. The plant's administrative limit on total reactor water gamma and alpha activity for fission products and corrosion products is 0.5 pCi/cc. Normal values of reactor water fission products and corrosion products are 0.01 pCi/cc and 0.0002 pCi/ce, respectively. The typical design N-16 concentration is 100 pCi/g. A representative normal value of N-16, however, is F- 25

1 1

4 1 .

23.5 pC1/g With expected operating increases in operating activity proportional t to the proposed power increase, the design shielding assumptions remain ,

bounding with significant margin at power rerate conditions. )

The equilibrium activity concentration of corrosion products that have plated out

on reactor coolant piping and other surfaces may theoretically increase by the i square of the power rerate increase. This is primarily due to 1) the additional corrosion products introduced into the primary system from the feedwater flow increase which is proportional to the power increase and 2) the increase in

, activation events from the core average flux increase which is also proportional

^

to power. For reasons stated in Section 7.1.1.A, however, the potential increase in the volume and activity of activated corrosion products at power

, rerate operating conditions is expected to be more than compensated for, and equilibrium activity concentrations are expected to continue to decline under l power rerate conditions. Consequently, operating and shutdown radiation levels a will not increase and are expected to continue to decline under power rerate conditions.

a l

Moreover, with higher fuel bumup at power rerate operating conditions, ,

3 occupational exposure is expected to decrease proportionally to the increase in '

bumup (Ref.18 and Section 3.1.4 of Ref. 21). Dose reduction programs will also compensate for the possible increases in individual doses due to power l rerate. The plant radiation protection program will be used to maintain individual ,

j doses consistent with Al. ARA policies and well below the established limits of !

l 10 CFR 20. Routine plant radiation surveys required by the radiation protection l l program will identify increased radiation levels in accessible areas of the plant

and radiation zone postings will be adjusted if necessary. Time within radiation

)

' areas is controlled under the radiation protection program. Administrative dose -

! control limits are established well below regulatory criteria and provide ;

significant margin to that allowed by regulatory dose limits. Administrative dose l limits are not routinely exceeded under present power conditions.

7.2.2 Offsite Doses at Power Rerate Conditions For power rerate, normal operational gaseous activity levels may increase slightly. The increase in activity levels is generally proportional to the

!- percentage increase in core thermal power. This slight increase does not affect

} the large margin to the offsite dose limits established by 10 CFR 20. Doses from liquid effluents are currently zero and will remain zero under power rarate

conditions.

The Monticello Technical Specifications implement the guidelines of 10 CFR 50 Appendix I which are well within the 10 CFR 20 limits. Adjusting present values for projected power rerate increases, the estimated offsite dose at power rerate conditions is presented in Table 7 below. As shown in Table 7, the offsite dose is not changed significantly and continues to be well within the conservative

Technical Specification dose limits.

F-26

1 1

Table 7. Radiological Effluent Doses l Noble Gases 1-131, Long Uved Liquid Effluents !

Particulates, and Tritium Technical 10 mrad / year 15 mrem / year and 3 mrem / year and Specification and5 7.5 1.5 Limits mrad / quarter mrem / quarter dose mrem / quarter dose gamma; to any organ to the total body; 20 mrad / year and 10 10 mrem / year and 5 mrad / quarter mrom/ quarter dose beta to any organ Nominal 0.1% of 5 - 0.2% of 7.5 NSP maintains a Operating mrad / quarter mrem / quarter dose zero liquid Values gamma; to any organ radioactive release 0.04% of 10 policy at MNGP.

mrad / quarter beta Adjusted 0.11% of 5 0.21% of 7.5 No change in zero l Power Rerate mrad / quarter mrem / quarter dose release policy.

Values gamma; to any organ 0.043% of 10 mrad / quarter beta Power rerate does not involve significant increases in offsite dose from noble gases, airbome particulates, iodine, or tritium. Radioactive liquid effluents are not routinely discharged from MNGP. In addition, radiation from shine is not presently a significant exposure pathway (Section 4.3 of Ref.19), and this radiation is not significantly affected by power rerate.

Power rerate does not create any new or different sources of offsite dose from MNGP operation, and power rerate does not involve significant increases in present radiation levels. Therefore, it is reasonable to conclude that under '

power rerate conditions, offsite dose will remain well within regulatory criteria with no significant environmental impact.

7.3 Radiological Consequences Of AccidentsSection VI of the Final Environmental Statement (FES) Identifies nine classes of l

, postulated accidents at MNGP that were evaluated by the Staff to determine the

associated environmentalimpact. " Accidents,"in this context, includes those accidents I

evaluated for environmental consequences by the Staff in addition to design basis j accidents contained in the MNGP Final Safety Analysis Report (FSAR).

}

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The Staff used information provided by NSP in Section 10 and Appendix C of the MNGP Environmental Report (Ref. 2) to determine the associated environmental impacts. According to Section 3 of Ref. 2, the radiological effects determination is conducted utilizing reasonable assumptions, justifiable calculation models and techniques, and realistic assessments of environmental effects. The following discussion addresses the impact of power rerate on the assumptions and conclusions for the environmental accident classes. Comparisons are made, where applicable, with the accident analyses previously submitted by NSP in the MNGP Environmental Report.

7.3.1 Class 1 - Small Leaks inside Containment l

In accordance with AEC guidance for environmental reports at the time, Class 1

~

accidents were not considered within the scope. These accidents are initiated j by small spills and leaks below the Technical Specification limits inside the primary containment or secondary containment. These leaks are bounded by those analyzed under Class 8 - LOCA inside or Outside Containment. The Staff considered that an incident of this type would cause releases that are commensurate with routine effluents (Section VI of Ref. 3). Because of plant improvements, the activity concentrations of reactor coolant are considerably ;

less than that predicted by the Staff in the FES, and the above conclusion !

remains valid.

7.3.2 Class 2 - Miscellaneous Small Leaks Outside Containment l

The postulated Class 2 accident is a continuous steam leak equivalent to 7 gpm leak on the turbine building floor that releases through the turbine building roof vent. Power rerate does not increase the probability of occurrence or severity of this event. The turbine building vents were permanently secured subsequent to initial operation, and turbine exhaust air is processed through the reactor .

building ventilation system.

At power rerate conditions, the gaseous activity concentration of the reactor coolant will not increase above the assumptions used by the Staff in the original analyses. These analyses assumed a coolant activity inventory of 0.2 pCl/cc to determine radiological effects (Section 10.b.(2)(b) of Ref. 2). At power rerate conditions, coolant gaseous activity levels are significantly less because of improvements in fuel performance that have been developed and implemented since initial operation. See typical activity values in Section 7.2.1 above. The discussion of fuel reliability is contained in Section 7.1.3 above.

Consequently, the dose conclusions of Table VI-2 of the FES for Class 2 accidents remain bounding for power rarate, and the radiological consequences of these accidents are not increased.

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4 7.3.3 Class 3 - Radwaste System Failures

) Class 3 accidents are included in Table VI-2 of the FES. Class 3.1 radwaste 1

system failures are due to a single operator error or single equipment i malfunction (Section 2.2 of Appendix C to Ref. 2). NSP selected two events to

represent Class 3.1. These are 1) a liquid radwaste discharge-operator error, I

and 2) a gaseous waste discharge drain line failure. These accidents were chosen becausa these particular events were considered most probable

(Section 7.0 of Appendix C to Ref. 2).

The Staff included a Release of Waste. Gas Storage Tank Contents Accident (Class 3.2) and a Release of Uquid Radwaste Storage Tank (Class 3.3)in Class

3. NSP analyzed these events as Class 8 accidents because oflow probability

, (Sections 12.4 and Sections 12.5 of Appendix C to Ref. 2 respectively). These

accidents will be addressed as Class 3 accidents herein to conform to the 2

Staff's determination. 1 A. Class 3.1 Equipment Leakage or Malfunction

1. Liquid Radwaste Discharge 1

i Section 7.1 of Appendix C to the Environmental Report (Ref. 2) describes the assumptions used in postulating this event. The release is

, the result of an inadvertent pumping of the floor drain sample tank containing 0.7 Ci to the discharge canal for 20 minutes. The event is initiated by one of the following three single operator errors, i

The operator commences pumping without taking a batch sample.

A batch sample is incorrectly analyzed prior to discharge.

The operator pumps the wrong tank.

j From the above, it can be deduced that this accident was postulated because liquid radwaste discharges were expected to be performed -

routinely. However, evolutionary changes to the liquid radwaste system and changes in NSP's liquid radwaste discharge policies make this event extremely unlikely for current power and power rerate operating .

conditions. Liquid radwaste discharge is not routinely performed at MNGP. The plant is administratively operated as a zero radioactive j- liquid discharge plant. Operators have not discharged liquid radweste to j the canal for 24 years. Inadvertent pumping of liquid radwaste would

require an implausible sequence of events involving multiple operator errors and malicious disregard for a variety of administrative controls. A procedure to pump liquid radwaste to the discharge canal, which does

, not currently exist and would likely be created for a one-time occurrence,

. would have to be developed and approved by a variety of environmental authorities. Operators are not authorized to perform evolutions without a j valid procedure. The liquid radwaste discharge valve in the plant is a F- 29 4

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

i manual valve that is maintained shut. A sign at the valve wams the i operator that management permission is required for operation.

Additional manual valves in the discharge line are shut.

The above accident is initiated by an operator error. The offsite dose consequences of an liquid radwaste equipment failure or operator error are bounded by a tank release. The radiological consequences of discharging the entire contents of the floor drain sample tank have been I analyzed and found to be well within the limits of 10 CFR 50 Appendix 1.

I See Section 7.3.3.C below.

l l l Given the above, the probability of this postulated environmental

! accident under power rerate conditions is significantly less than that l assumed at initial licensing and would require multiple operator errors to L occur. Moreover, the severity of this event at power rerate conditions is

! bounded by that previously evaluated by the Staff.

l 2. Gaseous Radwaste Discharge Section 7.2 of Appendix C to the Environmental Report (Ref. 2)' I describes the assumptions used in postulating this event. The release is i

the result of a loss of a drain line water seal. A modification to the

Offgas System removed these water seals such that gaseous effluents

are hard-piped and positively contained within closed drain tanks.

Consequently, the probability of a malfunction of this type is significantly reduced at present and power rerate conditions because a release of this type would require a passive Offgas System pressure boundary 1 failure instead of a single equipment failure.

i Because of modifications made to the MNGP Offgas System since initial I operation, it is difficult to directly analyze this postulated accident under i power rarate conditions. These changes were described to the Staff by severa' letters. See Section 2.1.1 of Ref. 23. For an updated system description, see Section 9.3 of the MNGP USAR (Ref.11).-

To the extent that a comparison can be made, the activity concentrations at power rerate are well bounded by the assumptions used in the original analyses. Operational defects of modem fuel are significantly less than those associated with earlier fuel. See Sechon 7.1.3 above. The diffusion and equilibrium components of the offgas mixture are sigr.ific.erdly less than that expected at initial licensing. Using results from a recent actual air ejector isotopic analysis, the full power activity at the air ejector outlet under present conditions is 5100 pCl/sec at five minutes (sum of all measured radionuclides). The extrapolated rate at 0 minutes is 56,500 pCi/sec. The activity is expected to approximately increase in linear proportion to power under rerate conditions. Adjusting for the power rerate increase, the resultant value of 60,052 pCi/sec at 0 minutes is significantly below the base value of 143,000 Ci/sec at 2 minutes used in the original analyses. Consequently, the dose conclusions of Table F- 30

_ _. ._ _ i

(

F VI-2 of the FES for equipment failures remain bounding for power rerate, 1 and the radiological consequences of this accident are not increased.

Gaseous waste discharges due to operator errors were not specifically :

analyzed by NSP in the Environmental Report. Two MNGP technical specification limits incorporated after the issuance of the Final Environmental Statement address this issue. The offgas storage tank i gross activity limit of 22,000 CI (Xe-133 equivalent) is based on limiting

the offsite dose following an operator error that results in an inadvertent j release of one decay tank after 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of decay. A typical value for this parameter at current power levels is 6 Cl. Power rerate will not :

L involve significant increases in storage tank activity, and a large margin j to the limit will be maintained. A separate technical specification limits

the maximum activity at the steam jet air ejector (260,000 pCi/sec at 30 l minutes) to limit dose within regulatory criteria due to exposures from i inadvertent discharges. From the discussion in the preceding

. paragraphs and in Section 7.1.3 herein, it is apparent that operation at

[' power rerate will not involve significant increases in offgas activity above present levels, and significant margin to this limit will be maintained. I i

Gaseous waste accidents initiated by single operator errors or equipment failures are bounded by the multiple tank release analysis.

l- See Section B below. t i

j B. Class 3.2 Release of Waste Gas Storage Contents Section 12.5 of Appendix C to Ref. 2 describes this accident. The accident i is the result of a hydrogen ignition in the holdup volume. The probability of this accident is significantly less likely since NSP has installed offgas recombiners in the Air Ejector Offgas System. See Section 9.3.3.4, j Hydrogen Explosion, of the MNGP USAR (Ref.11) for a description and

. analysis.

!- The hydrogen handling design of the augmented offgas system has been !

reviewed and approved by the Staff (Ref. 29). The offgas system is designed to withstand the pressure from a hydrogen detonation. Loss of i dilution steam resulu in a recombiner train shutdown. In addition, hydrogen t is monitored, and automatic shutdowns occur well before potentially

, explosive hydrogen concentrations are reached. An explosion in the j recombiner could cause a release via the recombiner's hydrogen analyzer equipment. This rslease has been analyzed and was found to be within limits. The analyzer release is bounded by the multiple tank failure accident described below.

The Offgas System has been designed to prevent an explosive mixture from propagating beyond the recombiner system. In 1973, the Staff evaluated

, the effects of an offgas tank failure for the augmented offgas system. By -

Section 6.1 of the safety evaluation for the full term operating license (Ref.

23), the Staff analyzed the radiological consequences of a simultaneous i

j l F- 31

, u

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

failure of five offgas storage tanks. The offgas release rate was assumed to be equivalent to the prevailing Technical Specification limits (270,000 pCi/sec gross radioactivity and 2.4 pCl/sec iodine). This release rate ,

significantly bounds that expected for power rerate operating conditions.

See Table 6 and Section 7.3.1.A.2 above. The Staff concluded that the 3 dose at the site boundary was well within the values given in 10 CFR 100. l This conclusion remains valid under power rerate conditions. Power rerate

! . Will not increase the probability of this accident and will not involve operation ,

above the release rates assumed by the Staff, and consequently the '

previously analyzed dose rates continue to bound operation at power rerste I

conditions.

C. Class 3.3 Release of Liquid Waste Storage Tank Contents According to Section 12.4 of Appendix C to Ref. 2, this accident involves a l catastrophic failure of a low level radwaste tank which included a l simultaneous failure of the tank's containment basin. The activity was

released to the discharge canal. The analysis assumed a total radwaste tank activity content based on the prevailing technical specification limits.

l .

l Technical specification inventory limits are provided for undiked temporary i radwaste tanks. The technical specification limit for undiked temporary L tanks is 10 Cl, excluding tritium and dissolved or entrained noble gases. '

i Power rerate will not, of itself, involve storage of low level radwaste outside l

of the radwaste building. If storage does occur, the temporary tank radioactivity limit of Technical Specification 3.8.A.4 will not be exceeded.

Concerning installed radwaste tanks, NSP analyzed radweste tank j discharges by its Appendix I filing (Section 1.1 of Ref. 30). In this analysis it !

was assumed that the entire contents of the floor drain sample tank after treatment was discharged to the Circulating Water System with no credit for ;

Mississippi River dilution. Conservative discharges of chemical wastes and laundry wastes were also assumed. Exposures were calculated using the ;

guidance of Regulatory Guide 1.109. The resultant doses were well below t the 10 CFR 50 Appendix I limits. Power rerate will not have a material ,

impact on the effectiveness of the liquid waste processing system or on the generation and activity level of liquid wastes at MNGP. Consequently, the results of the Appendix l rMwaste tank discharge analysis are bounding for-power rerate conditions.

7.3.4 Class 4 - Events that Release Radioactivity into Primary System According to Section 2.2 of Appendix C to the Environmental Report (Ref. 2), no Class 4 events were identified for MNGP. Table VI-2 of the FES includes dose l l estimates for Class 4 events. The assumptions for these dose estimates could not be located. It is reasonable to conclude, however, that these estimates will -

remain bounding for power rerate. According to Table VI-2, Class 4 events include releases due to fuel cladding defects and releases from fuel failures induced from transients. As explained above, fuel cladding defects have been F- 32

____, ~ , - -

L l significantly reduced since initial operation due to industry improvements. in i

addition, operational limits are calculated at MNGP for each cycle to prevent transients from inducing fuel damage. These limits involve significant margin to l fuel failure. These calculations will continue to be performed, and the appropriate limits will continue to be imposed under power rerate conditions.

7.3.5 Class 5 - Events that Release Radioactivity into Secondary System l Class 5 accidents were intended to apply to Pressurized Water Reactors 4 l

- (PWRs). A justification for not including Class 5 accidents was presented in Section 9 of Appendix C. Power rerate does not impact this justification.

7.3.6 Class 6 - Refueling Accidents inside Containment Class 6 accidents include refueling and fuel handling accidents. NSP chose the design basis refueling accident and a spent fuel cask drop to represent this i

i class. The refueling accident is specifically addressed in the design basis accident section below (Class 8). The following discussion addresses the spent fuel cask drop and fuel damage from heavy loads.

The spent fuel cask drop was analyzed in Section 10.2 of Appendix C to Ref. 2.

l A cask was assumed to drop from a crane while being lowered to a flatear.

l Because of cask design integrity and fuel capability, no fuel damage was l postulated. A 1000 Ci release was assumed in accordance with 10 CFR 71 criteria.

Since initial licensing the cask drop accident has been re-evaluated by NSP at the request of the Staff, in part to support actual fuel shipments made from l MNGP. These evaluations resulted in a variety of design and administrative )

, improvements in cask handling. By its review of cask handling at MNGP in May l l 1977 (Ref.16), the Staff concluded that, "the licensee has proposed adequate l l

measures to preclude the occurrence of a cask drop accident and to mitigate its I i effect in the very unlikely event that it should occur."

l Subsequent to this action, the Staff issued generic letters that requested that I licensees determine the extent of compliance with NUREG-0612. The safety l

, concems of a heavy object drop at MNGP are mitigated by compliance with l NUREG-0612. The crane system for lifting casks at MNGP is designed for ,

!' single failures. Procedural controls and safe load paths are in place to prevent handling of heavy objects above the core and the fuel pool.. By SER dated March 19,1984 (Ref.17), the Staff concluded that the guidelines of Section l 5.1.1 and 5.1.3 of NUREG-0612 had been satisfied. For additionalinformation on cask movement and crane safety at MNGP, see NSP's response to NRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, Or Over Safety Related Equipment (Ref. 27).

' Notwithstanding NSP's stated compliance with NUREG-0612, the severity of any heavy load drop involving fuel damage is less at power rerate conditions.

The FES analyses was based on the fractional activity of 7x7 fuel assemblies.

F- 33

i NSP has replaced all the 7x7 fuel at MNGP with 8x8 or 9x9 fuel. The effect of this change in fuel design was to lower the fuel pin centerline temperature, which lowered the release of fission product gases from the fuel. This, in tum, lowered the

. available inventory of gases in the fuel pin cladding gap available for release to the environment. According to Section 14.7.6.4.2 of the MNGP USAR, the fractional plenum activity of 8x8 fuelis approximately one-tenth that for7x7 rods. Therefore, for those accidents that assume fuel cladding failures caused by a heavy object drop, the radioactivity available for release and the subsequent magnitude of the -

release to the environment is still bounded by that previously analyzed in the FES. j

. 7.3.7 Class 7 - Accidents to Spent Fuel Outside Containment i i i i

Power rerate does not significantly impact the probability or consequences of a i transportation accident. NSP has evaluated the conditions and assumptions of i j Table S-4 of 10 CFR 51.52 for MNGP operation at power rerate conditions.-

l These conditions and assumptions are applicable for MNGP operation under power rerate conditions. Table S-4 of 10 CFR 51.52 presents a generic evaluation of the environmentalimpact of fuel and waste transportation 3 accidents. See Section 8.2 below for additionalinformation. i 7.3.8 Class 8 - Accident Initiation Events Considered in the Design Basis Evaluation ;

in the SAR l The environmental impact analysis made in the FES for Class 8 accidents was ;

based on information provided by NSP in its Environmental Report (Section 11 of ;

Ref. 2). These accidents included the Recirculation Line Suction Break, the l Main Steam Line Break, and the Control Rod Drop Accident. The radwaste tank ;

failure and the offgas accident, which were originally analyzed as Class 8, are evaluated in Sections 3.2. and 3.3 above. The design basis refueling accident, .

which was originally analyzed as Class 6, is included in the Class 8 evaluation.

The methodology used to determine the offsite doses for environmental impacts of Class 8 was based in part on subjective and realistic assumptions, and the -

, FES results were expressed in estimated fractions of 10 CFR 20. It is difficult to recreate this methodology, and the value of recreating it is questionable in light of some non-conservatisms such as the assumed availability of offsite power and because of evolutionary changes in dose calculations. Therefore, for power rerate, NSP has determined that it is prudent to use contemporary and conservative Staff approved methods for calculating accident dose for design -

basis accidents that is consistent with the MNGP environmental design bases and with the appropriate regulatory acceptance criteria.

This updated radiological consequence analysis for design basis accidents presented herein supersedes the previous submittal for the environmental impact of SAR design basis accidents (LOCA, MSLB, CRDA, and RFA) in its entirety. The updated analyses demonstrate that the dose consequences are less than those previously approved by the Staff in the review for MNGP's provisional operating license and less than regulatory acceptance criteria. The postulated design basis accidents were modelled and analyzed to determine F- 34 i

-- --- --m-- -

numerical dose outcomes under power rerate conditions for direct comparison with regulatory limits. The radiological consequences of these design basis accidents represent the worst case environmental contequences. The regulatory acceptance criteria for these accidents is delineated by 10 CFR 100 for offsite doses and by GDC 19 of 10 CFR 50 Appendix A for control room habitability. The results of these analyses demonstrate that power rerate has an insignificant environmental impact. The accident doses for postulated environmental accidents under power rerate conditions are bounded by those previously evaluated in the provisional operating license, and doses from design basis accidents remain well within regulatory guidelines.

The results of these studies are presented in Section 9.2 of Exhibit E to this licensing submittal (Ref. 24). It is noteworthy that these accidents were conservatively analyzed by assuming an initial power level of 1918 MWt. This postulated power !avel is 102% of a bounding analytical power level of 1880 MWt and 108% of the proposed rerate powerlevel. A notable exception is the MSLBA, which for the purpose of establishing worst-case radiological consequences, is assumed to occur during low power conditions. Section 9.2 of Ref. 24 demonstrates that offsite and control room dose levels under power rerate conditions are well within regulatory guidelines. Control room doses for the RFA and CRDA are bounded by the MSLBA. The assumptions are conservative with respect to power rerate operating conditions, shielding, and dose, and the methodologies have been previously approved by the Staff.

Given the above, the radiological consequences of design basis accidents

. under power rerate conditions are within the acceptance criteria of GDC 19 of Appendix A to 10 CFR 50 and of 10 CFR 100 and do not involve any signi'icant impact to the human environment.

7.3.9 Class 9 - Severe Accidents The environmental effects of severe accidents outside the design basis of protection and engineered safety systems were not evaluated in the MNGP FES. (See Section VI.A of Ref. 3.) The Staff did not evaluate these sequences on the premise that sufficient design conservatism, quality assurance, testing, and multiple physical barriers were in place such that the probability of a severe environmental accident is small, and the environmental risk of a Class 9 accident was extremely low. Power rerate will not involve any changes to the Staffs assumptions made in arriving at the above conclusion.

Notwithstanding the above, NSP conducted an evaluation to detennine the effect of power rerate on the MNGP IPE and IPEEE submittals. No new.

vulnerabilities were identified. See Section 10.5 of Attachment E to this licensing submittal (Ref. 24) for additional information. In the Summary and Conclusions Section of the FES, the Staff stated that the licensing of MNGP involved the " creation of a very low probability risk of accidental radiation exposure to nearby residents." There is nothing in the power rerste PRA that would serve to nullify the Staff's previous conclusions on environmental risk from accidents.

i F- 35 un _-. _ _ _ _ . . _

1 J

7.4 Other Potential Environmental Accidents Power rerate does not significantly change the inventory, storage, usage, or control requirements for chemicals, industrial gases, oil, oil products, or other hazardous ~

substances. Power rerate will not require the introduction or use of any new hazardous.

substances. Power rerste will not result in a significant increase in the probability or consequences of an oil spill, chemical spill, industrial gas release, or other event involving a non-radioactive hazardous substance.

8.0 ENVIRONMENTAL EFFECTS OF URANIUM FUEL CYCLE ACTIVITIES AND FUEL AND RADIOACTIVE WASTE TRANSPORTATION 8.1 Compliance With 10 CFR 51.51, Uranium Fuel Cycle Environmental Data (Table S-3)

Because Table S-3 of 10 CFR 51.51 was adopted after MNGP received its operating ;

l license, the MNGP FES does not contain a uranium fuel cycle environmental analysis !

l similar to Table S-3. The Staff, however, included the Table S-3 fuel cycle .

environmental data in its review of the MNGP full term operating license (Enclosure 3 of )

Ref. 22). The Staff concluded that the fuel cycle effects of Table S-3 combined with '

operation of MNGP did not significantly impact the environment. The impact of power I rerate on the Staff's previous evaluation is increased fuel bumup and U-235 i enrichment. .

The environmental effects of fuel cycle activities under power rerate conditions continue to be bounded by the Staff's evaluation that incorporated Table S-3 into the MNGP licensing basis as described above. The evaluation assumed that the fuel cycle would support a reference reactor of 1000 MWe that operated at 80% capacity factor which results in an adjusted daily electricity production of 800 MWe during a reference <

reactor year (RRY). Unoer power rerate conditions the daily output at 100% capacity is less than 650 MWe, and MNGP will not exceed the assumptions of the reference reactor year (RRY) used in the evaluation and as defined by the NRC in Ref. 21.

l The data presented in Tables S-3 and S-4 are, in part, based on an average bumup assumption of 33,000 mwd /MtU and a U-235 enrichment assumption of 4 wt.%. Fuel consumption is expected to increase under power rarate conditions such that the batch average bumup of the fuel assemblies will be in excess of 33,000 mwd /MtU but less than 60,000 mwd /MtU. The U-235 enrichments levels will also increase to greater than 4 wt.% but less than 5 wt.% to support extended bumup. The Staff has previously evaluated the environmental impact of increased bumup to 60,000 mwd /MtU with U-235 fuel enrichment to 5% wt.% on the conclusions of Table S-3. See Federal

, Register 53 FR 6040 dated February 29,1988 (Ref.18). Although some radionuclide l

inventory levels and activity levels are projected to increase, the Staff noted that little or no increase in the amount of radionuclides released to the environment during normal operation was expected. The Staff determined that the incremental environmental effects of increased enrichment and bumup on transportation of fuel, spent fuel, and waste were not significant. In addition, the Staff recognized the salient environmental benefits of extended bumup such as reduced occupational dose, reduced public dose, reduced fuel requirements per unit electricity, and reduced shipments. The Staff F- 36

~. .-. - _

l' concluded that the environmentalimpacts described by Table S-3 were bounding and l

were also applicable for burnup levels to 60,000 mwd /MtU and U-235 enrichment l levels to 5 wt.%.

Table S-3 does not include a determination of the environmental effects of the gaseous ,

i effluents of Rn-222 and Tc-99. By Enclosure 3 to Ref. 22, the Staff evaluated these effluents and concluded that the environmental impact from radon releases was not i significant and that consideration of Tc-99 releases at MNGP was unnecessary. In i addition, an industry study performed by the Atomic Industry Forum (Ref. 20) concluded that extending fuel bumup to 60,000 mwd /MtU and increasing U-235 enrichment to 5 wt.% results in insignificant environmental consequences from Rn-222 i and Tc-99.

8.2 Compliance With 10 CFR 51.52, Environmental Effects of Transportation of Fuel l and Waste (Table S-4)

The environmentalimpacts of transporting fuel and waste were analyzed by the Staff in ' ;

the FES. For the purposes of evaluating the environmentalimpact of power rarate at' ,

MNGP, however, NSP will demonstrate compliance with the more recent envinnmental l licensing criteria of Table S-4 of 10 CFR 51.52. Table S-4 presents a generic assessment of the environmentalimpacts of transporting fuel and waste to and from a ;

, reference reactor. For power rerate operating conditions, this demonstration supersedes the previous NSP submittals conceming environmental effects of ;

transportation of fuel and waste including Sections 11.0 of 13.0 of Appendix C to the i Environmental Report (Ref. 2).

Operation of MNGP under power rerate conditions meets all the conditions of part (a) of ]

10 CFR 51.52 with the exception of the enrichment and bumup conditions as described in the succeeding paragraphs. Each subsection of part (a)is addressed below for l power rerate conditions. The enrichment assumptions of paragraph (a)(2) and the bumup assumptions of paragraph (a)(3) are addressed separately below.

(a)(1) The core thermal power under rerate conditions is less than 3800 MW.

(a)(2) The reactor fuelis in the form of sintered uranium dioxide pellets, and the pellets are encapsulated in zircalloy rods.

(a)(3) No irradiated fuel assembly is shipped until at least 90 days after it is discharged from the reactor.

(a)(4) With the exception of irradiated fuel, all radioactive waste shipped from the i reactor is packaged and in a solid form.

L (a)(5) Unirradiated fuel is shipped by truck; irradiated fuel is shipped by truck, rail, or barge; and radioactive waste other than irradiated fuel is shipped from the reactor by truck or rail.

( (a)(6) in accordance with parai aph (a)(6) of 10 CFR 51.52, the environmental j impacts of transportation of fuel and waste to and from the reactor at power rerate l

F- 37 l

l l

i '

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

conditions with respect to normal and accident conditions of transport are as set forth in Table S-4 with the exception of fuel enrichment and bumup assumptions.

The values in the table represent the contribution of the transportation to the environmental costs of operating at power rerate conditions.

NSP complies with the conditions of Table S-4 for the MNGP power rerate except for the U-235 enrichment and fuel bumup assumptions. The conservatism and continued applicability of Table S-4, however, has been previously evaluated by the l Staff for enrichment to 5 wt.% and for average bumup to 60,000 mwd /MtU (Ref.

j 18). As stated in Section 8.1 above, power rerate will not involve operation outside j these limits.

l .

l l

Section V.E. of the FES contains an analysis of expected exposures during normal conditions of transport.Section VI.B contains an analysis of transportation accidents.

Table S-4 provides a generic assessment of the environmentalimpacts of transporting j fuel and waste to and from the reactor. Therefore, given NSP's compliance with Table <

i- S-4 for power rerate, these sections of the FES would be superseded upon approval of l the proposed power rerate changes. Compliance with the conditions of part (a) of 10 CFR 51.52, as described above, would then become part of the MNGP license. I 9.0 DECOMMISSIONING EFFECTS Other than financial set asides, the environmental effects of decommissioning were not evaluated by the Staff in the Monticello FES (Section Xill, Question 45, and Section Vill.C l of Ref. 3). The AEC deferred this review until the submittal of a decommissioning plan.

l NSP's decommissioning plan for Monticello will be submitted in accordance with regulatory criteria. Power rerate does not involve increases in decommissioning cost estimates and does not affect NSP's ability to maintain sufficient financial reserves for decommissioning.

The potential impact of power rerate on decommissioning is due to increases in feedwater ,

flow rate and increased neutron fluence. These effects could increase the amount of ;

activated corrosion products and consequently increase post-shutdown radiation levels.

For reasons cited in Section 7.1.1.A herein, however, the potential increases in activated corrosion products are obviated by other ongoing plant improvements.

10.0 CONCLUSION

S i Power rerate does not involve any significant impacts to the environment. There are no ;

new significant environmental hazards in addition to those previously evaluated. The environmentalimpacts and adverse effects identified by the Staff for MNGP operation at 1670 MWt in the Summary and Conclusions Section of the Final Environmental Statement (Ref. 3) continue to bound plant operation at power rerate conditions. The proposed changes do not, individually or cumulatively, affect the human environment. There is no significant change in the types or amounts of plant effluents.' Power rerate does not l Involve significant increases in individual or cumulative occupational radiation exposure.

The effect of power rerate on the environment does not prevent continued compliance with any MNGP environmental permit. None of the license conditions for environmental protection will be changed for power rerate. No effluent limits will be exceeded, and the F- 38

l present large margins to these limits will not be significantly changed.' Power rerate does not involve an increase in the discharge of hazardous substances, contaminants,' or pollutants and does not involve the use of any new hazardous substances, contaminants, or pollutants.

Power rerate does not involve any changes to air quality or water quality, it does not result

'in any changes to land usage and has an insignificant effect on groundwater and surface water usage. The amount of water withdrawn and consumed from the Mississippi Riveris not significantly increased above that previously evaluated. The slight increase in discharge canal temperature has an insignificant effect on river temperature and will not result in any changes to aquatic blota other than those previously evaluated. Power rarate will not involve new or different discharges of contaminants and does not involve changes ,

to any bioaccumulation effects for aquatic organisms. Power rerate does not accelerate i the introduction of any microbiological organisms into surface water pathways or significantly increase the population of any known pathogens. The quality of drinking water is not affected.

Power rerate does not involve any changes to wildlife habitat and does not result in any significant changes to aquatic or terrestrial blota. There are no deleterious effects on the diversity of biological systems or the sustainability of species due to power rerate. Power .!

rerate does not involve any additional changes to the stability and integrity of ecosystems.

Power rerate does not affect the previous conclusions on impingement or entrainment. ['

Power rerate does not affect MNGP's compliance with Sections 316(a) or 316(b) of the i

Federal Water Pollution Control Act.

Power rerate does not significantly change any doses to the public from radiological j effluents, and offsite doses will continue to be well within regulatory limits. By Section l 2.1.3 of the Safety Evaluation for the MNGP full term operating license, the Staff '

concluded that "the release of radioactive material in liquid and gaseous effluents from the -

Monticello Nuclear Generating Plant will meet the requirements of 10 CFR 50 for keeping i such effluent levels to unrestricted areas as low as practicable and will result in doses that are a small percentage of the 10 CFR 20 limits." The Staff based this conclusion on assumptions for effluent releases that bound releases _ expected for power rarate.

Occupational dose will be maintained well within regu;atory limits, and changes in radiation levels will not significantly increase the dose to the MNGP work force. For accident dose, 4 the methodology for certain design basis accidents was updated. This methodology is consistent with previously approved Staff methods, and the resultant dose is well within the applicable regulatory limits. Power rerate does not involve significant increases in the t) probability or consequences of previously evaluated environmental accidents.

This environmental evaluation has demonstrated that in most cases power rerate does not '

involve any environmental impacts that are different from those previously evaluated for the present power level. Where environmental impacts which differ from those previously evaluated have been identified, these impacts have been shown to be insignificant and '

well within regulatory environmental acceptance criteria.

l!

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11.0 REFERENCES

1. Monticello Generating Station Capacity increase System Reliability Study, Report to the MAPP Design Review Committee, NSP Delivery System Planning and Engineering Department, January 1996 1

2. Northem States Power Company, Environmental Report, Monticello Nuclear Generating Plant, November 5,1971

3. Final Environmental Statement Related to Operation of Monticello Nuclear Generating Plant Northem States Power Company Docket No. 50-263, United States Atomic Energy Commission Directorate of Licensing, November 1972

4. Northem States Power Company Environmental Report Supplement 1, Monticello Nuclear Generating Plant, April 4,1972

5.

Environmental Monitoring and Ecological Studies Program for the Monticello i Nuclear Generating Plant 1987 Annual Report, Water Monitoring Summary," D. J.

Orr, Environmental and Regulatory Activities Department, Environmental Sciences i Section,1967 i

6. "Sectir 3' W ) Demonstration for the Monticello Nuclear Generating Plant on the MissisC 9 <er at Monticello, Minnesota," NUS Corporation, Ecological Sciences Division, Ptt ourgh, Pennsylvania,1978 i

7. 1990 CPH-5-25,1990 Census of Population and Housing, Summary Social, Economic, and Housing Characteristics, Minnesota, U.S. Dspartment of Commerce, June 1992

8. NUREG-1437, Vol. I, " Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Rev. C, U.S. NRC Office of Nuclear Regulatory Research, December :

1995 l

9. "Northem States Power Company Before the Minnesota Public Utilities Commission, Application for Resource Plan Approval 1996-2010," July 1995, Docket Number: E-002/RP 95-589 10.* Effects of a Heated Discharge on the Ecology of the Mississippi River: 316(a) Type i Demonstration on the Monticello Nuclear Generating Plant, Monticello, Minnesota," NUS Corporation, Ecological Sciences Division, Pittsburgh, Pennsylvania,1975 l

11. MNGP Updated Safety Analysis Report, Rev.13

12. Letter from J. Bodensteiner, NSP, to T .J. Mader, MPCA, "Monticello Generating l Plant NPDES Permit #MN0000868 Application for Renewal," June 30,1995

13. Letter from T. Mader MPCA, to M. Hestick, NSP, October 27,1995

, 14."1995 WANO Performance Indicators for the U.S. Nuclear Utility industry," The Nuclear Professional. Vol.11, Number 2, Spring 1996 F- 40 t

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

15. DOE /EE-0040," Questions and Answers About EMF Electric and Magnetic Fields Associated with the Use of Electric Power," January 1995, National Institute of -

Environmental Health Sciences and U.S. Department of Energy. .

16. Letter from D. K. Davis, NRC, to L O. Mayer, NSP, May 19,1977

17. Letter from D. B. Vassallo, NRC, to D. M. Musoif, NSP, " Control of Heavy Loads .

(Phase 1)," March 19,1984

18. " Extended Bumup Fuel Use in Commercial.LWRs; Environmental Assessment and ;

Finding of No Significant impact," Federal Reoister 53 FR 6040, February 29,1988 ,

19. Letter from R. O. Anderson, to US NRC Document Control Desk, *1995 Annual Radiation Environmental Monitoring Report," April 24,1996

20. "The Environmental Consequences of Higher Fuel Bum-up," AIF/NESP-032, Atomic ' '

industrial Forum, June 1985

21. " Assessment of the Use of Extended Bumup Fuelits Light Water Power Reactors,"

j. NUREG/CR-5009, February 1988 l 22. Letter from D. G.'Eisenhut, NRC, to L O. Mayer, NSP, issuance of Full Term .

! Operating License DPR-22, January 9,1981

23. Safety Evaluation By The Directorate Of Ucensing U.S. Atomic Energy Commission For I Northem States Power Company Monticello Nuclear Generating Plant Unit No.1 Full Term Operating License, February 5,1973

24. NEDC-32546P, " Power Rorate Safety Analysis Report for Monticello Nuclear Generating Plant," (Exhibit E of this submittal)

25. NUREG/CR-3950, " Fuel Performance Annual Report for 1989," June 1992

26. NUREG-0559, "A Comparative Analysis of LWR Fuel Designs," Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, July 1980 l

27. Letter from W. J. Hill, NSP, to NRC Document Control Desk, " Response to l

NRC Bulletin 96-02 Movement of Heavy Loads Over Spent Fuel, Over Fuel in j the Reactor Core, Or Over Safety Related Equipment," May 8,1996 28.NUREG-1437, Vol. I, " Generic Environmental Impact Statement for Ucense

Renewal of Nuclear Plants," Draft Report for Comment, US NRC Office of Nuclear l Regulatory Research, August 1991 29.Monticello Nuclear Generating Plant, " Answers to DRL Questions of June 3, 1971," October 1971 l

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l 30. Letter from L. O. Mayer, to D. L. Ziemann, NRC, " Appendix l Filing Supplement 1," July 21,1976 f

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Reference:

Reference:

Reference:

SUMMARY

SUMMARY

1.0 INTRODUCTION

10.0 CONCLUSION

11.0 REFERENCES

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