ML20202H706

Proposed Tech Specs Increasing Plant Max Power Level from 1670 Mwt to 1775 Mwt

ML20202H706
Person / Time
Site:	Monticello
Issue date:	12/04/1997
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML19317C757	List: ML20202H675 ML20202H686 ML20202H706
References
FACA, NUDOCS 9712100345
Download: ML20202H706 (62)

Text

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EXHIBIT B

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i REVISION 1 TO LICENSE AMENDMENT REQUEST DATED JULY 26,1996

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PROPOSED CHANGES MARKED UP ON EXISTING OPERATING LICENSE AND TECHNICAL SPECIFICATION PAGES NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA LICENSE NO. OPR 22 DOCKET NO. 50-263 1

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9712100345 971204 DR ADOCK 0 3

Exhibit B Monticello Nuclear Generating Plant Revision 1 to License Amendment Request 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.

Affected Page Associated Change Affected Page Associated Change License page 3 A

30*

E, J 4

A 19 E

14 A

38 E

15 A

50*

F, J 6

B 53' G

56 B

65 G

16 8

104 H

114 8

39 l

23 C

112 1

24 C

68 l

150 C

il i

28 DJ iil 1

37 D

49 J

M J

Pages reflect proposed changes submitted with amendment request entitled

• Surveillance Test Interval /Atlowed Outage Extension Program - Part 2,* dated July 5,1995, TAC M92948.

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3 4

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required eny byproCuct, source or special nuclear material without restriction to chemical or physical form, for sample analys'.s 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 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 1.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of MM-mega-l5 watts (thermal) 8M g

j 2.

Technical Soecifications The Technical Specifications contained in Appendix A, as E

revised through Amendment No. 97, are hereby incorporated in the lI license. The licensee shall operate the facility in accordance with the Technical Specifications.

j 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 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 E

Security Plan," with revisions submitted through November 30, 1987; "Monticello Nuclear Generating Plant Guard Training and Qualification

E Plan," with revisions submitted through February 26, 1986; and u

"Monticello Nuclear Genersting Plant Safeguards Contingency Plan,"

E with revisions submitted through August 20, 1980. Changes made in 4

accordance with 10 CFR 73.55 shall be implemented in accordance vid E

the schedule set forth therein.

4

• 4 i

4. Protective Function - A system protective action which results from the protective action of the channels bonitoring a particular plant condition.

R. Rated Neutron Flux - Rated flux is the neutron flux that corresponds to a steady-state power level of 16,90- thermal megawatts.

15 es r '

S. Rated Thermal Power - Rated thermal power means a steady-state power level of 4699-thermal megawatts.

l :2 1775-T. Reactor Coolant System Pressure or Reactor Vessel Pressure - Unless otherwise indicated, reactor vesral pressures. listed in the Technical Specifications are those existing in the vessel steam space.

c U. Refuelinz Operation and Refueline Qutare - 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 i

of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled

' refueling outage; however, where such outages occur within 4 months of the completion of the previous refueling outage, the required 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 cladd*ng and primary 7

system integrity are assured.

Exceeding such a limit is cause for plant shutdown and review by the s'

l

[

Commission before resumption of plant operation. Operation beyond such a limit may not in itself result in serious conrequence2 but it indicates an operational deficiency subject to-regulatory review.

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

(

closed and the following conditions are met:

i r

)

1. At least one door in each accery opening is clesed.

a f

I

2. The standby gas treatment system is operable.

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

r i

X. Sensor Check - A qualitative determination of operability by observation of sensor behavior during I

operation. This determination shall include, where possible, comparison ther independent

(

sensors measuring the same variable.

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b

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AMIL1i-m s-2.3

-The abnormal operational transients applicable to operation of the Monticella Unit.have beenfanalysed i

throughout the spectrum of planned operating conditions up to the thermal power level of 1479-Mut.

l'y l The analyses were based upon plant operation in accordance with the operating map. The licensed maximum Tl s

power level 1474-MWe represents the maximum steady-state power which shall not knowingly be exceeded.

i ,

I1TS j

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as i

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 l

j' tranisent results as determined by the current analysis model. Conservatism incorporated into.the i

i transient analysis is documented in Reference 1.

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2.3 BASES 14 i

REV 10^ 7/20/69-I i

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Esses 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 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 recson. 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 allovable 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 Scra:n 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

1. p MWt).

Because fission chambers provide the basic input signals, the APRM system responds l _f, 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 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 120n ceram trip setting, none of the abnormal operational transients analyzed violete the fuel Safety Limit and there is a substantial margin form fuel damage. Also, f

the flow biased neutron flux scram (specification 2.3.A.1) provides protection to the fuel safety

/

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

l 2.3 BASES 15 REV -167 9/17/96-

.+

s.

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CIADDING INTECRITY 2.3 FUEL CIADDING INTERGRITY Anolicability Aeolicability Applies to the interrelated variables associated Applies to trip settings of the instruments and with Tuel thermal behavior.

devices which are provided to prevent the reactor system safety limits from being exceeded.

Obiective:

Objective:

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.

Soecification:

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 1.

APRM - The APRM flux scram trip setting minimum critical power ratio (MCPR) less than shall be:

1.08*, for two recirculation loop operation, or

a. For two recirculation loop operation less than 1.09* for single loop operation, (TLO):

45.4*

shall constitute violation of the fuel cladding S $ 0.66W + 70%

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integrity safety linit.

where S - Setting in percent of rated thernal pow A rated power being 167G hvr / N m,5/ %,

W - Percent of +hebflow required to produce a M 8

6 core flow of 57.6 x 10 lb/hr b.

For single recirculation loop operation (SID):

bb 69 SS 0.58(W - 5.4) + 42%

l

/

MCPR values are for cycle 18 only.

2.1/2.3 6

REV -149-10/20/07

i TABLE. J. Z. 3 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 Evoass Conditions **

Channels per Instrument Channels Required 3metion Trio Settints Refuel Startuo Run Trio System ner Trin System Corwittions*

6. SER s.

Upscale 55x105 I

X(d) 2 1(Note 1, 3, 6)

A or 8 or C cps b.

Detector X(a)

X(a) 2 1(Note 1, 3, 6)

A or B or C l

not fully inserted 7

i

2. IEg a.

Downscale 23/125 X(b)

X(b) 4 2(Note 1, 4, 6)

A or B or C full scale b.

Upscale s108/125 X

X 4

2(Note 1, 4, 6)

A or 8 or C full scale

3. APRM U

X 3

1(Note 1, 6, 7)

D or E (1)pscale a.

TIE.

58 '

l Flow-f 0.66U + 56%

Biased (Note 2)

Q w

(2) SLO 4

m i

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Flow f 0.-M(U - 5.4) + Set i

Biased (Note 2)

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(3) High $ 108%

Flow Clamp i

b.

Downscale

= 3/125 full scale X

3 1(Note 1, 6, 7)

D or E l

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E,atual,,,s Jaswsned.* MOC m SM,.f ty n

re nle co s/ths, in edw ik treld.on dreendreios rk aceyhb:4Jy of AwasA L s gs-I-ojs k

op %.

Bases Continued-Mt u.a)

Maximum Extended load Line Limit Analyses have been performed to allow operation at higher powers at flows j

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). E::: an:1 :: h..e r.ct changed-the allowed-7 e r rette-fer eingle leep eperation. The supporting analyses are discussed in CE NEDC-31849P report

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(

Reference:

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

/.

/

Increased Core Flowfa(red nalyses have been performed to allow operating at flows above 100% for powers equal to g

or less than 1004 (See Core Operating Limit Report). The supporting analyses are discussed in General Electric NEDC-31778P report (

Reference:

Letter from NSP to NRC 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 power provides adequate thermal margin between the setpoint and the safety limit, 25% of rated. 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 control rod patterns are constrained to be uniform by operating procedures.

Uorth of individuajL rods is very low in a uniform Tod pattern. Thus, of SI.1 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 perks, 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 St 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 APEM 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 Specifica-tion 2.3.A.I.

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 59% of design due to the deviations discussed on page 39.

B.

Deleted 2.3 BASES 16 REV 12' 1/27/??

Next Page is 18

?

L Bases 3.5/4.5 continued-l t

F. Recireslation System The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel desi n limits. To minimize the likelihood of a thermal-5 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 l

each fuel cycle the limits are contained in the Core Operating 1.imits Report. Specific directions are provided to avoid operation in this region and to immediately exit upon an entry. Entries into the exclusion region are not part of normal operation. An entry may occur as the r sult of an abnormal event 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 7

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

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

Power 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 i

outside of the stability exclusion region. Without these centrols, it is theoretically possible to operate the reactor in such a manner as to cause an oscillation outside of the exclusion region. A nominal 54 power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the analytically defined exclusion region. The buffer region may be entered only when the power distribution controls are in place.

3 Continu s operation with one recirculation loop was analyzed and the adjustments specified in specification 3.5.F.g were determined by NEDO-24271. June 1980. "Monticello Nuclear Generating Plant Single loop Operation *-i Specification 3.6.A.2 governs the restart of the pump in an idle recirculation loop. Adherence to this sp cification limits the probability of excessive flux transients and/or thermal stresses.

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3.5/4.5 114 REV 157 7/17/"

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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 ranse of the fu 1 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 translents.

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'Ihe 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 fand they must limit the reactor pressure to no more than 110 percent of design pressure.

"Ihe safety / relief valves are sized according to the Code for a condition of MSIV closure while operating mr 7490-MWe, follow,sd by no MSIV closure scram but scram from an indirect (high flux) l means.

With the safety / relief valves set as specified herein, the maximum vessel pres-Only E of the eight va,1ves are q

4 sure remains below the 1375 psig ASME Code limit.

assumed to be operable in this analysis and the valves are assumed to open ate above I

their setpoin with a 0.4 second delay. %e upper limit on safety / relief valve setpoint is established by the design pressure of the liPCI and RCIC systensgcf 112^ p;ig.

ol oro9tsj The operator will set the reactor coolant high pressure scram trip setting at 1075 psig 7

or lower. Ilowever, the actual setpoint can be as much as 10 psi above the 1075 psig y

indicated set point due to the deviations discussed in the basis of Specification 3.1.

1 In a like manner, the operator will set the reactor coolant syyt afety/ relief valve V

initiation trip setting at 1120, psig or lower. Ilowever, the actua. set point can be as

'v much as 2 pst above the 1120 psi indicated set point due to the deviations discussed in the basis of Specification 3.6.

g,g

( no9 r+3 + 4 )

i A violation of this specification is assumed t'o 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 7he desy, cup.Cily af ik H/t1 a,J/ctc yshns hcs boea (**Smdarby c4 e-starle.f 4 4* ecephfie af pr,ss,ars 3% yee/w /h.* /La 3di l"kef f

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2.4 BASES

>* be 5 !j'*.nt */ st31 s.j. yter s<J Ret y**an'* * '*].***/ fQ31SH"'

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e Bases Continued 3.6 and 4.6 D.

Coolart Y==k==e he 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 leakaga 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 Imakage 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 concainment. 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 l

is collected in the drywell floor drain sump.

Both sumps are equipped with level and flow trans-l, mitters connected to recorders in the control room. An annunciator and computer alarm are pro-vided 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 tem-perature and pressure. Systems connected to the reactor coolant systems boundary are also monitored l

for leakage by the Process Liquid Radiation Monitoring System.

l!

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

1 E.

Safetv/ Relief Valves

. h,4* est/t v 09+ MI A es-4=/

tk valve deterioration is detected.

Testing of sH safety / relief valves each refueling outage ensures that aq j

A0".0 1,iler y

Atolerance value ofgg for safety / relief valve setpoints isgpecified in4 ction III ef Om end Ne;;ure i";;;;l _dc.

Analyses have been performed with414-valves assumed ce M h!;h. As discussed l

the 1375 psig Code limit is not exceeded in any case.f la f*d 3% do.e flee.r salpeaf o l

in the Section 2.2 Bases, t

3.6/4.6 BASES 150 r

REV 00 11/16/"'

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TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRLMENT REQUIRDtENTS Modes in whicn func-Total No. ot Nin. No. ot. Operable tion must be Oper-Instrument or Operating Instru-Limiting able or Operatlng**

Channels per ment Channels Per Required Trip Function Trip Settings Refuel (3) Startup Run Trip System Trip System (1)'

Condition

• 1.-

Mode Switch in Shutdown

.X X

X 1

1 A

2.

Manual Scram X

X X

1 1

A' 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 a

a. High-High

b. Inoperative

c. High Flow Clamp $ 120 t

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I 5.

High Reactor e

Pressure s 1075 psig X

X(f)

X(f) 2 2

A (See Nate 9) 6.

High Drywell Pressure s 2 psig X

X(e.f) X(e,f) 2 2

A' (See Note 4) e 4.=4,5 7.

Reactor Iow lb s

Water Imvel 2 7 in.(6)

X X(f)

X(f) 2 2

A 8.

Scram Discharge

^

volume High Level

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 Low Vacuum a M in. Hg X(b)

X(b f) X(f) 2 2

A'or C la s

Z Z.

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3.1/4.1 2s 4

REV 1M 1/"/*3 -

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Bases Continued:

3.1 condenser vacuum initiates a closure of the turbine stop valves end turbine bypass valves which eliminates the heat input to the condenser.

Closure of the turbine stop and bypass valvea causes a 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.

limit from being axceeded in the event of a turbine trip transient without bypass.The tu: vine stop Reference FSAR Section 14.5.1.2.2 and supplemental informatica 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 23" Hg vacuum, stop i2

^

valve closure occurs at 20" Hg vacuum, and bypass closure at 7" Hg vacuum.

22.

2 The main steamline isolation valve closure scram is set to scram when the isolation valves are s10% 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 Section 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 functicns appropriate to the particular plant cperating status. Reference Section 7.7.1 FSAR.

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

The IRM system provides protection against excessive power levels and short reactor periods in the a

t 3.1 BASF.S 37 REV 133 0/10/02 1

4

Table 3.1.1 - Continued Dele lzL m

6. C. se-- i--he -- tv m eer 1--a % r--- t ie i n <-

b.

x-

-x-ec- - r ts- -- t s-r-i -- --

2 Isi

--r 7.

Trips upon loss of oil pressure to the acceleration relay.

l i

8.

1.imited 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.

N s Reauired Conditions when minimum conditions for operation are not satisfied. (ref. 3.t.5)

A.

All operable control rods fully inserted.cithir ? her e.

"f* *

Buyes fer B.

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

Af9 4 4' M C.

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

Q D.

Reactor power less than 454 (M1. MWt. ).

N

'NtM

• • Allowabie Bvoass Conditions It is permissible to bypass:

The scram discharge volume High Water Level scram function in the refuel mode to allow reactor protection a.

system reset. A rod block shall be applied while the bypass is ir'effect.

The Low Condenser vacuum and MSIV closure scram functionin Jhe/

b.

is below 600 psig.

Refuel and Startup modes if reactor pressure (s

c.

Deleted.

d.

The turbine stop valve closure and fast control valve closure scram functions when the reactor thermal power is s 45% (-M1-5 MWt).

g Q' 71F. 7f 3.1/4.1 30 REV 133 8/18/92 i

~

l 4

Bases Continued:

meeting their criterion. To raise ths 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 >6'6" <6'10" above the top of the active fuel. Howev-r, 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 enticipate 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 failurs and therefore adequate margin exists. Specific analyses have generated specific limits which allow this scram to be bypassed below 45% rated thermal power.

In order to ensure the avsilability of this scram above 45% rated thermal e

s s

} N3 power, this scram is only bypassed below 30% turbina first stage pressure. This takes into account the possibility of F6% power being passed dir tly to the condenser through the bypass valves.

,1 1*f Nmljewer u induledby F. Turbine Ston Valve Sc *, The 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 45% rated thermal power, this scram is only bypassed below 30% turbine first stage

~

pressure. This takes into account the possibility of 164 power being passed dir tly to the condenser D

through the bypass valves.

IV

%pu w G. 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 lot valve closure there is no increase in neutron flux.

H. Main Steam Line Low Pressure Initiates Main Steam 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 poser 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 h

REV 70 11/10/04 f

Eases 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 requited 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 there-fore, 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 roactor 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 condensor vacuum trip is bypassed until 600 psig. This bypass also applies to the main steam isolation valves for the same reason. Ref.

Section 7.7.1.2 FSAR.

a betw soy, twtpuee e s J. a. led by b b=" M deje f ** *

• An automatic bypass of the turbine control valve fast closure scram and turbine stop closure scram is effective! hcncver thc turbinc first s cgc prcssurc is Icss thcn 200 cf it, r-ted..luo.

1 [f This insures that reactor therac1 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.

3.1 BASES 38 REV 52

?/9/S!

L 4

TABLE 3.2.1 - Continued 3

Min. No. of Operable Total No. of Instru-or Operating Instru-ment Channels Per ment Channels Per Trip Required Function Trip Settings Trip System System (1,2)

Conditions b.

High Drywell Pressure

$2 psig 2

2 D

(5) t

.3.

Reactor Cleanup System (Group 3) s

'l, (em s**)

g j

a.

Low Reactor Water

>10'0" L..

2 2

E i

' Level S.,

L.,. of 0.,

t aiv. f1 b.

High Drywell Pressure

$2 psig 2

2 E

HPCISteamLines(Grop4) 4.

a.

HPCI High Steam Flow

$150,000 lb/hr 2(4) 2 F

with <60 second

-# 3Yg ej,T, fu.

time ' delay t

M V koke-75 M.

)

i

,b. HPCI High Steam Flow l

$300,000 lb/hr 2(4) 2 F

F f

c.

HPCI Steam Line

$200*F 16(4) 16 F

Area High Temp.

a 5.

RCIC Steam Lines (c,9T) f a.

RCIC High Steam Flow

$45,000 lb/hr 2(4) 2 C

with 5 1 2 e

see time delay

{

b.

RCIC Steam Line Area 200*F 16(4) 16 C

6.

Shutdown Cooling Suoolv Isolation j

f s.

Reactor Pressure 575 psig 2(4) 2 C

s at N the "*d Interlock R

M==

steem 4*M i

~

j 3.2/4.2 50 REV 07 1/20/0' s.

s..

~, -...

.w.

Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems Minimum No. of Oper-Minimus No. of able or Operating Operable or Total No. of Instru-Instrument Channels Operating Trip ment Channels Per Per Tr!p System Required Function Trio Settinz SvSteres (M Trio System (I) Conditions

• 1 "F"I #4#^8* ["

M

(

I D.

HPCI System NSf lettre 7 S TS.

1.

Higli Drywell 52 psig 1

4 4

A.

Pressure (1) 2.

Iow-Iow Reactor 26'6*s6'10" 1

4 4

A.

Water Level

/

C.

Lutomatic Deerss-

/

l surization i

1 1.

Low-low Reaetor 26'6"s6'10" 2

2 2

B.

I Water level and 2.

Auto Blowdown s120 secorals 2

1 1

B.

Timer snd 3.

Low Pressure Core 5100 pMg-2 12(4) 12(4)

B.

Cooling Pumps Dis-2 up 3

_a Charge Pressure s 9 p3 i

Interlock 1

3.2/4.2 53 Rev. 10 7/12/Ob 4

Bases Continued:

3.2 'Ihis trip setting level was chosen to be low enough to prevent spurious operation but high encugh to initiate ECCS operation and primary system isolation so that no melting of the fuel c.1 adding 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 Croup 2 and Group 3 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation ct

/

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 the four RER pumps. Two trip systems are provided in the control logic such that either trip system can permit automatic depressurizatien.

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 l

within their respective trip system.

l l

The pump pressure permissive control logic is designed such that no single ftilure (short or open circuit) will prevent ruto-blowdown or allow auto-blowdown when not rcquired.

l l

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

In additio to monitoring steam flow,

% s., sert.;,,y Ca ok k.> ressere EMS f"? f""'A 0

(,a a

., sa I s,< h }+a4 ao *

• I' LS lha ~ O ' f****f

'

• 4 *')# 1l

?

i

.,.,.,., r ~

• y,

a n~ ~ ~

e,s.h 8,.y,a p ta w s y e-,/.t.*~

I w << d 's i

3.2 BASES In a d.s et,*rge f**

• W'* f**~n*** " E4 Pf'*"

G5 REV

',3 3/2/G1 opp of,c,/;g.

I i

I O

LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMDITS C.

Containment Sprey/ Cooling Syst em C. Containment Spray / Cooling System 1.

Except'as specified in 3.5.C.2x 2 cri '- below, 1.

Demonstrate the RHR Service Water pumps l

both Containment Spray / 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 / spray cooling subsystem consista 2.

Test the valves in accordance eith of the following equipment powered from one Specification 4.15.B.

division:

3.

Demonstrate the operability of the drywell i 4F EUHt/ Service Vater Pumps-spray headers and nozzles with an air test idHeat Exch during each 10 year period.

-p-l ib MKl Pumpe, anger Valves and piping necessary for:

,s Torus Cooling 7p Drywell Spray 2.

Guo RUR Sm..icc "atcr " ump ::y br,-ineper ble far 2C dej5.

2.

^: "1" S reic: ":ter "=p i:. ::ch ;;bsysece-e>y bc inspcrahic for 7 days.

I L -4.

One Containment Spray / Cooling Subsystem ray be a

inoperable for 7 days.

fo<

l 3'4.

If the requirements of 3.5.C.1x 2f 3 and ^

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

1

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

104 3.5/4.5 REV 145 '/3/^0

s s

Gl salp.t m. tis'qy perdset.>, IfDC St31L, heaani E1*eh.t Sefp;n 6)Aal*lry, O

n' esed

• s gy; us te,. es/ef t.%y salp.s,fs.

s o Bases Continued:

by the heat balance method such that 120/125 of full scale on the The IRMs are calibrated 3.1 The highest IRM range is be.ow 20% of rated neutron flux (see Specification 2.3.A.2).

detectors be inserted in the core assures that the heat balance requirement that the I calibration is not inv lidated by the withdrawal of the detector.

ill set the set points within the trip settings specified on Table 3.1.1, Although the operator he various set points can differ appreciably from the value the operator the actual values of The deviations could be caused by inherent instrument error, operator setting is attempting to set.

error, drift of the set point, etc.

Therefore, such deviations have been accounted for in the various transient analyses and the actual trip settings acy vary by the following amountr.

Trio Function Deviation Trio Function Deviation

3. High Flux IRM

+2/125 of scale

• 7. Reactor Low Vater

-6 inches Level

5. High Reactor Pressure

+10 psi

8. Scram Discharge Volume

+& gallon High Level

6. High Drywell Pressure

+1 psi

9. Turbine Condenser Low

-1/2 in. Hg Vacuum

• This indication is reactor coolant temperature sensitive, The calibration is thus made for rated conditions. The level error at low pressures and temperatures is bounded by the safety analysis l

which reflects the weight-of-coolant above the lower tap, and not the indicated level.

i A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting, or a sufficient number of devices have been affected by any means such the automatic function is incapable of operating within the allowable deviation while in a re-f'"

that actor mode in which the specified function must be operable, or the actions specified in 3.1.B are not initiated as specified.

is desirable to determine as soon If an unsafe failure is detected during surveillance testing, it as possible if other failures of a similar type have occurred and whether the particular function in-To meet the require-volved is still operable or capable of meeting the single failure criterion.

is necessary that all instrument channels in one trip system be operable ments of Table 3.1.1, it 39 3.1 BASES REV 131 4/16/92 1A

Bases 3.5/4.5 Continued:

N

[

automatically contrels three selected safety-relief valves although the safety analysis only takes credit It is therefore appropriate to permit one valve to be out-of-service for up to 7-days g

for

.ro valves.

without materially reducing system reliability.

aw d*Wi-9s

• M ~~ f**sMI'N ***"'I*!**

4e,5ala :s.,.<he s.

s

/y,,, p,,,p n, ge. al/,, N 4'/ = A* M5s #4"*/ "a N' >J '

H B. RilR Intertie Line

.tw., m u at u.s r.r.g.

N This An intertie line is provided to connect the RIiR suctio_n line with the two RHR loop return lines.

four-inchlineisequippedwiththreeisolationvalvejs The purpose of this line is to reduce the potential

13:n rith en isciated er idle for water hammer in the recirculation and RHR system 1when required t:

-ent-ferm-ee:Idcr rf de <

v.rtre.a.rtnn.yages ne imletion-valves-are-opened during-a-cooldown-te encur:

der cel c:: nd :::::icted piping-

-RHR-inj eesion-piping r---I f-ene-reelecula tion-loop-i s-is olated--e r-idl e,The RHR loop return line isolation valves ell:r de Operating leep t 2001 de 1: elated-or-idle leep e In the event of an inoperable return line isolation valve, receive a closure signal on 1.PCI initiation.

is a potential for some of the LPCI flow to be diverted to the broken loop during a loss of coolant there been established to periodically cycle the RHR intertie line Surveillance requirements have accident.

In the event of an inoperable RHR loop return line isolation valve, either the inoperable isolation valves.

The RHR valve is closed or the other two isolation valves are closed to prevent diversion of LPCI flow.

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 Spray / Cooling Systems Two containment spray / cooling subsystems of the RHR sys

. ara provided to remove heat energy from the A

containment and control torus and dryvell pressure in the event of a loss of coolant accident.

a RHR heat exchanger, 2 RER containment spray / cooling subsystem consists of 2 RHR service water pumps, Torus Spray is not considered pumps, and valves and piping necessary for Torus Cooling and Drywell Spray. Placing a containment spray / cooling subsystem part of a containment spray / cooling subsystem.

following a loss of coolant accident is a manual operation.

mis loss of The most degraded condition for lor.g term containment heat removal following the coolant accident results from the loss of one diesel generator. Under these ce' nly one RHR pump

' cooling. The

\\

and one RHR service water pump in the redundant division can be used for coi

<mL r l

w vice water and containment temperature and pressure have been analyzed under these condit'-

initial suppression pool temperature are both 90*F.

Acceptable margins '.- W.alaced

.gn conditions han,nple to provide the have been demonstrated. Therefore the containment spray / cooling system i

.l.3.

required heat removal capability. Refer to USAR Sections 5.2.3.3, 6.2.3..

the containment spray / cooling system prevides cooling of the suppression pool During normal plant operation, l

water to maintain temperature within the limits specified in Specification 3.7.A.I.

4 112 3.5/4.5 Rev. 140 7/25/^7-

-^'

t

?

Bases Continued:

3.2 The RBM bypass time delay is set low enough to assure minimum rod movement while ups; ale tripe are

. bypassed.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the Safety Limit (T. S. 2.. l. A).

A downscale indication of an APRM or IRM 1r an indication the instrument has failed or the instrument is not sensitive enough.

In either case the-instrument will not respond to changes in control rod-motion and thus control rod motion is prevented. The downscale rod blocks assure that there will be proper overlap between the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation. The downscale trips are set at 3/125 of full scale.

For effective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief

[

system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.

The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings i

given in the' specification are adequate to assure the above criteria is met.

Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods i

I of maintenance, tecting, or etlibration, and also minimizes the risk of inadvertent operation; 1.e., only one instrument channel out of service.

i t

{'

I Four radiation monitors (two reactor building vent plenum and two refueling floor) are providad which

,f initiate isolation of the reactor building and operation of the scandby gas treatment system following a t

I refueling accident. The monitors measure radioactivity in the reactor building ventilation exhaust and on the refueling floor. -One upscale trip signal or two downscale/ inoperable trip signals, from a pair of monitors performing the same function, will cause the desired action. Trip settings of 100 mR/hr for the reactor building vent plenum monitors and the refueling floor monttors are based upon initiating normal ventilation isolation and standby gas trestment system operation so that none of the activity released during the refueling. accident leaves the reactor buildind via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

-f' '

i recircelstier p" p trip desoription-and-performano: :n:ly i: i: di::::::d in T:pieel Rep;rt s

t Th-

"EDO 25015, Sept rr' r 1975, "E e:1ection-of-Ansielpeted Trencient: L'ithout Scr:n der-thn-Montie:11:

F4 4

]

Muclear Cenerating Plant".

'C ; C ptcat r 15, 1975 Ictter frea Mr L 0 Meycr, MCP, t+.ir 0 L Ziemenn, s

L'S"SC. )

The pump trip is provided to minimize reactor pressure in the highly unlikely avent of a plant trans ne coincident.with the failure of all control rods to scram. The rapid flow reduction rancelabon 3.2 BASES 68 4

REV of sf iu l

U

1 i

Page 3.4 and 4.4 Standby Liquid control Sy'. tem 93 sp k A.

W=:.1 Operation 93 9..-.'m

.$.. 5 ! 5 E.. b!.... 5555.. [.,!f! [ 5 _3 5. 5. E..

b e

3.4 and 4.4 Bases 99 3.5 and 4.5 Core and containment / Spray cooling Systems 101 A.

ECCS 3phas 101 l

B.

RHR Intertie Return Line Isolation Valves 103 e

C.

Contaiament Spray /Coolin5 System 104 D.

RCIC 105 E.

Cold Shutdown and Refuelin6 Requirements 106 F.

Recirculation System 107 3.5/4.5 Bases 1

-M9- / t o i

3,6 and 4.6 Primary System Boundary 121 N

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 C.

Jet Pumps 128 H.

Snubbers 129 3.6 and 4.6 Bases

--44v &

1 3.7 and 4.7 Containment Systems 156 A.

Primary Containment 156 B.

Standby Cas Treatment System 166 C.

Seconc'.ary Containment 169 D.

Primi.ry Containmer.t Isolation Valves 170 E.

Combustible Gas Centrol System W4+ O2 l

3.7 Bases 175 4.7 Bases 183 il REV 120

'/0/01-

f.111 3.8 and 4.8 Radioactive Effluents 192 A.

Liquid Effluents 192 B.

Caseous Effluents 197 C.

Solid Radioactive Waste 198e ; r;> -

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 u) -G42 3.

W Wuey 3rdws sh 3.9 Bases 204 4.9 Bases 1

205 206 3.10 and 4.10 Refueling j

A.

Refueling Interlocks 206 B.

Core Monitoring 207 s

C.

Fuel Storage Pool Water Level 207 D.

Movement of Fuel 207 i

E.

Extended Core and Control Rod Drive Haintenance 208 3.10 and 4.10 Bases Jo9 3.11 and 4.11 Reactor Fuel Assemblies 211 A.

Avarage Planar Linear Heat Generation Rate 211 B.

Linear Heat Generation Rate 212 C.

Minimum Critical Power Ratio 213 3.11 Bases

.216

'I 4.11 Bases zie -495' l

3.12 and 4.12 Sealed Source Contamination 219 A.

Contamination 219 B.

Records 221 3.12 and 4.12 Bases 222 111 REV 'o 11/15/?':

l l

-_-_-________.__.____.___a.___

1

~

Table 3.2.1 Instrumentation "that Initiates Primary Containment Isolation R.tuccions 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

• 1.

Main Steam and Recire Samole Lines (Croup 1) a.

Low Low Reactor 26'-6" s6'10" 2

2 A

Water Level b.

liigh Flow in Main s1404 rated 8

8 A

Steam Line c.

liigh temp. in Main s200* F 8

2 of 4 in each A

of 2 sets Steam Line Tunnel d.

Low Pressure in Main 2825 psig 2

2 B

Steam Line (3)

RilR System, Head Cooling,

^

Drywell, Sump, TIP (Croup

, c....r,>

)

i a.

Low Reactor Water 210'6" ebove-the 2

2 C

Level t p Of th: cetive-Juel-l 49 3.2/4.?

REV 133 ?/18/92-

Bases:

In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents 3.2 which are beyond the operators ability to control, or terminate a single operator error before it This set of specifications provides the limiting conditions of results in serious consequences.

operation for the primary system isolation function, initiation of the emergency core cooling system, The objectives of the Specificatioris are (i) to assure the effectiveness of the protective instrumentation when required, and (ii) to prescribe the trip sett ngs 4 and other safety rclated functions.

i This set of Specifications also provides the limiting required to assure adequate performance.

conditions of operation for the control rod block system.

inscalled in those lines that penetrate the primary containment and must be the radiation dose limits are not exceeded during Isolation valves are isolated during a loss of coolant accident so thatActuation of these valves is initiated by protective instrumentation shown in Such instrumentation must be an accident condition.

Table 3.2.1 which senses the conditions for which isolation is required.The objective is to isolate the primary containment integrity is required.

available whenever primary the guidelines of 10 CFR 100 are not exceeded during an accident.

containment so that The instrumentation which intriates primary system isolation is connected in a dual bus arrangement.

the discussion given in the bases for Specification 3.1 is applicable here.

Thus, The low reactor water level instrumenmtion is set to trip when reactor water level is 1^'0"

'7" on D

the tcp rf th: : ti:: fuel. This trip initiates.

the instrument.rr-1^^t ret-d ther-1 pe er) chez:

isolation valves. Reference ection 7.7.2.2 FSAR.

For trip retting ef 10'5" ni::: :h; :sp cf thm curi.c feci, the valyes will be closed before perforation

\\j closure of group 2, and 3 primary containment line nd therefore thel setting is adequate.

l of the clad occurs even for the maximum break in that i

frf 4/f:nj prov.b s esv<ence that The The low low reactor water level instrumentation is set to trip when reactor water level is 6*6" above a

isolation the top of the active fuel. This trip initiates closure of the Group' 1 Primary containment valves c Reference Section 7.7.2.2 FSAR, and also activates the ECC systems and starts the emergency diesel generator.

7Ls co<rayo Jr & < lour-natr<to I &c fu hp of e/cee (;,1 s! too% fo.oro~ de< fe fx, p enore drq vr ss /f e d< yea. / sq <re f***,

y afheM I~s ~d %*/ys f.

ggy 43: :,733,y 6,,,orcoun h / fa'r.- Iaa 11,.s hag 1

I

I EXHIBIT C REVISION 1 TO LICENSE AMENDMENT REQUEST DATED JULY 26,1996 REVl3ED OPERATING LICENSE AND TECHNICAL SPECIFICATION PAGES NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA LICENSE NO. DPR-22 DOCKET NO. 50-263

Exhibit C Monticello Nuclear Generating Plant Revision 1 to License Amendment Request 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 Change License page 3 A

30*

E, J 4

A 19 E

14 A

38 E

15 A

50*

F, J 6

8 53*

G 56 B

65 G

16 8

104

-H 114 B

39 1

23 C

112 1

24 C

68 1

150 C

li 1

28 D. J iii I

37 D

49 J

64 J

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

/

1

/

3 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 s

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 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission,-

or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1.

Maximum Power Level The licensee is anthorized to operate the facility at steady state reactor core power levels not in excess of 1775 mega-l watts (thermal) cn 2.

Technical Soecifications gz The Technical Specifications contained in Appendix A as u

5 revised through Amendment No. 97 are hereby incorporated in the l4 license.

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

E 4

3.

Physical Protection The licensee shall fully implement and maintain in effect all provirions of the Commission-approved physical security, guard training and qualification, and safegrards 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).

ao*

The plans, which contain Safsguards Information protected under 10 CFR 73.21, are entitled: "Monticello Nuclear Cenerating Plant Physical d*

Security Plan," with revisions submitted through hovember 30, 1987; "Monticello Nuclear Generating Plant Guard Training and Qualification U

Plan," with revisions submitted through February 26, 1986; and E

"Monticello Nuclear Generating Plant Safeguards Contingency Plan,"

I with revisiens submitted through August 20, 1980.

Changes made in j

accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

4.

Prorective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

R.

Rated Neutron Flux - Rated flux is the neutron flux that corresponds to a steady-state power level g

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 Refueline 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 required surveillance testing need not be performed until the next regularly scheduled outage.

V.

S a fe ty Limit - The safety limits are limits below which the maintenance of the cladding and primary systen 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.

i W.

Secondary Containe it Integrity - 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 operabic.

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 i

operation. This determination shall include, where possible, comparison with other independent sensors measuring the same variable.

1.0 4

REV i

h BASES:

L 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 MUt.

- 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.

1 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 r. tapes.. These factors are selected conservatively with respect to their effect.on the applicable tranisent results as determined by'the current analysis model. Conservatism incorporated into the j

transient analysis is documented in Reference 1'

)

q k

i i

i 1

5 t

i

.t t

t i

2.3 BASES

~ 14

.REV-

.q-,

y

Bases Continued:

For analyses of the thermal consequences of the transients, the Operating MCFR Limit (T.S.3.ll.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 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 specifL 4 the actual values of the various setpoints can vary from the specified trip setting by the allowabl

.ieviation.

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 MUt).

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

2.3 BASES 15 REV

2.0 ' SAFETY L'IMITS LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CIADDING TNTECRITY 2.3 FUEL CLADDING INTERCRITY Applicability Applicability 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:

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 >300 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 1.

APRM - The APRM flux scram trip setting minimum critical power ratio (MCPR) less than shall be:

1. 08*, for two recirculation loop operation, or a.

For two recirculation loop operation less than 1.09* for single loop operation, (TLO):

shall constitute violation of the fuel cladding S $ 0.66W + 65.6%

integrity safety limit.

where S - Setting in percent of rated thermal power, rated power being 1775 MUT U - Percent of recirculation drive flow required to produce a core flow of 57.6 x 106 lb/hr b.

For single recirculation loop operation (SLO):

S$ 0.66(W - 5.4) + 65.6%

I MCPR values are for cycle 18 only.

c.

No greater than 120%.

2.1/2.3 6

REV

u-k 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-

. Instrument able or Operating able Bypass Conditions **

Channels per Instrument Channels Required Function Trio Settines Re fuel Startup Run Trio System per Trio' System Conditions *

1. SRM a.

Upscale s5x105 cps 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. IE6 a.

Downscale 23/125 X(b)

X(b) 4 2(Note 1, 4, 6)

A or B or C full scale b.

Upscale s108/125 X

X 4

2(Note 1, 4, 6)

A or B or C full scale

3. APRM a.

Upscale X

3 1(Note 1, 6, 7)

D or E (1) TLO Flow

$ 0.66W + 53.6%

j Biased (Note 2).

(2) SLO Flow

$ 0.66(W - 5.4) + 53.6%

l Biased (Note 2)

(3) High $ 108%

Flow Clamp i

1 b.

Downscale 2 3/125 full scale X

3 1(Note 1, 6, 7)

D or E 3.2/4.2 56

Bases Continued:

Maximum Extended Load Line Limit Analyses-(MELLLA) have been performed _to allow operation at higher l

powers 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 anslyses are

]

discussed in CE NEDC-31849P report (

Reference:

Letter from NSP to NRC dated October 1,1997).

Increased Core Flow (ICF) analyses have been performed to allow operating at flows above 100% for powers 1

equal to or less than 1004 (See Core Operating Limit Report). Tbc supporting analyses are discussed in General Electric NEDC-31778P report (

Reference:

Letter from NSP to FRC dated October 1, 1997).

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

In addition, the evaluation demonscrated 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.

The margin is adequate to accom:nodate 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 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 w'th 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 3 aver rise is very slow.

Generally, the heat flux is in near equilibrium with the fission rate.

In an assuned 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 i

2.3 BASES 16 REV Next Page is 18

"ases 3.5/4.5 continued:

F. Recirculation System The reactor is designed such that thernal 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.

Since the exclusion regi..n may change 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 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 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.

Power distribution controls are established to ensure the reactor is 'oerated within the bounds of the stability analysis. With these controls in place, there is confidence chat 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%

power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the analytically defined exclusion region. The buffer region may be entered only when the power distribution controls are in place.

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specificat.ra 3.5.F.3 were determined by NEDO-24271, June 1980, "Monticello Nuclear Generating Plant Single Loop Operation;"NEDC-30492, April 1984, " Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Generating Plant;" and NEDC-32456P, July 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.

NEXT PAGE IS 121 3.5/4.5 114 REV

-s 4

I Bases Continued:

2.2-

' The normal operating pressure of the : reactor. coolant system is approximately 1010 psi 5-Evaluations.have determined that the 'most severe pressure transient is bounded by. the, closure of '

~

' all MS1vs, 'followed by a reactor scram on high neutron flux (failure of the direct scram associat' d e

with MSIV. position is assumed)~. The USAR discusses the analysis of this event. - The analysis'.

7 results demonstrate the safety / relief valve capacity is capable' of maintaining pressure below the '

' ASME Code limit' of '1104 of. vessel design pressure 1110% X 1250 psig - 1375 psig). ;The safety,limiti

~

j;

~

ensures' that 'the: acceptance limit' of 1375 psig is met during the. design basis event: at the vessel.

-location with the highest pressure.

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2.2 BASES 23-

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REv l

j

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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'eladding integrity safety limit.

The AFRM neutron flux scram and the turbine bypass system also provide protection for these safety limits.

In addition to preventing power operation abeve 1075 psig, the pressure scram backs up the APRM neutron flux scram for steam line isolation type transients.

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 l

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 MWe, followed by no MSIV closure scram but scram l

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 3% 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 conservatively demonstrated to be acceptable at pressures 3% greater than the safety / relief valve setpoint of 1109 psig. HPCI and RCIC pressures required for system operation are limited by the low-iow-Set SRV systen to well below these values.

The operator will set,the reactor coolant high pressure stram trip setting at 1075 psig or lover.

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

+ 1%) 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 the limiting trip setting, or when a sufficient number of devices have been affected by any means.

2.4 BASES 24 REV

t Bases Continued 3.6 and 4.6:

D.

Coolant feakare 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 s mewhat 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.

i i

Two leakage collection eumps are provided inside' primary containment. Identified leakage is piped from die 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 transnitters 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 cor inuously monitored as well as drywell atmospheric temperature and pressure. Systems connected to the reactor coolant systems i

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 -

gpm 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 i

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 e

setpoints is specified in ANSI /ASME OM-1-1981.

Analyses have been performed with the valves assumed to open i

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 i

I e

.a,,

TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Modes in which lunc-Total No. ot liin. Zio, 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 *'

l.

Mode Switch in 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 B

a. High-High

b. Ino erative

c. Hi Flow Clamp < 120 t 5.

i!igh 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.(annulus) X X(f)

X(f) 2 2

A l,

8.

Scram Discharge Volume High Level l

a. East

.s 56 gal.(8)

X(a)

X(f)

X(f) 2 2

,A j

b. West s 56 gal.(8)

X(a)

X(f)

X(f) 2 2

A l

9.

Turbine Condenser l

Iow Vacuum 2 22 in. Hg X(b)

X(b f) X(f) 2 2

A or C l

l 3.1/4.1 K

28 REV I

I l

Bases Continued:

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

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 510%

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 Section 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 providing 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 6.

Delesed 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.

.0.

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

• Recuired Conditions when minimum conditions for operation 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.).

l

• • Allowable Bypass Conditions It is permissible to bypass:

The scram discharge volume liigh Water Level scram function in the refuel mode to allow reactor protection a.

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

3.1/4.1 30 REV

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 26'6" 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 ejection 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 45% 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 Ston Valve Scram The 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 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.

G. 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.

H. Main Steam Line low Pressure Initiates Main Steam Isolation Valve Closure The low pressure isolation of the main steam lines at 825 psig was provided to g've 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 REV l

Bases continued:

start-up and intermediate power ranges.

Ref. Section 7.4.4 FSAR. A source range monitor (SRM) 3.1 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" Ref. Section 7.4.5.2 FSAR.

modes.

In the power range the APRM system provides required protection.

Thus, the IRM system is not required in the "Run" mode.

The AFRM's cover only the power range, the and therefore, the APRM's are IRM's provide adequate coverage in the start-up and Intermediate range, 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 subcritical and depressurized. They are,

.therefore, required to be operational for all modes of reactor operation except in the " Refuel" mode with the reactor subcritical 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 ad condenser vacuum at relatively low reactor power, the main condenser vacuum trip is bypassed until Ref.

600 psig. This bypass also applies to the main steam isolation valves for the same reason.

Section 7.7.1.2 FSAR.

An automatic bypass of the turbine control valve fast closure scram and turbine stop closure scram is l

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.

38 3.1 BASES REV j

i

9 L

v.

6 TABLE 3.2.1 -' Continued Min. No. of Operable Total No. of Instru or Operating Instru-ment Channels Per ment Channels Per Trip"

-Required:

_2 ;

. Function Trio Settines Trio System System (1.2)

Conditions

'~

b.

.High Drywell Pressure s2 psig 2-2 D

.(5)

-3.

Reactor Cleanup System

+

(Group 3) a.

Iow Reactor Water 27"(annulus) 2 2

"E.

Level b.

High Drywell' Pressure s2 psig 2

2

-E-l

( 4'.

HPCI Steam Lines-(Grouo'4) a.

HPCI High Steam Flow s150,000 lb/hr 2(4) 2 F,

with s60 second time delay.

b.

IIPCI High Steam' Flow -

s300,000 lb/hr 2(4) 2

.. F c c.

IIPCI Steam Line' s200* F 16(4) 16 F

' Area liigh Temp.

5.

RCIC Steam Lines (Group 5) a.

RCIC High Steam Flow s45.000 lb/hr 2(4) 2

,C with 5.i 2 see time delay b.

RCIC Steam Line Area s200* F 16(4) 16

'C ~

'6.

Shutdown Cooling-

'f Sucolv Isolation 4

a.

Reactor Pressure 575 psig 2(4)-

2 C

t Interlock at the reactor steam done

'I 3.2/4.2 50 l

REV-a r

?

l

, ~, -.

[

rJ

' 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 Channelsi Operating Trip ment channels.Per.

Per. Trip System Required-Function Trin Settine Systems (3.6)

Trio System

'(3.6)

Conditionr*

B.-

HPCI System 1.

High Drywell-s2 psig 1

.4.

4' A. -

Pressure..- (1) ;

-2.

Iow-Low Reactor.

'26'6"s6'10" l

'4 4

A.

Water Ievel i

C.

Automatic Deores-i surization 1.

Low-Low Reactor 26*6"s6'10" 2

2 2

'B.

Water Level and' 2.

Auto Blowdown-5120 seconds 2

1 1.

B..

Timer and 3.

Low Pressure Core 260 psig 2

12(4) 12(4)

B.

[

Cooling Pumps Dis-sl50 psig Charge Pressure-Interlock.

l l

l3.2/4.2-53 REV L.

~

. - ~,

Bases Continued:

This trip setting level was chosen to be low enough to prevent spurious operation but high enough to 3.2 initiate ECCS operation and primary system isolation so that no melting of the fnel cladding vill occur and so thct post accident cooling can be accomplished and the guidelines of 10 CFR 100 vill not For the complete circumferential break of a 28-inch recirculation line and with the trip be violated.

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 Reference Section 6.2.7 FSAR.

range or spectrum of breaks and meets the above criteria.

The high drywell pressure instrumentation is a back-up to the water level instrumentation and in For the addition to initiating ECCS it causes isolation of Group 2 and Group 3 isolation valves.

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 isolatior. valves include the drywell vent, purge, sump isolation, and RUCU system valves.

Two pressure switches are pravided on the discharge of each of the two core spray pumps and each of the four RHR pumps.

Two trip syst-ss 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 (thort 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 wher. the pumps are not operating.

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

In addition to monitoring steam flow, 65 3.2 BASES REV

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS C.

Containment Spray / Cooling System C. Containment Spray / Cooling System I

1.

Except as specified in 3.5.C.2 below, both 1.

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

temperature is greater than 212*F. A containment / spray coolfng subsystem 2.

Test the valves in accordance with consists of the following equipment powered Specification 4.15.B.

from one division:

3.

Demon

'te the operability of the drywell 1 RHR 3ervice Water Pump spray headers and nozzles with an air test 1 RHR Heat Exchanger during each 10 year period.

1 RHR Pump p

Valves and piping necessary for:

l, Torus Cooling Drywell Spray 2.

One Containment Spray / Cooling Subsystem may be inoperable for 7 uays.

3.

If the requirements of 3.5.C.1 or 2 cannot be net, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

104 3.5/4.5 REV

._________o A

Bases Continued-

'Ihe IRMs are calibrated by the heat balance method such that 120/125 of full scale on the 3.1 The highest IPJi range is below 20% of rated neutron flux (see Specification 2.3.A.2).

requirement that the IRM detectors be inserted in the core assures that the heat balance calibration is not invalidated by the withdrawal of the detector.

Although the operator will set the set points within the trip settings specified on Table 3.1.1 the actual values of the various set point e can differ appreciably from the value the operator CE setpoint methodology provided in NEDC 31336, " General Electric Setpoint i

i is attempting to set.

Methodology." is used. as applicable, in establisning setpoints. The deviations could be caused by etc.

Therefore, such inherent instrument error, operator setting error, drift of the set pc4nt, l

deviations have been accounted for in the various transient analyses and the actual trip settings may i

vary by the following amounts:

Trio Function Deviation Trio Function Deviation

3. High Flux IRM

+2/125 of scale

• 7. Reactor Lov Vater

-6 inches I4 vel

5. High Reactor Pressure

+10 psi

8. Scram Discharge Volume

+1 gallon High level l

6. High Drywell Pressure

+1 psi

9. Turbine Condenser low

-1/2 in. Hg vacuum

• his indication is reactor coolant temperature sensitive. The calibration is thus made for rated conditions. The level error low pressures and temperatures is bounded by the safety analysis at which reflects the weight-of-coolant abo re the lower tap, and not the indicated level.

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting, or a sufficient number of devices have been affected by any means such 7: bat the automatic function is incapable of operating within the allowable deviation while in a re-in which the specified function must be operable, or the actions specified in 3.1.B are actor mode not initiated as specilled.

If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as po:sible if other failures of a similar type have occurred and whether the par;icular function in-To meet the require-volved is still operable or capable of meeting the single failure criterion.

ments of Table 3.1.1, it is necessary that all instrument channels in one trip system be operable 39 3.1 BASES REV

W Bases 3.5/4.5 Continuedi automatically controls three selected safety-relief valves although the safety analysis only takes credit for two valves. It is t'erefore appropriate to permit one valve to be cut-of-service for up to 14 days l

without materially reducing system reliability.

B. RHR Intertie Line An intertie line is provided to connect the RER 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 RER systems. The isolation valves are opened during a cooldown to establish recirculation flow through the RHR suction line and return lines, thereby ensuring a uniforc cooldown of this piping. The RHR loop return line isolation valves receive a closure signal cn LFCI 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 FGGt 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 it.tertie 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.

l C.

Containment Spray /Cooiing Systems l

Two containment spray / cooling subsystems of the RHR system are provided to remove heat energy from the l

containment and control torus and dryvell pressure in the event of a loss of coolant accident. A containment spray / cooling subsystem consists of 2 RHR ser. rice water pumps, a RHR heat exchanger, 2 RER pumps, and valves and piping necessary for Torus Cooling and Dryvell Spray. Torus Spray is not considered part of a containment spray /couting subsystem. Placing a containment spray / cooling subsystem into operation following a loss of coolant accident ;s a manual operation.

The most degraded condition for long term containment heat removal following the design basis loss of coolant accident results from the lose of one diesel generator. Under these conditions, only one FOGt pump and one RHR service w.ter pump in the redundant division can be used for containment spray / cooling. The containment temperature and pressure have been analyzed under these conditions assuming service water and initial suppression pool temperature are both 909F. Acceptable margins to containment design conditions have been demonstrated. Therefore the containment spray / cooling system Is more than apple to provide the required heat removal capability. Refer to USAR Sections 5.2.3.3, 6.2.3.2.3, and 8.4.1.3.

During normal plant operation, the containment spray / cooling system provides cooling of the suppression pool water to maintain temperature within the limits specified in Specification 3.7.A.l.

3.5/4.5 112 Rev.

s

t 1

Eases Continued:

3.2' The RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip sett!ng is less than a factor of 10 above the indicated level. Analysis of the worst case accident results la rod block action before MCFR approaches the Safety Limit (T.S.2.1.A).

A downscale indication of an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod i

motion and thus control rod motion is prevented. The downscale rod blocks' assure that there will be proper overlap between the neutron monitoring systems and thus, that adequate coverap is provided for all ranges of reactor operation. The downscale trips are set at 3/125 of full scale.

For effective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spuricus operation. The trip settings

=

given in the specification are adequate to assure the above criteria is met.

Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Four radiation monitors (ts o reactor building vent plenum ar.d two refueling floor) are provided which initiate isolation of the reactor building and operation of the standby gas treatment system following a refueling accident. The monitors measure radioactivity in the reactor building ventilation exhaust and on the refueling floor. One upscale trip signal or two downscale/ inoperable trip signals, from a pair of monitors performing the same function, will cause the desired action. Trip settings of 100 l

mR/hr for the reactor building vent plenum monitors and the refueling floor monitors are based upon initiating normal ventilation isolation and standby Eas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

The recirculation pump trip is provided to minialze reactor pressure in the highly unlikely event of a l

plant transient coincident with the failure of all control rods to scram. The rapid flow reduction 3.2 BASES 68 AEV

l I

bia 3.4 and 4.4 Standby L'. quid Control System 93 A.

System Operation 93 B.

Boron Solution Requirements 95 C.

96 3.4 and 4.4 Bases 99 3.5 and 4.5 Cors and Containment / Spray cooling System 101 A.

ECCS Systems 101 l

B.

RHR Intertie Return Line Isolation Valves 103 C.

Containment / Spray Cooling System 104 L.

RCIC 105 E.

Cold Shutdown and Refueling Requirements 106 F.

Recirculation System 107 3.5/4.5 Bases 110 l

3.6 and 4.6 Erimary System Boundary 121 A.

Reactor Coolant Heatup and Cooldown 121 B.

Reactor Vessel Tec)erature and Pressure 122 C.

Coolant Chemistry 123 D.

Coolant Leakage 126 E.

Safety / Relief Valves 127 F.

Deleted C.

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 B.

Standby Cas Treatment System 166 C.

Secondary Containment 169 D.

Primary Containment Isolation Valves 170 E.

Combustible cas Control System 172 l

3.7 Bases 175 4.7 Bases 183 11

EAtt 3.8 and 4.8 Radioactive Effluents 192 A.

Liqvid Effluents 192 B.

Caseous Effluents 197 C.

Solid Radioactive Waste 198e D.

Dose from All Uranium Fr.el Cycle Sources 198f 3.8 and 4.8 Eases 198u 3.9 and 4.9 Auxiliary 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 A.

Refueling Interlocks 206 B.

Core Monitoring 207 C.

Puol Storage Poo** Water Level 207 D.

Movement of Puel 207 E.

Extended Core and Control Rod Drive Maintenance 208 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 C.

Minimum Critical Power Ratio 213 3.11 Bases 216 4.11 Bases 218 l

3.12 and 4.12 Sealed Source Contamination 219 A.

Contamination 219 B.

Records 221 3.12 and 4.12 Bases 222 iii REV

Table 3.2.1 Instrumentation That Initiates Primary Containment Isolation Functions Min. No. of Operable Total No. of Instru-or Operating Instru-ment Channels Per ment Channels Per Trip Required Trio Settines Trio System System (1.2)

Conditions

• Function 1.

Main Steam and Recirc Sample Lines (Groun 1) a.

Iow Low Reactor 26'-6" s6'10" 2

2 A

Water Level b.

High Flow in Main s140% rated 8

8 A

Steam Line High temp. in Main s200* F 8

2 uf 4 in each A

c.

of 2 sets Steam Line Tunnel d.

Iow Pressure in Main 2825 psig 2

2 B

Steam Line (3) 2.

RHR System, Head Cooling.

Drywell, Sump. TIP (Group 2) l a.

Low Reactor Water 2 7" (annulus) 2 2

C Ievel 49 3.2/4.2 REV I

l

Bases:

3.2 In addition to reactor protection instrumertation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences oi accidents which are beyond the operators ability to control, or terminate a single operator error before it results in serious consequenus. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, and other safety related functions. The objectives of the Specifications are (1) to assure the effectiveness of the protective instrumentation when required, and (ii) to prescribe the trip settings required to assure adequate periormance. This set of Specifications also provides the limiting conditions of operation for the centrol rod block system.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by protective instrumentation shown in Table 3.2.1 which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The objective 1.s to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arra_.gement.

Thus, the discussion given in the bases for Specification 3.1 is app)' :able here.

The low reactor water level instrumentation is set to trip when reactor water level is 7" on the instrument. This corresponds to a lower water level above the top of active fuel at 100% power due to the pressure drop across the dryer / separator. This has been accounted for in the affected transient analysis. This trip initiates closure of group 2, and 3 primary containment isolation valves.

Reference Section 7.7.2.2 FSAR. The trip setting provides assurance that the valves will be closed before perforation of the clad occurs even for the maximum break in that line and therefore the setting is adequate.

The low low reactor water level instrumentation is set to trip when reactor water level is 6'6" above the top of the active fuel. This trip initiates closure of the Group 1 Primary contaitument isolation valves, Reference Sectior. 7.7.2.2 FSAR, and also activates the ECC systems and starts the emergency diesel generator.

3.2 BASES 64 REV

EXHIBli D REVISION 1 TO LICENSE AMENDMENT REQUEST DATED JULY 26,1996 b

i

SUMMARY

OF PLANT MODIFICATIONS FOR POWER RERATE IMPLEMENTATION 1.

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

l LICENSE No. DPR 22 DOCKET NO. 50 263

=_

l Exhibit D Summary of Plant Modifications for Power Rerate implementation Hardware Changes

- High Pressure Turbine rotor replacement to support power rerate steam flow pain.

Modification to piping or equipment supports for some plant systems due to load changes. Th5 i

involves approximately twelve (12) pipe supports.

Modification to Control Room Emergency Filtration Train (EFT) system to reduce control room l

ventilation finer bypass leakage to establish consistency with control room dose calculation inputs.

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

Modification of the four Molsture 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 sy+...

nhance cooling capacity.

Replacement of both condensate pumps with more efficient high capacity pumps and replacement of one condensate pump motor with a more efficient model (one motor has elready been replaced).

Increase the torque switch setpoint in motor operator for one shutdown cooling suction isolation valve.

Non-Hardwarv Changes Perform adjustments to installed plant instrumentation as necessary to support changes to system process conditions. Examples of affected systems:

Condensate Domineralizer Flow Controllers, Main Steam Line High Flow, -

Main Steam Line High Radiation, Pemss for Turbine Stop Valve Closure and Turbine Control Valve Fast Closure, Main Generator Protective Devices, APRM flow biased neutron flux scram, and Condenser low vacuum scram.

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

D-1 t

t 9

w

.-,.y y

- ~, - - - -

,,,,m

,,,a e

a

<r-e.---

,m

L i

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

valves.

l Re-certify 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 uia EFT system filtration efficiency to 985 Revised the SBGT system filtration efficiency to 85%

implement 1R transformer changes described in Exhibit 1.

f i

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

Erosion / Corrosion program change's to address power rerate effects on inspection frequency, f

t l

l l

\\

i I

l l

l D-2 I

. ~.. -..

~

.. ~..

i I

EXHIBIT F REVISION 1 TO LICENSE AMENDMENT REQUEST DATED JULY 26,1996 MONTICELLO POWER RERATE ENVIRONMENTAL EVALUATION i

NOTE: There were no changes to Exhibit F from Revision 1 to the License Amendment Request. This page was included for continuity. See the original License Amendment Request submitted by letter dated July 26,1996 for a copy of Exhibit F.

4 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA LICENSE NO. DPR-22 DOCKET NO. 50 263 t-m y

.. ~~

v-

-