Text

Amends 106 & 99 to Licenses DPR-42 & DPR-60,respectively, Revising TS Section 3.1.A.2.C & Table TS 4.1-2A & Associated Bases in Response to Generic Ltr 90-06
	ML20045E008
Person / Time
Site:	Prairie Island
Issue date:	06/21/1993
From:	Bill Dean; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20045E009	List: ML20045E008; ML20045E010;
References
	GL-90-06, GL-90-6, NUDOCS 9306300315
	Download: ML20045E008 (30)
	v • d • e

.

l p1L Rf C UNITED STATES -

o ye j

j NUCLEAR REGULATORY COMMISSION.

2 WASHINGTON, D.C. 205EE0001

%...* /

NORTHERN STATES POWER COMPANY DOCKET NO. 50-282-PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 MENDMENT T0 FACILITY OPERATING LICENSE.

Amendment No.106 '

License No. DPR-42 1.

The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated June 25, 1991,; complies with the standards and requirements'of the Atomic _ Energy Act of 1954, as amended (the;Act),

and the Commission's rules and regulations set forth in~10 CFR Chapter I;

B.

The facility will operate.in conformity with the application,~the provisions of 'the Act, and the' rules and regulations of the Commission; C.

There is reasonable assurance (1) that the-activities authorized by this amendment can be conducted without endangering the-health and 3

. safety of the public, and (ii) that such activities: willL be' conducted.

. in compliance with the Commission's regulations; D.

The issuance of this amendment will not be -inimical' to the common defense and security or to the health and safety of the public;' ano E.

The issuance of this amendment.is in'accordance with:10 CFR Part 51'of' the Commission's regulations and-all applicable requirements have been satisfied.

2.

Accordingly,-the license is-amended by changes t'o the Technical Specifica-tions as indicated in the attachtaent to this license amendment, and para-graph.2.C.(2) of Facility Operating License No. DPR-42 is hereby amended i

to read as follows:

l I

1 o

9306300315'930621

'PDR-ADDCK 050002B2

P-PDR

i

-i Technical Specifications-The Technical. Specifications contained in Appendix A, as' revised through Amendment No.-~106,- are hereby incorporated in the license.-

4 The licensee sha11' operate:the facility in.accordance with the Technical Specifications.

3.

This license' amendment is effective as-of the date'of issuance..

FOR THE NUCLEAR REGULATORY COMISSION A

't

~W 1 iam M. Dean, Acting Direct'or Project Directorate 111-1 Division of Reactor' Projects III/IV/V l

Office.of Nuclear Reactor-Regulation'

Attachment:

Changes to the Technical Specifications-Date of Issuance:

June'21,.1993 l

i l

ATTACHMENT TO LICENSE AMENDMENT N0.106 FACILITY OPERATING LICENSE NO. OPR-42 DOCKET NO.'50-282 Revise Appendix A Technical Specifications'by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

i REMOVE INSERT i

TS.3.1-4 TS.3.1-4 TS.3.1-5 TS.3.1-5 Table TS.4.1-2A Table TS.4.1-2A' l

l B.3.1-2 B.3.1-2 i

B.3.1-3 B.3.1-3 i

B.3.1-4 B.3.1-4 B.3.1-5 B.3.1-5 B.3.1-6 B.3.1-6 B.3.1-7 B.3.1-7 B.3.1-8 B.3.1-8 B.3.1-9 B.3.1-9 B.3.1-10

TS.3.104 3.1.A.2.c Pressurizer Power Overated Relief Valves.

(1) Reactor Coolant System average temperature greater than or eaual to 310*F*

(a) Reactor coolant system average temperature shall not exceed 310* F* unless two power operated relief Lvalves (PORVs) and their associated block valves' are.0PERABLE (except as.specified in 3.1. A.2.c(1)(b) below).

(b) During STARTUP OPERATION or POWER'0PERATION, any one of the.

following conditions of inoperability may exist for each unit.

If OPERABILITY is not r,estored within the time specified or the required action cannot be completed, be in at least HOT SHUTDOWN within~the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature.below 310*F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months '.

1.

Withione or both PORVs inoperable because-of excessive seat leakage, within one hour either restore the.PORV(s)1to OPERABLE' status or close'the associated block valve (s) with power' maintained to the block valve (s).

2. With one PORV inoperable due.to-causes'other than excessive seat leakage, within one hour either restore the PORV to OPERABLE status or close and remove power from the associated block valve.. Restore the'PORV to.0PERABLE status within.the following~72 hours,

3. With both PORVs inoperable due to causes other than excessive seat leakage', within one hour either restore.-at least one PORV to OPERABLE status or close and' remove power from the associated block valves and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature below 310* F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4. With one block valve inoperable, within one hour either.

restore:the block valve to OPERABLE. status or. place its associated PORV'in manual' control. Restore the block valve-to CPERABLE status within'the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months '.

5. With both block valves inoperable, within one hour either restore the block ~ valves to OPERABLE status or place the PORVs in manual control' Restore at least one block valve to OPERABLE status within the next hour.

1 (2) Reactor Coolant System average temperature greater than

. or ' coual to 200* F and below 310* F*

l Vith Reactor Coolant System temperature greaterLthan or equal to 200*F and less than 310*F*; both pressurizer power operated relief valves (PORVs) shall be OPERABLE (except as specified in 3.1.A.2.c.(2).(a) and 3.1.A.2.c.(2).(b)'below) with the Over Pressure Protection System enabled. the associated block' valve open, and.the associated backupfair supply charged.

-CValid until 20 EFPY Prairie Island thit 1

-/= rd - t No. W, 91, 106

- Prairie Island thit'2

. Axraat h 73 N, 99

TS.3.1-5 3.1.A.2.c.(2).(a)

One PORV may be inoperable for 7 days.

If these conditions cannot be met, depressurize and vent the reactor coolant system through at least a 3 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(b) With both PORVs inoperable, complete depressurization and venting of the RCS through at least a 3 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(3) Reactor Coolant System averare temperature below 200'F

~

With Reactor Coolant System temperature less than 200*F when the head is on the reactor vessel and the reactor coolant system is not vented through a 3 square inch or larger vent; both Pressurizer power operated relief valves (PORVs) shall be OPERABLE (except as specified in 3.1. A.2.c. (3). (a) and 3.1.A.2.c. (3). (b) below) with the Over Pressure Protection System enabled, the associated block valve open, and the associated backup air supply charged.

(a) One PORV may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If these conditions cannot be met, depressurize and vent the reactor coolant system through at least a 3 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(b) Vith both PORVs inoperable, complete depressurization and venting of the RCS through at least a 3 square' inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3.1.A.3 Eeactor Coolant Vent System a.

A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200'F unless Reactor Coolant Vent System paths from both the reactor vessel head and pressurizer steam space are OPERABLE and closed (except as specified in 3.1.A.3.b and 3.1.A.3.c below).

b.

During STARTUP OPERATION and POWER OPERATION, any one of the following conditions of inoperability may exist for each unit provided STARTUP OPERATION is discontinued until OPERABILITY is restored. 1NE any one of these conditions is not restored to an OPERABLE status within 30 days, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in' COLD SHUTDOWN within the following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s:

)

(1) Both of the parallel vent valves in the reactor vessel head vent path inoperable, or (2) Both of the parallel vent valves in the pressurizer vent path inoperable, or (3)' The vent valve to the pressurizer relief tank discharge line inoperable, or l-L (4) The vent valve to the containment atmospheric. discharge line inoperable.

L c.

With no Reactor Coolant Vent System path OPERALLE, ' restore at least one vent path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Prairie Island thit 1 AruAmt No. 80,' 9f,106 Prairie Island thit 2__

Tar 1@ m d h ]) 4 99

m '

'TabloLTsi4.162A jj MINIMUM FREQUENCIES FOR EQUIPMENT TESTS

. FSAR SAct.

Isa.t Frecuency

Reference:

j 4

1. Control Rod Assemblies Rod Drop' Times

$11 rods'duringiach.

17 L9 of full length refueling shutdown or-rods' followingleach removal'

]

of the reactor-vessel' head;(affacted: rods; following maintenance i

.onTor modification:to:

theLeontrol. rod-drive' 4

systies which;could affect performance ~of-t those'specified: rods-g

-2. Control Rod' Assemblies' Partial noveO

!Every'2l weeks.

-7 j

ment ofLalli~

-}

rods

~

3. Pressurizer Safety' Set pointi Pe'r ASME. Code,iSection XI-:

)

-Valves Inservice' Testing Program-

4. Main Steam Safety Set point Per ASME Code,Section XI i-Valves

. Inservice Testingi rogram j

P

't j

5. Reactor Cavity Water Level'

. Prior'to' moving' fuel:q j

c assemblies'or control'

1 rods'andlat least once' j

every' day lwhile.the, i

cavity.isiflooded.

.l l

6, Pressurizer PORV Functional

-QuarterlykunlessLthel

-I j

Block Valves.

block valve ~has been

~;

j closed perjSpecification?

>j l.

3.1.A.2.c.(1).(b).2 or' i

3.1. A. 2. c. (1).'(b). 3.,

a

~

7. Pressurizer PORVs

-Functional:

Every 18 months 1 j

q

8. Deleted e

9. Primary S'ystem'LAakage-.. Evaluate

Daily.

-4<

n 1

10.' Deleted 4

2.:

11. Turbine.stop. valves,

'Functionalz

.See'(1).

110' i

1

, governor valvesNand-

'l 0

intercept valves.

(Part of: turbine l

f oyerspeed protection)-

~f 4

~

.(1)Turbinestopvalves.governorvalves

and)interAept' valves'are;tobe'teste'dat 1a: frequency consistent withitheLmethodology presented.in'WCAP-11525;

_,l j

1

~

""Probabilistic; Evaluation of:Reduetion.in Turbine Valva Test Frequency".andl W

fin accordanceLwith the established'NRC acceptance.-: criteria'forithe^

1

probability of a. turbinei missile l ejection: incident of 1.0x10*t per..' year., ;In JI t

Eno case'shall?thelturbine'valvi.testeintervaliexceed:one' year.

....c Amen &mmt Nof 75iWh10h

. :.i Prairie Isladd-thit li

?

Wrairie Island thit 2i Q ',s, Amendumt[Noi M.J. 99( u,&,l

^

,3;

l a

B.3.1-2 3.1 REACTOR COOLANT SYSTRi Bases continued A.

Operational Components (continued)

Reactor coolant pump start is restricted to RCS conditions where there is pressurizer level indication or low differential temperature across the SG tubes to reduce the probability of positive pressure surges causing overpressurization.

r The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated. transients.

Each of the pressurizer safety valves is designed to relieve 325,000'1bs per hour of saturated steam at the valve set point.

Below 350'F and 450 psig in the reactor. coolant system, the residual heat removal system can remove decay. heat and thereby contro1' system temperature and pressure.

If no. residual heat were removed by any of the means available, the amount of steam which could be denerated at safety valve relief pressure would be less than half the valves' capacity.

One valve therefore provides adequate defense'against over-pressurization of the reactor coolant system for reactor coolant temparatures less than 350'F.

The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load (Reference 1).

The requirement that two groups of pressurizer. heaters be OPERABLE provides assurance that at least one group will be available during a loss of offsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus.

Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus.

Tests have confirmed the ability of either group to maintain natural circulation conditions.

The pressurizer power operated relief valves (PORVs) operate to relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotelyLoperated block valves to provide a positive shutoff capability should a relief valve-become inoperable.

The PORVs are pneumatic valves operated by instru-ment air.

They fail closed on loss of air or loss of power to their DC solenoid valves.

The PORV block valves are motor. operated' valves supplied by the 480 volt safeguards buses.

The OPERABILITY of the PORVs and block' valves lis determined on the basis of their being capable-of performing the following functions:

a.

Manual control of'PORVs to control reactor coolant pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown.

b.

Maintaining the-integrity of the. reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor. coolant j

pressure boundary leakage.

- Piairie Island thit 1 Aiu s uit No.'91, 106 Frairie Island Unit 2 Aiuauit No. 84, 9)

B.3.1-3 3.1 REACTOR COOIANT SYSTEM Bases continued A.

Operational Components (continued) c.

Manual control' of the block valve to:

(1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item b. above).

d.

Manual control of a block valve to isolate a stuck-open PORV.

The OPERABILITY of two PORVs or an RCS vent opening of at least'3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS-temperature is less than 310*F*.

The PORV control switches are three position switches, Open-Auto-Close. A PORV is placed in manual control by placing its control switch in the closed position.

The minimum pressurization temperature (310*F *) is determined from Figure TS.3.1-1 and is the temperature equivalent to the RCS. safety relief valve setpoint pressure.

The RCS safety _ valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low

~

temperature operational transients.

Inadvertent pressurization of the RCS at temperatures below 310*F*'could result in the, limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded. Thus the low temperature overpressureprotectionsystem,whichisdesignedtopreventpressurizing.l the RCS above the pressure limits specified in Figures 1TS.3.1-1 and TS.3.1-2, is enabled at 310*F*.

Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limits of Figures TS.3.1-1 and TS.3.1-2 from being exceeded.

The setpoint for the. low temperature overpressure protection system is' derived by analysis which models the performance of the low temperature overpressure protection system assuming.various mass input and heat input transients. The low temperature overpressure protection system setpoint:

is updated: whenever the RCS heatup and cooldewn curves (Figures' TS.3.1 and TS.3.1-2) are revised.

The 3 square inch RCS vent opening is based on the 2.956 square inch cross sectional flow area of a pressurizer PORV. Because-the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.

Valid until 20 EFPY Prairie Island thit 1 -

Auauit No 91,106-Prairie Island thit 2 Auduit 10. 84, 99 l

B.3.1-4 l

3.1 REACTOR COOLANT SYSTEM Bases continued A.

Operational Components (continued)

The OPERABILITY of the low temperature overpressure protection system is determined on the basis of their being capable of performing the function to mitigate an overpressure event during low temperature operation.

OPERABILITY of a low temperature overpressure protection system PORV requires that the low pressure set point has been selected (enabled),.the upstream isolation valve is open and the backup air supply is charged.

The low temperature overpressure protection system is designed to perform l its function in the event of a single failure and is designed to meet the requirements of IEEE-279.

The backup air supply provides sufficient air to operate the FORVs following a letdown isolation with one charging pump in operation for a period of ten minutes after receipt of the overpressure alarm. These specifications provide assurance that the low temperature overpressure protection system will perform its intended function.

The reactor coolant vent system is provided to exhaust noncondensible gases from the reactor coolant system that could inhibit natural circulation core cooling.

The OPERABILITY of at least one vent path from both the reactor vessel head and pressurizer steam space ensures the capability exists to perform this function.

The vent path from the reactor vessel head and the vent path from the pressurizer each contain two independently emergency powered, energize to 2

open, valves in parallel-and connect to a common header that discharges either to the containment atmosphere or to the pressurizer relief tank.

The lines to the containment atmosphere and pressurizer. relief tank each contain an independently emergency powered, energize'to open, isolation valve. This redundancy provides protection from the failure of a single vent path valve rendering an entire vent path inoperable. An inoperable vent path valve is defined as a valve which cannot be opened or whose position is unknown.

A flow restriction orifice in each vent path limits the flow from an inadvertent actuation of the' vent system to less than the flow.of the 4

reactor coolant makeup system.

References 1.

USAR, Section 14.4.8.

2.

Testimony by J Knight in the Prairie Island Public Hearing on January 28, 1975.

3.

NSP Letter to.USNRC, " Reactor Vessel Overpressurization", dated July 22, 1977.

Prairie Island thit 1

.imuauitNo.yp,'106-Prairie Island thit 2 Amndment No. 84, 99

j B.3.1-5 l

3.1 REACTOR C00iANT SYSTEM Bases continued B.

Pressure / Temperature Limits Appendix G of 10 CFR Part 50, and the ASME Code require that the reactor coolant pressure boundary be designed with sufficient margin to insure that, when stressed under operating, maintenance, testing,'and postulated accident conditions, the boundary behaves in a nonbrittle manner the probability of rapidly propagating fracture is minimized and'the design reflects the uncertainties in determining the effects of irradiation on material properties. Figures TS.3.1-1 and 2 have been developed (Reference 1) in accordance with these regulations. The curves are based on the properties of the most limiting material in either unit's reactor vessel (Unit i reactor vessel weld W-3) and are effective to 20 EFPY.

The curves have been adjusted for possible errors in the pressure and temperature sensing instruments.

The curves define a region where brittle fracture will not occur and are

~

determined from the material characteristics, irradiation effects, pressure stresses and stresses due.to thermal gradients across the vessel wall.

Heatuo Curves During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall'to tensile at the outer wall. At the inner wall of the vessel,. the thermal induced compressive stresses tend to alleviate the tensile stresses-induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite-heatup rates when the inner wall of the vessel is treated as the governin5 location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses'at the outer wall of the vessel are dependent on both the rate ' of heatup and the - time along the heatup ramp; therefore, a lower. bound curve similar to that described for the heatup of the inner wall cannot be defined. For the cases in which the outer wall of the vessel becomes. the stress controlling. location,. each heatup rate of '

interest must be analyzed on an individual basis. The heatup. limit curve is a composite curve prepared by determining the most conservative case, l

with either the inside or outside wall controlling, for any heatup' rate up l

to 60*F per hour.-

=i Cooldown Curves i

During cooldown, the thermal gradients in the reactor vessel wall produce i

thermal stresses which vary from tensile at the inner wall to compressive at the outer wall. The thermal induced tensile stresses at the inner wall j

are additive to'the pressure induced tensile stresses which.are already q

present.

Therefore, the controlling location is always the inside vall.

Prairie Island Unit 1 AmendmentNo.91,106

. Prairie Island thit 2 Amendnent Jb.. N, 99 3

B.3.1-6 l

3.1 REACTOR C001 ANT SYSTEM Bases (continued)

The cooldown limit curves were prepared utilizing the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside wall.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

Criticality Limits Appendix C of 10 CFR Part 50 requires that for a given pressure, the reactor must not be made critical unless the temperature of the reactor vessel is 40*F above the minimum permissible temperature specified on the heatup curve and above the minimum permissible temperaturn for the inservice hydrostatic pressure test.

For Prairie Island the curves were prepared, requiring that criticality must occur above the maximum permissible temperature for the inservice hydrostatic pressure test.

The criticality limit specified in Figure TS.3.1-1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressuriza-tion whenever the reactor vessel is in the nil ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters.

The pressurizer heater and associated power cables have been sized for continuous operation at full heater power.

ASME Code Section XI Inservice Test Limits The pressure temperature limits for the ASME Code Section XI Inservice Test Limits (hydrostatic pressure test) are less restrictive than the heatup and cooldown curves to allow for the periodic inservice hydrostatic test.

These limits are allowed to bc less restrictive because the hydrostatic test is based on a 1.5 safety factor versus the 2.0 safety factor built into the heatup and cooldown curves and because the test is run at a constant temperature so the thermal stresses in the vessel are minimal.

Steam Generator Pressure / Temperature Limitations The limitations on steam generator pressure and temperature ensure that the pressure induced stress in the steam generators do not exceed the maximum allowable fracture toughness stress limits and thus prevent brittle fracture of the steam generator shell.

Pressurizer Limits Although the pressurizer operates at temperature ranges above those for which there is reason for concern about brittle fracture, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with ASME Code requirements.

Reference 1.

USAR Section 4.2 ML"hMW R?M !b ?

B.3.1-7 l

3.1 REACTOR COOLANT SYSTEM Bases continued C.

Reactor Coolant System Leakage Leakage from the reactor coolant system is collected in the containment or by other systems. These systems are the main steam system, conden-sate and feedwater system and the chemical and volume control system.

Detection of leaks from the reactor coolant system is by one or more of the following (Reference 1):

1.

An increased amount of makeup water required to maintain normal' level in the pressurizer.

2.

A high temperature alarm in the leakoff piping provided to collect-reactor head flange leakage.

3.

Containment sump water level indication.-

4.

Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the reactor coolant, the radiation monitoring system provides a sensitive indica-

{

tion of primary system leakage. Radiation monitors which' indicate

{

primary system leakage include the containment. air particulate and gas

)

monitors, the area radiation monitors, the condenser. air ejector i

monitor, the component cooling water monitor, and the steam generator blowdown monitor (Reference 2).

A leak rate of 1 gpm corresponds to a through wall crack less than 0.6 inches long based on test data. Steam generator tubes having a 0.6-inch long through-wall crack have been shown to resist failure at. pressures resulting from normal operation, IDCA, or steam line break accidents (Reference 3).

~

Specification 3.1.C.3 specifies-actions to be taken in the event of' failure or excessive leakage of a check valve which isolates the high pressure reactor coolant system from the low pressure RHR system i

piping, jLeferences 1.

USAR, Section 6.5 2.

USAR, Section 7.5.1 3.

Testimony by J Knight in the Prairie Island public hearing on I

January 28, 1975, pp 13-17-.

Prairie Island thit 1 Amndmmt No. 91,106

' Pmirie Island thit 2 A u d uit No. #, 93' u__

_ - - _ - _ _ _ _ - _ _ _ _ _ = - -

B.3.1-8

-l' 3.1 REACTOR C001 ANT SYSTEM Bases continued j

D.

Maximum Coolant Activity The limitations on the specific activity of the primary-coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube. rupture accident in-conjunction with an assumed steady state primary-to-secondary steam generator leakage' rate of 1,0 gpm.

The values for the limits on' specific activity' represent limits based' upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Prairie Island site, such as SITE, BOUNDARY. location and metaorological conditions, were not considered in this evaluation.

Specification 3.1.D.2, permitting POWER OPERATION to continue for.

limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within'the allowable limit shown on Figure TS.3.1-3,> accommodates'possible iodine spiking phenomenon which may occur following changes in THERMAL' POWER.

Operation with-specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown-on Figure TS.3.1-3 should be minimized since the activity levels allowed by Figure TS.3.13 increase the 2_ hour thyroid dose at-the SITE BOUNDARY by a factor-of

~

up to 20 following a postulated steam generator? tube rupture.

Reducing RCS temperature to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of tlui atmospheric steam relief valves. The surveillance requirements in-Table TS.4.1-2B provide adequate assurance thatl excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

l i

l 1

1

-)

I l

l Prairie Island Unit I h h t No. fi, 106

' Prairie Island thit Amen &nent No. - lM,. 99. -

.. ~-

,~

3 i~

j

~B.3.1-9

.l-1 1

t 3.1 REACTOR'C00fANT SYSTEM i

i 4

[

Bases continued E.

Maximum Reactor Coolant Oxygen.. Chloride'and Fluoride Concentration

.i a

u 1-By maintaining the oxygen,l chloride:-and fluoride concentrationsoin'the reactor coolant below the normal steady-state operation'11mits specified;.

~

.the integrity of the reactor coolant system!is, assured (under-allu

~

operating conditions (Referencel1).-

d 4

If these steady-state: limits are' exceeded,-measures can beftaken tol correct the conditionLduring reactor. operation',.o;.g N replacement'of l

ion exchange resin.or adjustment of-thel hydrogenLeoncentration.in the l

volume control: tank.(Reference l2). Because"of:the:timeLdependent)

. nature of.any; adverse effectscfrom, oxygen,1chloridet and' fluoride 1.

concentrations in-excess:of the. limits, ityis unnecessary tofshutLdown*

T 3

~

'immediately since the. conditions for' corrective' action;to restore?

. concentrations within-theEsteady-state limits has been established 4

'If*

2 the corrective. action has not been'effectiveLat the end-of>the 24-houri

[

Eperiodi then the reactorLwill?be-broughti to lthe COLD SHUTDOWN cionditioni and the; corrective action will continue Theieffeets of. contaminants in the reactor lcoolanti are Ltssperature i

dependent..It.is' consistent, therefore,Sto-permit" transient;concentra "

i-tions to exist for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for coolant temperaturesiless than 250*Fland still provide thetassurance'theLintegrity.of the primary coolant system; y

will,be maintained.

, 4 1

In order to. restore the. contaminant concentrations to withinsspecifica-tion limits in the event such limits;were; exceeded, mixing of the-

-t i.

primary coolant with the reactor coolant pumps may be required.' This:

E will result in-~a smal1~heatup ofLshort durationLand will?not increase

[

7 the average coolant temperature labove-250*F.

1 i

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- References '.

j q'

II. :USAR; SSction 4'.5;2 i

~~

2. 'USAR,.Section.10.2.3; 1

1 y

Prairie Island thit 1 I Ammdrrint No. 9/[106)

Prairie Island thit.2 J Anendmmt E lWT99 0,

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1 3.1 REACTOR C001 ANT SYSTEM j

i j

18111 continued F.

Isothermal Temperature' Coefficient-(ITC)'

At the beginning of a fuel cycle the moderator _ temperature coeffi-cient has its most positive or.least. negative value.: As.the boron concentration is reduced throughout the fuel cycle, thel moderator.

l temperature coefficient becomes more negative. The. isothermal temperature coefficient is defined as tl.a reactivity change associated with a unit' change in'the moderator and fuel tempera-tures.. Essentially,-the isothermal, temperature coefficient;is.the' sum of the moderator and' fuel' temperature coefficients... This

.j_

s coefficient is measured directly:during low power PHYSICS TESTS in order:to verify analytical prediction. =The units"of:.the isothermals

. temperature coefficient are pen /?F, where 1pcm '.. lx10* Ak/k,

-l Forextendedoptimum. fuel _burnupitisnecessary3toeither[ load.;the-l

~

reactor with burnable pois'ons or increase the boron;concentrationL in the: reactor coolant system.

If the lacter approachiis5 emphasized.

it is possible that a positive 1 isothermal temperature: coefficient-could exist at beginning.of cycle;(500). ' Safety analyses verify thee acceptability.of the1 isothermal-temperature coefficient:for? limits;

specified in 3.1.F...Other conditions,te.'g.,Lhigher: power or partialf rod insertiot would cause the isothermal, coefficient to have a more '

negative value'..-These analyses demonstrate.that' applicable l criteria-in the NRC Standard Review Plan l(NUREG-.75/087) are met.-

Physics measurements.and analyses are conducted _during the reload ~

startup test _ program to (1) verify.that the-plant will. operate within safety analyses'. assumptions'and:(2) establish operational procedures to ensure safety analyses.:assumptionsLare meth-The; 3.1.F requirements are waived during. low power PHYSICS TESTS to -

permit' measurement of-reactor temperature" coefficient and other physics design parameters of.-interest.': = Special-operating-precautions will be taken during'.these-PHYSICS TESTS. 'In' addition, the strong.

negative Doppler coefficient (Reference 1) and thetsmall integratedE ok/k would limit the magnitude of a power; excursion resulting fromia' s

reduction of moderator ~ density.

' References?

'1.

FSAR Figure 3.2.10 f

-i

- A,.;- -- (g 1 Prairie Island thit 1:

>nendmentNo;91,100 Prairie, Island Ibit 2.

Amendnent No N,199.;,

'f X

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.7 UNITED STATES-

[

]

NUCLEAR REGULATORY COMMISSION

's WASHINGTON, D.C. 20666-0001

\\,*...*/

NORTHERN STATES POWER C'OMPANY-DOCKET NO.~50-306:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING ~ LICENSE i

. Amendment ' No.' 99 4

~ License No. DPR-60

~

1.

The Nuclear' Regulatory Commission (the Commission) has found that:

A.

The application for amendment.by Northern States Power' Company;(the licensee) dated June.25, 1991, complies with the~ standards and requirements of the Atomic Energy Act of 1954, Las amended (the Act),

and't~e Commission's= rules and regulations set'forth in'10 CFR Chapter n

I; B.

The facility.will operate in conformity with:the application, the provisions of.the Act, and the ' rules and regulations. of the Commission; i

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of~the public, and (ii) that'such-activities will be. conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will-not be inimical to the common defense and security or to the _ health and. safety 'of the public; and E.

The issuance of this amendment:is in accordance'with-10 CFR Part'51 of j

the Commission's regulations and'all applicable requirements haveLbeen satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-;

tions as indicated in the' attachment to this. license ~ amendment,.and para 4 graph.2.C.(2) of F_acility Operating License No.-DPR-60 is hereby amended-to read as follows:

j 4

1

=

?~,.

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~

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UNITED STATES

[

]

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20056 4 001

$g

,/

...s NORTHERN STATES POWER' COMPANY DOCKET NO. 50-306 EBAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 License No. DPR-60 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application-for amendment by Northern States Power' Company (the licensee) dated June 25,.1991, complies with the standards and requirements 'of the Atomic Energy Act-of 1954,.as. amended (the Act),

and the Commission's rules and regulations set-forth in 10 CFR Chapter I;.

B.

The facility will operate in conformity witithe application, the provisions of the Act, and the ' ules and regulations of the r

Commission; C.

There' is reasonable assurance-(1)' that the activities authorized by this amendment can be conducted without endangering the health and-safety of the public, and (ii) that such' activities will be condu'cted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the= common

~

defense and security or to'the health and safety of'the public; and E..The issuance of this amendment is in~ accordance with-10 CFR Part 51 of the Commission's regulations and all. applicable requirements have been satisfied.

2.

Accordingly,- the. license.is' amended by changes to the Technical Specifica.

tions~as indicated in the attachment to this license amendment,.and para-graph-2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as-follows:

4 s

6 p

ym

Technical Specifications The Technical Specifications contained in Appendix A,.as; revised through Amendment No. 99, are hereby incorporated'inLtheLlicense.-

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

3.

This license amendment is effective as of the-date of' issuance.

l

- FOR THE NUCLEAR REGULATORY COMMISSION' W

iam M Dean, Acting-Director;

- Project Directorate III-1;.

Division of Reactor' Projects III/IV/V!

Office aof Nuclear Reactor: Regulation-

Attachment:

Changes to the Technical i

Specifications Date of Issuance:

June 21, 1993 q;

a i

i F

5

\\l 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 99 FACILITY OPERATING LICENSE NO. DPR DOCKET NO 50-306

- Revise Appendix A Technical Specifications by; removing.the pages ' identified' below and inserting 'the attached, pages.. The revised.pages f are-identified. by amendment number and contain vertical lines indicating the area ofl change.

-BEMOVE INSERT TS;3;1-4

'TS.3.1.TS.3.1 L TS.3.'l Table TS.4.1-2A

-Tabl e / TS.4.1-2A L B.3.1-21 B.311 1

- B.3.1-3

!B.3.13 B.3.1-4~

B.3;1-4' B.3.1-5 B.3.1-5 :

B.3 i B. 3 '.1 -6 ~

B.3.1-7:

z B.3. li7--

B.3.1-B B.3.1-B B.3.1 ~B.3.1-9:

B.3.1-10; i =

1 E

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3 k

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TS.3.1-4 l

3.1.A.2.c Pressurizer Power Overated Relief Valves (1) Reactor Coolant System average temperature greater than or eaual to 310*F*

(a) Reactor coolant system average temperature.shall not exceed 310* F* unless two power operated relief valves (PORVs) and their associated block valves are OPERABLE (except as specified in 3.1.A.2.c(1)(b) below).

(b) During STARTUP OPERATION or' POWER OPERATION, any one of.the following conditions of inoperability may exist for each unit.

If OPERABILITY is not r,estored within the time specified or the required action cannot be completed, be in at least HOT SHUTDOWN within the~next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature below 310* F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1.

With one or both PORVs inoperable because of excessive seat-leakage, within one hour either restore'the PORV(s) to OPERABLE status or close the' associated block valve (s) with power maintained to the block valve (s).

2. With one PORV inoperable due to causes other than excessive seat leakage, within one hour either restore the PORV to OPERABLE status or close and remove power from the associated.

block valve. Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour either restore at least one PORV to OPERABLE status or.close and remove power from the associated block valves and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system average temperature below 310*F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4. With one block valve inoperable, within one hour either restore the block valve to OPERABLE status or place its associated PORV in manual control.

Restore the block valve to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5. With both block valves. inoperable, within one hour either-restore the block valves to OPERABLE status'or place the PORVs in manual control. Restore at least one block valve to OPERABLE status within the next hour.

(2) Reactor Coolant System average temperature greater than or eaual to 200* F and below 310* F*

With Reactor Coolant System temperature greater than or equal to 200* F and less than 310* F*; both pressurizer power operated relief valves (PORVs) shall be' OPERABLE (except as specified in 3.1.A.2.c.(2).(a) and 3.1.A.2.c.(2),(b) below) with the Over Pressure Protection System enabled. the associated block valve open..and'the

-associated backup air supply charged, evalid until 20 EFPY Prairse Island Unit 1

/ = a m t No. N, 91, 106 Prairie Island. Unit 2-kendrent No. 73mW 99

i.4 TS.'3.1-5 3.1.A.2.c.(2),(a)

One PORV may be inoperable for.7 days.

If these conditions cannot-be met, depressurize and vent the reactor coolant system through at-a lesst a 3. square inch vent within tho'next-8 hour's.

(b) With b~oth PORVs inoperable, complete _depressurization and venting of the RCS through'at-least.a'3 square inch: vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(3) Reactor Coolang System averare tennerature below'200'F'

~

t With Reactor Coolant-System temperature less.than 200*F when the. head is-7' on the reactor vessel and the. reactor l coolant: system'is not vented.'.

.i through a 3. square inch-or-larger ^ve'nt; both Pressurizer power operated relief valves (PORVs) shall'be. OPERABLE,(except-as specified.in s

l 3.l.A'.2.c.(3).(a) and 3.1.A.2.c.(3)'.(b)lbelow).with the Over Pressure-

{

Protection System enabled,-the-associated block ~ valve open, and the

[

associsted backup air-supply; charged..

(a) One PORV may be inoperable for.2ie hours.. If the'se. conditions --

cannot be met, depressurize and-vent;the'reactoricoolant system' l

l through at least s'3. square? inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(b) With both PORVs inoperable complete:depressurization'and venting of the RCS through at least'a'3 square inch. vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1 3.1.A.3 Reactor Coolant' Vent System a.

A reactor'shall not'be made or maintained critical nor,shall:

reactor coolantLsystem. average temperature exceed'200'F unless Reactor. Coolant, Vent System paths from both the-reactor vessel head and pres'surizeristeam space-are 0PERABLE:

and closed-(except'as.specified in 3.1.A.3.b and;3.1.A 3.'c.

below),

b.

During STARTUP OPERATION and POWER OPERATION,.any onejof;the=

1 following conditions of inoperability may; exist.for each unit' provided STARTUP OPERATION is^ discontinued until OPERABILITY 4 is restored. 'If any one'of these conditions is not: restored-to an OPERABLE-status within'30 days, be:in'atileast HOT; SHUTDOWN 'within tho'next^6 hours"and in. COLD SHUTDOWN within'the~following 30l hours:

-(1)'Both 6fEthe parallel vent valves.in the' reactor vessell

~

s

' head vent path; inoperable, or-~

i (2)'Both>of:the' parallel, vent valves'in'the press'uriser. vent:

- i path inoperable, or.:

q

'(3)iThelvent valv'e.to'the pressurizer reliefltank discharge'-

^

line inoperable,Lcr

~

(4);The-vent valve tolthe containment atmospheric discharge line; inoperable, c.

With no Reactor; Coolant Vent : System path ' OPERABLE,: restore 'ati

least one vent. path to. OPERABLE. status'withinL72 hours-or.6ei

~

l

, ?in at least HOT. SHUTDOWN'withinEthe next: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD; d

.2

a. SHUTDOWN withinTthe following-30. hours.-

< ~

' Prairie Island thit 13

~

S Amnhenti nom 9/Ol'06i Prairie < Island thit 23

~

2 1._.4 4 N6.173, IW 99?

x...

~

l

+

Tcblo TS.4.1-2A' MINIMUM FREQUENCIES FOR ROUIPMENT TESTS FSAR Sect.

I I331 Frecuency Reference

1. Control Rod Assemblies Rod-Drop Times:

.All rods'duringLeach

7 j

ofifull length refueling shutdown.or?

il rods following each removal of the-reactor vessel head; affected rods.

following maintenance

'on'or' modification'toL the control?rodEdrive.-

system which could'

.]

affect performance,of<

those-specified rods-5 Every 2 Lweeks'.-

7

2. Control' Rod' Assemblies' Partial move' -

ment lof.:all -

~

j rods:

23. Pressurizer Safety Set point-

. Per ASME; Code',E Section -XI: -'--

Valves Inservice?TestingiProgram, l

~

4. Main Steam Safety-Set point-

PerjASME; Code;:SectionlXI'. -

Valves

.In arvice Testing Program M

S. Reactor Cavity.

Water Level-Prioritoinoving fuel' j

assemblies or? control

-rods and at:leastoonce-every day.whilelthe

cavity is.. flooded.

6. Pressurizer PORV Functional-

! Quarterly,t unle'ssithe:

Block Vilves block, valve has been :.

-l closed ~per Specification f

311.A.2;c'4(1)l(b).2 or' 3.1.'A.2;c.(1);(b).3.-

-i oi

..a 7 1 Pressurizer PORVs--

Functional-Every 18 months-

-8.'DeletedL n

9.. Primary System Leakage

Evaluate:

.' Daily; 14

l

10..' Deleted.~

111~ Turbine stop valves.

Functional:

See7(1)[

10 intercept valves..

j

.governorLvalves,1and)

~~

(Part;of. turbine' overspeedl protection)-

[

-(1) Turbine..stop valves,-governor valves and intercept valvesare to be7 tested a't in frequency consistent with the: methodology presented in)WCAP-115251 "Probabilistic Evaluation of. Reduction ~in TurbineLValve Te'st: Frequency",'andj fin accordance~with the established NRC' acceptance criteria;fori he; t

- probability; of al. turbine missilef ejection Cincident of /1.0x10-Nper year. s tin"

~

L P ~

no caseishallLthe: turbine valve test'intervaltexceed onel year,u

! rairie Isladd thit.1;

) brie Island thitE2 1

' Amer &ent No.i 75, N,1106r

%: AnundmentNo."N,79,199) c;1

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

4

.- ~

l.4 1

-4 1

F' B.3.1-2 i-[

3.1 REACTOR COOLANT SYSTEM j.

Agggi continued-j A.

Operational Components (continued)-

I Reactor coolant pump start is restricted to RCS conditions where there is o

pressurizer level indication or lowJdifferential. temperature across the SG-1

[

tubes to reduce the probability of positive pressure surges causing?

{

overpressurization.

.j l

The pressurizer is.needed to maintain la'eceptable_ system pressure.during l

normal plant' operation', including surges.that may' result.following a

anticipated transients. Each of.the

~

l to relieve.325,000.lbs per hour of sa;pressuriser safety _ valves;is designed-i.

turated steam at the. valve. set point'...

[

Below 350*F and 450 psig,in' the reactor coolant system,D the. residual. heat 4

j removal system can-remove decay heat"and'thereby control-system 1

j temperature and pressure.

If no;. residual; heat were removed bylanyLof.the' i

j.

means lavailable,- the amount of. steam which could bel generated;at safety "

j i

valve relief pressure would be less;than half;the.valvest capacity..

j One valve therefore provides~ adequate defense-.against~over-pressurization-ofJ

{

.the reactor coolant system:for reactor coolant: temperatures lless.than 350*F.

The combined capacity lof both' safety: valves is greater! than' the-d i

maximum surge' rate resulting from; complete-lossLof load (Reference 1).

l The requirementithat two groups.of pressurizer heaters bel OPERABLE-'

j-provides assurance.that atLleas.t one group will be;availablelduringia"

+

i loss of offsite power to maintain naturalc'circulationk ? Backup heater.

~

group "A" is normally supplied by one safeguards bus. } Backup heater j

1-group "B",can be manually transferred within minutes to'the redundant j

t safeguards bus. - Tests have confirmed, the ability of. either: group to.

t maintain natural circulation conditions.,

a

+

j.-

The pressurizer power operated relief-valvesx(PORVs); operatei to relieve.

reactor coolant system pressure below the: setting of:the pressurizerf l-code safety valves..These relief valves,have remotelyJoperatedzblock-I valves to provide.a' positive shutoff capability,should'afrelief valve'.

1 become inoperable. The.PORVs are pneumatic valves ~ operated by;instrue 3~

-ment. air. ' They' fail closed 'onl loss of air or loss' of power-to their. DC-.

l solenoid valves.

The PORV block valves are motor. operated l valves

'i supplied by the'480. volt safeguards buses.

t' i

The' OPERABILITY of the PORVs and b1'ockLvalves.is; determined [on:the basis.

of theiribeing capable;of performing theifollowing' functions:

Manual' contro1 of PORVs to' control' reactor coolant pressure. This'is

-a.

a function.thatlisiused for the: steam generator tube ruptureinceident?

l and for plant shutdown.

1

b. 1 Maintaining the-integrity of the1 reactor; coolant; pressure boundary;

[

This 1sia: function.thatiis.related to controlling; identified leakage:

j 2

and' ensuring the ability.to detect: unidentified reactor coolant

~

- pressure boundary leakage.:

'i h

l 1

q'.

~

!. _. L ut No. 9 C 106 1

Piairie Island thit.1 j

Grairie Island thit.21 Ammdent No. 84, 99f.

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"f*-'*'

B.3.1-3 3.1 REACTOR C001 NIT SYSTEM Bases continued A.

Operational Components (continued) c.

Manual control of the block valve to:

(1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item b. above).

d.

Manual control of a block valve to isolate a stuck-open PORV.

The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures.that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS temperature is less than 310*F*.

The PORV control switches are three position switches, Open-Auto-Close.: A PORV is placed in manual control by placing its control switch in the closed position.

The minimum pressurization temperature (310*F *) is determined from Figure TS.3.1-1 and is the temperature equivalent to-the RCS safety relief valve setpoint pressure. The RCS safety valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low temperature operational transients.

Inadvertent pressurization of the RCS at temperatures below 310'F* could result in the limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded.

Thus the low temperature overpressureprotectionsystem,whichisdesignedtopreventpressurizing-l the RCS above the pressure limits specified in Figures TS.3.1-1 and TS.3.1-2, is enabled at 310'F*.

Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limf.ts of Figures TS.3.1-1 and TS.3.1 2 from being exceeded.

The setpoint for the low temperature overpressure protection system is derived by analysis which models the performant:e of the low temperature overpressure protection system assuming various mass input and heat input transients. The low temperature overpressure protection system setpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 and TS.3.1-2) are revised.

The 3 square inch RCS vent opening is based on the.2.956 square inch cross sectional flow area of a pressurizer PORV. Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.

Valid until 20 EFPY Prairie Island thit 1 Aw&mt No 91, 106 Prairie Island thit 2 Anmdment Ms. 84, 99

m _..

go I,

Ll' B.3.1-4 l.

i 4

i, 3.1 REACTOR COOLANT SYSTEM

}

13111 continued 1

0 l

A.

Operational Components (continued)-

]

The OPERABILITY of the low temperature overpressure protection syster;is

(

' determined on'the basis of their being'capablejof' performing the_ function.

1 to mitigate an overpressure event.during low 3 temperature operation.

?'

OPERABILITY of a.Iow temperature overpressure protection l system.PORY:.

requires that the low pressure: set point has,been selected.(enabled),xthe

~

i upstream: isolation valve is'open and the backup air aupplyLis charged.

4 Thelowtemperature'overpressurepro"tectionfsystemis'designedto' perform-l-

~

l its function in thel event of;a single, failure and is; designed to meet theD' j

l requirements of IEEE-279.':The. backup airfsupply provides sufficient airt to operate f the PORVs: following 'a letdown ~ l' solation with'onel charging pump 1:

j in operation for a period of. ten minutes after, receipt of the_ overpressure:

alarm. These specifications provide assurancefthat[the; low temperature-

[

' overpressure protection system will perform,its: intended function.

q 1 --

l The reactor coolantivint system'is providedt tofexhaust noncondensible' f

gases from the reactor coolant system that..could; inhibit-natural 4

l circulation core cooling; The 0PERABILITY of;at leastlone'; vent pathi l

from both the reactor 1 vessel head and pressurizerKateam! space ensures the?

capability exists to-perform this function.

L The vent path from the reactor vesselthead andTche-vent path from:thei l

pressurizer each contain twoiindependently: emergency powered,Lenergizeito j

open;. valves inLparallel,and' connect'to 'a common headerjthatl discharges-l' either to the containment' atmosphere orctoithe pressurizsr relief, tank.

L Tho. lines to ' the containment atmosphere and pres'surizer(relief ~ tank eachh

{

contain an independently. emergency poweredi energize:f to:open,- isolation

[

valve. This redundancy provides protection from thesfailureioffa sing 1'e vent path valve rendering an entire vent'pathiinoperable.; An" inoperable;

.i vent path. valve:is-defined.as a valve which'cannot'beTopened_or whose:

41 q

position is unknown.

i A flow restriction orifice in each; vent path limits the; flow'from an

+

inadvertent actuation of the.' vent-system to lessi han'the flow'of the' t

reactor coolant makeup system.,

y t

4 Ti Re ference s 1, LUSAR,ESection~14.4.8.

2. JTestimony by J, Knight in the PrairisiIsland Public2 Hearing on; January.28.-1975; L

.j e

~3.'

NSF Letter'to USNRC, " Reactor Vessel Overpressurization",:datedt, fJuly:22,L1977.

4 l

~

Prairie Island thit:1.

Amen &nent Noi pf 100 a

. : Prairie: Island thit12 ~

Anerxhent No. 84,: 99

,,S s

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E. L _,

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}

tI B.3.1-5' l

3.1 REACTOR COOLANT SYSTEM Raggi continued i

l B.

Pressure / Temperature Limits l

Appendix G of 10 CFR Part 50,-andjthe ASMElCodeLrequire that the reactor-1 coolant pressure boundary be designed with sufficient. margin to. insure that, when stressed under operating,; maintenance. tasting. and postulated ~.

j accident conditions,.the: boundary behaves ~in a nonbrittle-manner,Lthe-

l j

probability of: rapidly propagating; fracture is minimized and,the design-1 e

reflects the uncertainties'in determining the effects:of. irradiation'oni j

j material properties.= Figures TS.3.1 1 and 2 have been= developed!,..

j.

(Reference 1)!in accordance;with these regulations..The curves are based 1 on the properties of;tho'most' limiting (materialTin:either unit's; reactor-vessel'(Unit 11Ereactor vessel. weld W-3);and are' effective to;20 EFFY.'-The~

j

~

curves have been' adjusted.for:possible Lorrors in the, pressure 'and "

i I

temperature sensing instruments.-

j 1

?

i The curvesidefine's.. region'where brittle fracture will not occur and are?

]

' determined from the material characteristics. irradiation effects,.

1 4

pressure stresses'and'str'essesTdue to th'ernal' gradients.acrossiche vessel.

i-wall.

E 3

l

-Heatuo Curves'

]

4 h

.During heatup, the thermal gradients in~thetreactor.vesse1~ wall produce; thermal stresses which vary from compressive:at thelinner walljto tensile at the outer-wall. At the:iuner wall'of:the. vessel, the thermal' induced, j.

compressive. stresses; tend-to alleviateethe tensile (stressesiinduced.by.the.

l interna 11 pressure. Therefore,:a pressure-temperature curve. based on ~

j. '

steady state conditions:(i.e.,;no; thermal stresses)< represents.a lower..

F bound of-all similar curves.for. finite heatup rates;whenithefinner' wall of.

7 the vessel-:is treated as theigoverning11ocation.

l The heatup analysisiaiso covers.theldeterminationfof pressure-temperature' l-limitations for the case in which:the< outer: wall:of the vessel.becomesi'the; 4

. controlling location. The' thermal gradients established.during heatup j

[

produce. tensile stresses at the outer wall of' the Lvessele Thess' stresses are additive to the)pressureiinduced-tensile. stresses;which are alreadyg j

present..'The-thermal-induced l stresses-at the outer wall;of the vessel are dependent'on both the rate of heatup 'and the timejalong the.heatiup ramp;

' therefore, ' a lower bound curve ~ similarf to: that: described for the heatup of1 the' inner wall'cannot'be; defined.a For the' cases-'in which.the-outerfwall:

~

of the ves'sel;becomes1the< stress = controlling location, each heatup; rate of-

~ interest aust be analyzed"on'an individualibasis. :The'heatup limit curve:

j is'a composite curve prepared by: determining the most conservative. case,,

with either the'.-inside or outsidetwallfcontrolling'L forlany heatup rateLup

.to 60*F per hour.

~

.Cooldown Curves.

u p

During cooldown,.the thermalEgradientsLin~ thei reactior vessel.wal1~ produce E z

. thermal stresses ~which; vary from tensileLat;the inner wall"to compressive' D

~

.'atithe outer. wall. The' thermal induced tensile stres'ses at the" inner-valli

_ (are' additive to;the pressure induced tensile: stresses'which arelalready;

~ 4 lpresentR'Therefore Rthe controlling location is..always thelinside~ wall,c Prairie Island thit 11 Anun&mntNo;91,L106f

~

M T Prairie Island thit 2 f'

g Anendent No( W,1990 m

s

,A a

.u n.m a..

=

4

B.3.1-6 l

3.1 REACTOR COOLANT SYSTEM Bases (continued)

The cooldown limit curves were prepared utilizing'the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside wall.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

Criticelity Limits Appendix C of 10 CFR Part 50 requires that for a given pressure, the reactor must not be made critical unless the temperature of the reactor vessel is 40*F above the minimum permissible temperature specified on the heatup curve and above the minimum permissible temperature for the inservice hydrostatic prassure test.- For Prairie Island the curves were prepared, requiring that criticality must occur above the maximum permissible temperature for the inservice hydrostatic pressure test.

The criticality limit specified in Figure TS.3.1-1 provides increased assurance that the proper relationship _ between reactor coolant pressure and temperature will be maintained during system heatup and pressuriza-tion whenever the reactor vessel is in the nil ductility temperature range.

Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters.

The pressurizer heater and associated power cables have been sized for continuous operation at full heater power.

ASME Code Section XI Inservice Test Limits The pressure temperature limits for the ASME Code Section XI Inservice Test Limits (hydrostatic pressure test) are less restrictive than the heatup and cooldown curves to allow for the periodic inservice. hydrostatic test.

These limits are allowed to be less restrictive because the hydrostatic test is based on a 1.5 safety factor versus the 2.0 safety factor built into the heatup and cooldown curves and because the test is run at a constant temperature so the thermal stresses in the vessel are minimal.

Steam Generator Pressure / Temperature Limitations The limitations on steam generator pressure and temperature ensure that the pressure induced stress in the steam generators.do not exceed the maximum allowable fracture toughness stress limits and thus prevent brittle fracture of the steam generator shell.

Pressurizer Limits Although the pressurizer operates at temperature ranges above'those for which there is reason for concern about brittle fracture, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with ASME Code requirements.

Reference i

1.

USAR Section 4.2 mMgm s+m t m. u

. nnwast No. N,, m 99

t B.3.147

]:

0

_d.

l 3.1 REACTOR COOLANT SYSTEN Bases continued C.

Reactor Coolant System Leakage

]

14akage from the reactor coolant system is collected in tho' containment or by other systems. These systeas'are the. main _ steam: system,.conden-

i sate and feedwater system and the chemical and volume control system.-

m;

.T Deteetion of. leaks from the reactor coolant system is'by one' or more of -

I i

the following'(Reference-1):_

l 1.

An increased amount of, makeup _ water requibed to maintain normal' l

level. in..the ' pressurizer.

2.

A high temperature alarm,in the' leakoff piping provided to collecta.

1 reactor head flange' leakage.

~

- t 3.

Containment' sump water l level; indication.

4. Containment pressure, temperature,and, humidify [ indication.

If thereLis significant radioactive ~ contamination o'f thA reactori coolant,:the radiationjaonitoring system provides a(sensitive? indica->

i

~

tion of primary system 11eakage.; Radiation monitors which indicatej

. primary system leakage _. include the; containment air particulate"and gas.

7 monitors, the area _ radiation ' monitors, the condenser ~ airl ejectort monitor, the component cooling water monitor, and.thejsteamTgenerator

~!

blowdown monitor-(Reference 12).

1 A leak rate of l'gpm corresponds to a throughLwa11 crack 1.ess than:0.6 1

inches:long based on. test data. Steam generatoritubes having a'O.6-inch?

j long through wall crack have;beenlshown to resist failure at pressures; i

resulting from' normal l operation,- IDCA,.or steam line break accidents :

1 (Referencet3).-

q Specification 3.1.C.3' specifies actions:to be?taken in thelevent of.

failure or excessive leakage'.of._a check 1 valve'which isolates the high 4

i pressure reactor coolant system'f on;the? low pressure RHR system.

1 piping.

j 6

+

i s

Re fe renc e s --

.)

'1.

.USAR.1Sectionl6.5~

l

2../USAR,iSection'7.5.1:

3.. Testimony'~oy J-Knight in'the Prairie-Island public-hearing:on,

]j January 28, 1975,1pp.13 17.

=i j

q

/ Prairie' Island thit 1

, AmEnhmt Noi 91,71% ~

ej

Prairie Island thit 2.-

Amendrent No. W, 99 ~

]

8 j

.a u..

= :. ~ x. +

a

~

j -.

j.

p B.3.1-8 l-

? e 3.1 REACTOR COO 1 ANT SYSTEM l-Bases continued i

l i

D.

Maximum Coolant Activity l

l' r

The limitations on'the specific activity'of the primary. coolant ensure j-that the resulting 2. hour-dosesLat the SITE BOUNDARY will not exceed an

{

. appropriately. small fraction of: Part.100 limits following.a steam :

{

j generator tube rupture -accident;in' conjunction' with an assumed steady j-atate' primary-to-secondary steam' generator leakage rate Lof 1.0 spa.

l The values.for thel limits on' specific ~ activity represent limits based l

4 l

upon a parametric evaluation by the NRClof typical: site locations.

These values are conservative;in thatjspecific: site. parameters of'the-Prairie Island site. such as SITE' BOUNDARY: location and meteorologica1'

[

conditions,:were not considered lin this~. evaluation.

,=

1 i-Specification.3.1.D.2~,ipermitting POWER OPERATION'to continue;for i

j.

limited time. periods with the; primary coolant's specific activity.

l

~

l greater than 1.0-microcuries/ gram. DOSE EQUIVALENT I-131, but within the-allowable. limit shown on Figure TS.3.1-3,Laccommodates possible iodine?

. spiking phenomenon ~which may. occur lfollowing. changes.in THERMAL POWER.'

Operation with specific, activity levels; exceeding 1.0'aicrocuries/ gram DOSE EQUIVALENT I-131cbut within-the limits;shown on Figure TS.3.1-3,.

L should be' minimized sinceL the ' activity levels: allowed'byl Figure TS.3.1-3 J 1

increase the-2: hour thyroid dose'at the1 SITE BOUNDARY by;a factor of l

up to 20 following a postulated steam'generatoritube: rupture..

~

Reducing RCS ' temperature j to les's thanf 500*F-preventis. the: release ' of '

~

activity should a, steam generator. tube rupture since.'the' saturation.

.i pressure of the primary coolant is:below the 11ft pressure of:the

~

atmospheric steam; relief valves. The surveillance requirements in' Table TS.4.1-2B; provide adequate 1 assurance.that:excessiveispecific l

activity levels.in the primary coolant will:be detectedLin sufficient-

. time to-take corrective-action.

.iq t

t i

+

'l

~

-I y

~

3 4'

e i,a

y i

p d-r

Prairie Island linit: 1b Amewhent N6. Al; 1061 Prairie Island thit-2 Amendamt No. Eh 99 x-1

7...

s

d ', _

.,s.

.i

-w-

+ +

~..v w,

-4

+

e #-

-- *~

+a"'-"

L

~

s B.3.1-9 ll 3.1 REACTOR COOIANT SYSTEM 1

Raggi continued

~E.

Maximum Reactor Coolant Oxygen,. Chloride and Fluorida concentration By maintaining the oxygen, chloride and fluoride concentrations.in the reactor coolant below the normalisteady-state; operation'11mics'specified,;

the integrity of the. reactor coolant' system is assured under all-operating conditions - (Reference.1).

.If these' steady-state limits.'are.exceededc measures can be taken to correct the' condition during reactor: operation,4e_.g.n replacement of ion exchange; resin or adjustment of thel hydrogen; concentration.in' the:

volume control tank'(Referen.o.2).,,Becausefof the; time dependent" c

nature of any.adverseLeffeets;from, oxygen, chloride,land-fluoridei

~

concentrations. in excess of ? the limits, ?it :is: unnecessary to shut-down immediately sincelthe conditions for= corrective action to restore concentration's within_the steadyistate limits has been established.. If the corrective action has not been effective:at the end.offthe'24-hourf

. period, then the' reactor will bel brought to;the Col.D. SHUTDOWN condition and the corrective-action will.'continun 1.

The' effects of contaminants <in the reactor. coolant!are temperature dependent.. It is consistent, therefore, :to permit' tran' ient. concentra--

s

-tions to exist'for 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months s:for coolant.' temperatures 11ess than'250*F'and-z

.still: provide the assurance-the. integrity of the primary coolant systen will:be maintained.

In order to restore the contaminant concentrations to within;specifica-

~

tion limits in the ' event-such limits were. exceeded, mixing of: the primary coolant with the. reactor coolant pumps lmay'be required. This will result in a small heatup of short; duration and will not' increase-the average coolant-temperature above 250*F.

.)

i Re fe renc e's '~

1.

USAR, Section 4.5.2:

2.

USAR, Section'10.2.3=

f Praihe Islahd Unit 1; l a-t No. 91,106)

Prairie Island Unit 2

l Anerdnent No. Nt99?

_._.__._____.m,+m<g.,

.v4-e

iMi, gp 9:

4 g

9 g,,,,,49y,J.

9 p ye9

,pQ p.yygp up p y.p, gr,g..y J

yam 9 9 9

-,g, g

es i.ey

l*

B.3.1-10 l

3.1 REACTOR COO 1 ANT SYSTEM Bases continued F.

Isothermal Temperature Coefficient (ITC)

At the beginning of a fuel. cycle'the moderator temperature coeffi-cient has its-most positive or least negative value. As the boron concentration is reduced throughout the fuel cycle, the moderator temperature coefficient becomes more negative. The isothermal temperature coefficient is defined as the reactivity change associated with a unit change in the moderator and fuel tempera-tures.

Essentially, the isothermal temperature coefficient is the sum of the moderator and fuel temperature. coefficients. 'This-coefficient is measured directly during low power PHYSICS. TESTS in.

order to verify analytical prediction. The units of the isothermal temperature : coefficient are pcm/*F, where Ipcm - 1x10-8 ak/k, For extended optimum fuel burnup it.is necessary to either load the-reactor with burnable poisons or increase the boron concentration in the reactor coolant system.

If.the latter approach is emphasized, it is possible that a positive isothermal temperature coefficient could exist at beginning of cycle L (BOC).

Safety analyses verify the acceptability of the isothermal temperature coefficient for limits.

specified in 3.1.F.

Other conditions, e.g. ' higher' power or partial rod insertion would cause the-isothermal-coefficient to have a more negative value. These analyses demonstrate that applicable criteria in the NRC Standard Review Plan (NUREG 75/087) are, met.

Physics measurements and analyses are conducted during the reload startup test program to (1) verify that the plant will operate within safety analy;es assumptions and (2) establish-operational procedures to ensure safety analyses assumptions are met.'

The 3.1.F requirements are waived during low power PHYSICS TESTS to i

permit measurement of reactor temperature coefficient-and other' physics design parameters of interest. Special operating precautions will be taken during these PHYSICS TESTS.

In addition,.the. strong negative Doppler coefficient (Reference 1) and the small integrated Ak/k would-limit the' magnitude of a power excursion'resulting from a reduction of moderator density.

References:

-1.

FSAR Figure 3.2.10-Prairie Island Unit 1.

' kodiust k.' 91,106

. Prairie Island thit 2 kodimit k. N, 99

=...

n

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