Text

Proposed Tech Specs to Improve Reliability of PORVs & Availability of Low Temp Overpressure Protection Sys
	ML20077H099
Person / Time
Site:	Prairie Island
Issue date:	06/25/1991
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20077H092	List: ML20077H091; ML20077H096; ML20077H099;
References
	NUDOCS 9107050168
	Download: ML20077H099 (29)
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Exhibit B Frairie Island Nuclear Generating Plant License Amend. ment Request Dated June 25, 1991 Proposed Changes Marked Up On Existing Technical Specification Pages Exhibit B consists of existing Technical Specification pages with the proposed changes written on those.pages.

Existing pages affected by this License Amendment Request are listed below:

TS.3.1 4 TS.3.1 5 TABLE TS,4,1-2A B.3.1 2 B.3.1 3 B.3.1 4 B.3.1 5 B.3.1 6 B.3.1 7 B.3.1 8 B.3.1 9 i

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TS.3.1 4

-hEV-H--10MH&

3.1.A.2.c Pressuriter Power Operated Relief Valver (1) Reactor Cociant Syster. average temperature greater than or equal to 310'F*

shall not (a) -A-reseter-shah r.

be nade-er-eaintained-eri-t+ea-1-nee-ehe-H \\eactor coolant system average temperature *cxceed 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 POVER OPERATION, any one of the following conditions of inoperability may exist for each unit pre?!&d--STARTUP-OFERATION 1 die ++nt4 m+ed-utred action tmtu-GPERAMbHV-i*-rest +eed r If OPERABILITY is not t cannot be completed t restored :: the bleek-velve-eennet b: cle::d within the L ___ _ _.

time specifiedt be in at least 110T $HUTDOWN within the bocause of next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reacto coolant system average excessive temperat ow 310'F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , bd seat leaka

/*

Vithoneormen.PORVs inoperable(,withinonahoureither%" " ge Insert New Paragraphs restore the PORV(s) to.0PERABLE ctatus or close the 3.1.A.2.c.(1),(b).2 and 3 associated block valve (s) thpower[ma'in(ained Y.

Vith one er er--blo.ck valve / inoperable, within one hour Insert New Parigraph either restore the blocV. valvejs1 to OPERABLE status or 2.1.A.2.c.(1) (b).5 OMC

'l"C-f (2) Reactor Coolant System averare teteerature*below 310'r f

With Reactor Coolant System temperature less than 310'F*;

both pressurizer power operated relief valves (PORVs) 7 shall be OPERABLE with the Over Pressure Protection System enabled, the associated block valve open, and the associated l

(Q) backup air supply charged 3 One PORV may be inoperable for 7 days.

If these conditions cannot be met, th: ::::::: ::elen+

syst: was-be depressurize/ and ventpf-t+-the-*te+epher: ::

l

the prc aur4::: :1!+f--tenk withir. S heura -

l (except as specified in 3.1.A.2.c.(2),(a) and greater than or equal to 200'F and 3.1.A.2.c.(2).(b) below) fr the reactor coolant system through at least place its associated PURV in a 3 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

manual contrel.

R., store the block w.ive tw OPERABLE status

Valid until 20 EFPY within the io11owing 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

A e A._ - h 7

4 New Paragraphs 3.1.A.2.c.(1).(b).2, 3 and 5:

2.-With one PORV inoperable due to causes other than excessive seat leak ge, within one hour either restore the PORV to OPERABLE status or close and remove power frotn 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 110T SilVTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant system averags temperature below 310'F* within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 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 wit.hin the next hour.

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i TS.3.1 5 tEV 91 10hh469-3.1.A.3 Reactor Coolant Vent System a.

A reactor shall not be made or maintained critical no shall reactor coolant cystem average temperature exceed 200'F un_ess 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 STAPTUP OPERATION is discontinued until OPERABILITY is restored.

If any one of these conditions is not restored to an OPERABLE status within 30 days, be in at least HOT l

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 inoperabie, or (3) Tha vent valve to the pressurizer relief tank discharge line inoperable, or (4) The vent valve to the containment atmospheric discharge line-inoperable.

c.

With no Reactor Coolant Vent System path OPERABLE, 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 il0T 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 nours.

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4 New Paragraph 3.1.A.2.c.(2).(b):

(b) With both PORVs inoperable, corrplete 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 .

New Section 3.1.A.2.c.(3):

(3) Egaetor Coolant System averare temocrature below 200'F With Reactor Coolant Systern temperature less than 200* F when the head is on the reactor vessel and the reactor coolant systern 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 valvo 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, depressurite 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) Wit:h both PORVs inoperabic, 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 .

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Tcb10 TS.4.1 2A

-ftEV-06 2 /7-/09-MINIMUM FREOUENCIES FOR EOUIPMENT TESTS TSAR Sect.

1g11 Frecuen!v Refereneg 1.

Control Rod Assemblies Rod drop times All rods during each 7

of full length refueling shutdown or rods following each removal of the reactor vessel head; affected rods following maintenance on er modification to tha control rod drive system which could affect performance of those specific rods

-2.

Control Rod Assemblies Partial move-Every 2 weeks 7

ment of all rods 3.

Pressurizer Safety Set point Per ASME Code,Section XI -

Valves Inservice Testing Program 4

Main Steam Safety Set point Per ASME Code,Section XI '

Valves Inservice Testing Program 5.

Reactor Cavity Vater level Prior to moving fuel assemblies or control rods and at least once every day while the cavity is floode 6.

Pressuriter PORV Functional Quaterly Block Valves p

7.

Pressurizer PORVs Functional Every 18 months 8.

Deleted 9.

Primary System Leakage Evaluate Daily 4

10.

Deleted 11.

Turbine stop valves.

Functional See (1) 10 governor valves, and

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. unless the intercept valves.

(Part of turbine block valve has been overspeed protection.)

closed per Specification 3.1.A.2.c.(1),(b) 2 or

-42.

Ocketed-

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

1 ms (1) Turbine stop valves, governor valves and intercept valves are to be tested at a frequency consistent with the methodology presented in l

WCAP 11525 "Probabilistic Evaluation of Reduction in Turbine Valve test Frequency". and in accordance with the established h'RC acceptance criteria for the probability of a turbine missile ejection incident of 1.0x10'3 per year.

In no case shall the turbine valve test inte rval exceed one year.

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B.3.1 2

-nt" 9; 20/27/09 3.1 REACTOR COOLANT SYSTEM BaMs 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.

The pressurizer is needed te maintain acceptable system pressure during noren1 plant operation, including surges that may resu? t following anticipated transients.

Each of the pressurizer safrty valves is designed to relieve 325,000 lbs 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 control system tenperature and pressure.

If no residual heat were removed by any of the means available, the amount of staam which could be generated 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 temperatures 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).

l The requirement that two groups of pressurizer heaters be OPERABLE l

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 l

reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated 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 valvcs Inl S6/T supplied by the 480 volt safeguards buses.

A e --

The minimum pressurization temperatue (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 th) limits of Figures TS.3.1 1 and TS.3.1 2 being exceeded. Thus the low temperature overpressure

- itigatir.g system, which is designed to prevent pressurizing the RCS above

[]fratoct.os the pressure limits specified in Figures TS.3.1-1 and TS.3.1 2, is enabled a t 310 ' F*.

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

Valid until 20 EFPY

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The OPERABILITY of the PORVs and block valves is determined on the basis of their being capabic of perfortning the following functions:

a.

Manual control of PORVs to control reactor coolant pressure. This is a function that is used for the stearn generator tube rupture accident l

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 e

and ensuring the ability to det-et unidentified reactor coolant pressure boundary leakage.

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_0PERABILITY 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 1

temperature is less than 310'F*,

3 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.

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3.1 FEACTOR COOLANT SYSTEM

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Operational Components (cutinued) g/Uo t, c B

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OPERABILITY of ahverpressure -toi+igat4ng 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.

Theisystem is designed to perform its function in the esent of a single failure and is designed to meet the requirements of IEEE 279.

The backup air supply provides sufficient air to operate the PORVs following a letdown isolation with one charging pump in operation for a period of ten minutes after receipt of the overpressure alarm. These specifica-perform its intended function [he overpressure mR4,aths system will%sv t rech u. c e.

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v The reactor coolant vent system is_pToW ded to e aust noncondensible gases from the reactor coolant system that could inhibit natural t

circulation core cooling. The OPERABILITY of at least one vent path l

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 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 prr. ides protection from the failure of a singic 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 reactor coolant makeup system.

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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 l

July 22, 1977.

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INSERT B:

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The setpoint for the low teroperature overpressure prctectlon system is derived by analysis which trodels the performance of the low temperature overpressure protection system assuming various mass input and heat input transients. The low teitperature overpressure protection system sotpoint 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 prersurizer 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 recet the RCS vent requirements.

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 overpressitre event during low tettperature operation.

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B.3.1 A

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3.1 REACT 0P COO 1M;T SYSTEM Bases continued B.

Tressure/ Temperature 1.init s Appendix G of 10 CFR Part 50, and the ASME Code require that the reactor coolant pressure boundary be designed with sufficient mar & n to insure i

that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner, the prc', ability 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'r reactor vessel (Unit i reactor vessel weld V 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.

Heatun 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 vess ?, the thermal induced compressive stresses tend to alleviate the tensile serceses induced by the internal pressure.

Therefore, a pressure-temperature surve 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 governing 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 heatuo and t us time along the heatup ramp; therefore, a lower bound curve similar to : hat described for the heatup of the inner wall canrot be defined.

For the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup ra*.e of interest must be analyzed on an individual basis, The heatup limit curve is a composite curve prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour.

Cooldown curves During cooldown, the ther:nal gradien i in the reactor vessel wall produce 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 are additive to the pressure induced tensile stresses which are aircady present. Therefore, the controlling location is always the inside wall.

g. 3.1 h(#

-4tEV-41 1EW 3.1 ELACTOR C001MT SYSTM Bases (continued)

The cooldovn limit curves were prepared utilir.ing the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside vall.

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

Criticality Limits Appendix G 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 permissibla temperature 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 vill be maintained during system heatup and pressuriza-tion whenever the reactor vessel is in the nil ductility temperature range.

Heatup to this temperature vill be accomplished by operating the reactor coolant pumpa and by the pressuriter heaters.

The pressurizer heater and associated pouer cables have been sized for continuous operation at full heater power, bSME 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 or. 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 ftacture toughness stress limits and thus prevent brittle fracture of the steam generator shell, fressurizer-Limits Although the pressurizer operates at temperature ranges above those for which there is reason for concern about brittle fracture, operating limits tre provided to assure compatibility of operation with the fatigue analysis performed in accordance with ASME Coes requirements.

Reference 1.

USAR Section 4.2

B.3.1\\

-RE" 01 leh?ffe9-3.1 EACTOR C001E T SYSTEd P.ases continued C.

Reactor Coolant System Leakage Leakage from the reactor coolant system is collected in the containment or by other systems. These systets 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):

l 1.

An iner ased 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 icvel indication.

4 Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the teactor coolant, the radiation monitoring system provioes a sensitive indica-tion of primary system leakage. Radiation monitors which indicate primary system Icakage include the containment air particulate and gas monitors, the area radiation monitors, the condenser air ejector monitor, the, component cooling water monitor, and the steam generator blowdown monitor (Reference 2).

l 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, LOCA, 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 raactor coolant system from the low pressure RER system piping.

References 1.

USAR, Section 6.5 2.

USAR, Section 7.5.1 3.

Testimony by J Rnight in the Prairie Island public hearing on l

January 28, 1975, pp 13 17.

I I

B. 3.1h '

RC" ^1 10/27/0')

3.1 REACTOR C001MT SYSTEM Bases continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure 4

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 vill not exceed an l

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.

l-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 sits parameters of the Prairie Island site, such as SITE BOUNDARY location and meteorological l

conditions, were not considered in this evaluation.

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

-liriited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQU1VA1.ENT I 131, but within the allowable limit shown on Figure TS 3.1 3, accommodates possible iodino spiking phenomenon which may occur following changes in THERMAL POWER.

l 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 sinco the activity levels allowed by Figure TS.3.1 3 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 the atmospheric steam relief valves.

The surveillance requirements in Table TS.4.1 2B provide adequate assurance that excessive specific activity levels in the. primary coolant will 'oe detected in sufficient time to take corrective action.

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-Rt-V-H --tW44&9-3.1 REACTOILCM1 ANT SMIfd Bases continued E.

Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration By maintaining the oxygen, chloride and f'.uoride concentrations in the reactor coolant below the normal steady state operation limits specified, the integrity of the reactor coolant system is tssured under all operating conditions (Reference 1).

l If these steady state limits are exceeded, measur es can be taken to correct the condition during reactor operation, s.g., replacement of ion exchange resin or adjustment o:7 the hydrogen con:entration in the volume control tank (Reference 2).

Because of the time dependent l

nature of any adverse effects from oxyger, chlorics, and fluoride concentrations in excess of the limits, it is unnecessary to shut down immediately since the conditions fer corrective action to restore 1

concentrations uithin the steady state limits has been established-.

If the corrective action has not been effective at the end of the 24-hour period, then the reactor will be bro"ght to the COLD SHUTDOVN condition I

and the corrective action will continue.

The effects of contaminants in the reactor coclant are temperature dependent.

It is consistent, therefore, to permit transient concentra-tions to exist for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for coolcnt temperatures less than 250'F and still previde the assurarce the integrity of the primary coolant system will be maintained.

In order to restore the contaminant concentratiot.s to within specifice-tion limits in the event such limits vote exceeded, mixing of the primary coolant with the reactor coolant pumps may be required.

This will result in a small heatup of short duration and vill not increase the average coolant temperature above 250*F.

Etf1Ltn911 1.

USAR, Section 4.5.2 2.

USAR, Section 10.2.3

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3. 3.1.v -
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-REV-W--l O /4MB 9-3.1 REACTOR COO 1/ST SYSTEM Enses continued F.
Isothermal Tutperature Coefficient (1TC)
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 toderator tenperature coefficient beccnes more negative.
The isothermal temperature coefficient is defined as the reactivity change associeted with a unit change in the inoderator and fuel tempera-tures.
Essentially, the isothermal teniperature coefficient is the sum of the moderator and fuel tertperature coefficients. This coefficient is measured d!.rectly during low power PlWSICS TESTS in g
teroperature coefficient are pcm/'F, where 1pera - 1x10'ge isothermal order to verify analytical prediction.
The units of t 6k/k, For extended optimum fuel burnup it is necessary to either load the reactor with burnable poisons e,r increase the boron concer,tration in the reactor coolant system.
If the latter approach is emphasized, it is possible that a positive i nthermai tertperature coefficient l
could exist at beginning of cyc1tc OOC),
Safety analyses verify the acceptability of the isothermal terepeature coefficient for limits specified in 3.1.F.
Other conditions, s.g., higher power or partial l
rod insertion would cause the isotha.rmal coefficient to have a teore negative value.
These analyses deoetstrate that applicable criteria in the !E Standard Review Plan (NUREC 75/087) are met.
Physics ineasureroents and analyses are conducted during this reload startup test program to (1) verify that the plant will operate within safety analyses assumptions and (2) es.tablish operational procedures to ensure safety analyses asstunptions are roet.
The 3.1.F requirernents are waived durinr, low power PlWSICS TESTS to l
permit measurement of reactor teroperature coefficient and other physics design parameters of interest.
Spacial operating precautions will be taken during these PlWSICS 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 trom a reduction of moderator density.
References:
1.
FSAR Figure 3.2.10
Exhibit 0 Prairie Island Nuclear Generating Plant License Amendment Request Dated June 25, 1991 Revised Technical Specification Pages L
Exhibit C consists of revised pages for the Prairie Island Nuclear Generating Plant Technical Specification with the proposed changes incorporated. The revised pages are listed below:
7 TS,3.1-4 TS 3.1-5 TABLE TS 4,1-2A B,3,1-2 g
B,3,1-3 3.3,1-4 B.3,1-5 B,3,1-6 B,3,1-7 B,3.1-8 B.3,1-9 B,3,1-10 m
N I
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a
-l i*".
TS.3.1-4 i
J 3.1.A.2.c Pressurizer Power Operated Relief Valves (1) Reactor Coolant System average temperature greater than or caual to 310'F*
(a) Reactor coolant system average temperature chall 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 restored 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 110T 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 balow 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 salve 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 Ooolant' System._ average temperature greater than or caual 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 4
/alves (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 suoply d'arEed.
i* Valid until 20 EFPY
. ~..
g, 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, dep*)ssurize and vent-the reactor _ctolant 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 ventin3 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 average temocrature 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 2PERAbLE (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 inoperabic 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 squaro 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 .
3.1.A.3 Reactor Coolant Vent System a.
A recetor 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.
Suring STARTUP OPERATION and POWER OPERATION, any one of the following conditions of-inoperability may exist for each unit provided STARTUP OPERATION is discontinued 2ntil OPERABILITY is rectored.
If any_one of these conditioas is not restored to an OPERABLE status within 30 days, be in at least 110T SIIUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SilUTDOWN i
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 L
head vent path inoperable, or (2) -Both of the parallel vent valves in the' pressurizer vent path _ inoperable, or l_
(3) The vent valve to the pressurizer relief tank discharge line inoperable, or (4) The vent va)ve to the containment atmospheric discharge line inoperable, ia c.
With no Reactor Coolant Vent System path OPERABLE, restore at I
l_
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 SilUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
Table TS.4.1 2A e*-
MINIMUM FREQUENCIES FOR EQUIPMENT TESTS FSAR Sect.
Test Preauency Re fe rence
1. control Rod Assemblies Rod Drop Times All rods during each 7
of full length refueling shutdown or rods following each removal of the reactor vessel head; affected rods following maintenance on or modification to the control r d drive system which could affect performance of those specified rods
2. Control Rod Assemblies Fartial move-Every 2 weeks 7
ment of all rods
3. Pressurizer Safety Set point Per ASME Code,Section XI
.nservice Testing Program
4. MainLSteam Safety Set point Per ASME Code,.Section XI Inservice Testing Program
5. Reactor Cavity-Water Level Prior to moving fuel i
assemblies or control rods and at least once every day while the cavity-is flooded.
6. Pressurizer-PORV-Functional Quarterly, unless the Block-Valves block valve has been closed per Specification 3.1.A.2.c.(1).(b).2 or 3.1.A.2.c.(1),(b).3.
7. Pressurizer PORVs Functional Every 18 months 8.' Deleted
9. Primary System Leakage Evaluate Daily 4
'10. Deleted 11.- Turbine stop valves,
Functional See-(1)
'10-governor valves, and
-intercept valves.
(Part of turbine overspeed-pretection) l (1) Turbine-stop 7:=1vcs,. governor valves and intercept valves are to be tested at a frequency consistent with.the methodology presented in WCAP-11525 j.
"Probabilistic Evaluation of Reduction in Turbine Valve test Frequency", and in accordance with the established NRC acceptance criteria'for the probability of'a turbine missile ejection incident of 1.0x10 5 per year.
In no case shall the turbine valve test interval exceed one year.
.. - - -. - ~
~.- _,
i B.3.1-2 1
3.1 REACTOR COOLANT SYSTEM 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.
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 lbs 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 control system temperature and pressure.
If no residual heat were removed by any of the
- means available,- the amount of steam which could be generated at safety
valve relief pressure would be less than half the valves' capacity.
One valve therefore provides adequate defense ascinst over pressurization of the reactor coolant system for reactor coolant temperatures 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 r.atural circulation conditions.
The pressurizer _ power operated relief valv$s (PORVs) operate to relieve reactor coolant system pressure _below the setting of the pressurizer code cafety valves. These relief valves' have ' remotely: operated' 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 or 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 PORVe and block valves is determined on the basis of their being capable of pe forming the following functions:
r 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 :he reactor coolant pressure baundary.
This-is a function that'is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.
p.
y.,
B.3.1-3 3.1 EMQTOR-COOLANT SYSTEM Sases continued A.
Operational-Components (continued)
J.
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 inehes enrures that the RCS will be protected from pressure transients which could exceed the liuits of Apper. dix G to 10 CFR Part 50 when the RCS temperature is less than 310* F*.
The PORV controlcswitches 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 se tpoint. prassure.
The RCS safety valves and normal setpoints on the pressurizer FORV'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 overpressure protection system, which is-designed to prevent pressurizing l the RCS above the pressare limit.: specified in Figures TC.3.1 1 and TS.3.1-2,. is enabled _ at 310* F*.
Abere 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 sotpoint-for the;10w temperature overpressure protection system is derived by analysis which models the performance.oi the low temperature overpressure protection system assuming various mass input and heat input transients..The low temperatv.e overpressure protection nystem setpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 ar.d TS.3.1-2) are revised.
The T 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 ventJrequirements, L
l
Valid until 20 EFPY a
w e e-T er a.+at7 v
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m e
cc 3-4 re-
%e W~
o--*
6
-re--e e<>rie
e-B.3.1-4 l
4 3.1 REACTOR COOLANT SYSTEM Bases continued A.
Operational Components (continued)
The OPERABILITY of the low temperature overpressure protection system.is cetermined 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 j
upstream isolation valve is open end 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 PORVs 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 cystem 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 open, valves in parallel and connect to a common header that discharges either.co 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
-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.
B.3.1-5 l
3.1 REACTOR-COOLANT 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 nonbr5etle manner, the probat'lity of rapidly propagating fracture is minimiaed and the design refit t the uncertainties in determining the effects of irradiation on mater.a1 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 +.te vessel wall.
Heatup_Cartea Daring 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 governing 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 cuter wall of the vessel, These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stressos at the outer wall of the vessel are
' dependent o both the rate of l eatup 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 l
is a composite curve, prepared by determining the most conservative case, l
with either the inside or outside wall controlling, for any heatup rate,,
l to 60*F per hour.
Cooldown Curves 1
i During cooldown, the thermal gradients in the reactor vessel wall produce thern strest-es which vary from tensile at the inner wall to compressive at th uter wall. The thermal induced tensile stresses at the inner wall l
are a:ditive t_
the pressure induced tensile stresses which are already present. Therefore, the controlling location is always the inside wall.
L
~ _.._.
_ - ~
_m_
,.7
.B.3.1 6 l
' 3.1 REACTOR COQJR{T SYSTEM i
HAEu (continued)
The cooldown 'imit curves were prepared utilizing the same type of analysis usedao calculate the heatup curve except that the contro111ni; location 1s always the inside wall.
Limit lines for cooldown rates between those presented may be obtaineri by interpolation.
Gr.lticality Limits Appendix G 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 pressure test.
For Prairie Island the curves were prepared, requiring that criticality must occur above the maximum permissible temperature for the inservice hydrostatit. pressure test.
The criticality limit s,pecified in Figure TS.3.1-1 provides increased
' assurance-that the proper relationship between cactor coolant pressure and temperature will be maintained during system heatup and preneuriza-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 l;
. Test Limits (hydrostatic pressure test) are less_ restrictive than the f
heatup'and cooldown curves to allow for the periodic inservice hydrostatic l'
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 l
factor built into the'heatup and cooldown curves =and because tho test is run at a constant temperature so the thermal stresses in the vessel are minimal.
y Steam Generator Pressure /Termerature 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 ot.' ration with the fatigue analysis performed in accordance with ASME Code requirements.
~
Reference 1.
USAR Section 4.2
m.
.p 4 a
B.3.1-7 l
g 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 f?.ange leakage.
3.
Containment sump water 1cvel indication.
4.
Containment pressure, temperature, ar.d 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 uonitors which indicate primary system leakage include the containment air particulate and gas monitors, the area radiation monitors, the condenser air ejector 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, LOCA, or steam line break accidents
[
(Reference 3).
l Specification 3.1.C.3 specifies actions to be taken in the event of i-failure or excessive leakage of a check valve which isolates the high l
pressure reactor coolant system from the low pressure RHR system i
piping.
References I
1.
USAR, Section 6.5 2.
USAR, Section 7.5.1 3.
Testimony by J Knight in the Prairie Island public hearing on January 28, 1975, pp 13-17
1 i
B.3.1-8 l
3.1 REACTOR C001 ANT SYSTER Bases-continued 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 generntor_ leakage rate of 1,0 gpm.
The values for the limits on specific netivity represent licits based upon a parametric evaluation by the NRG of typical site locations.
These values are conservative in that specific site parameters of the Prairie Island site, such'as SITE BOUNDARY location and meteorological 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.3.1-3 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the SITE b0UNDARY 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 coolent is below the lift pressure of the atmospheric steam _ relief valves.
The surveillance requirements in Table-TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient-time to take corrective action.
i l
l
5' s
l B.3.1 9 l
l 3.1 REACTOR COOLANT SYSTEM Bases continued E.
Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the normal steady-state operation limits specified, the integrity of the reactor coolant system is assured under all operating conditions (Reference 1).
If these steady-state limits are exceeded, measures can be taken to correct the condition during reactor operation, e.g., replacement of ion exchange resin or adjustment of the hydrogen concentration in the volume control tank (Reference 2).
Because of the time dependent nature of any adverse effects from oxygen, chloride, and fluoride concentrations in excess of the limits, it is unnecessary to shut down immediately since the conditions for corrective action to restore concentrations within the steady-state limits has been established.
If the corrective action has not been effective at the end of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, then the reactor will be brought to the COLD SHUTDOWN condition and the corrective action will continue.
The effects of contaminants in the reactor coolant are temperature dependent.
It is consistent, therefore, to permit transient concentra-tions to exist for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for coolant temperatures less than 250* F and still provide the assurance the integrity of the primary coclant system will be maintained.
In order to restore the contaminant concentrations to within specifica-tion limits in the event such limits were exceeded, mixing of t'e n
primary coolant with the reactor coolant pumps may be required. This will result in a small heatup of stort duration and will not increase the everage coolant temperature above 250'F.
References 1.
USAR, Section 4.5.2 i
2.
USAR, Section 10.2.3
&s*>
' *h B.3.1 10
- l-'
3,1 REACTOR COOIANT 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 in the sum of the moderator and fuel temperature coefficients. This coefficient is measured directly during low power PilYSICS TESTS in order to verify analytical prediction. The units of the isothermal temperature coef ficient are pcm/* F, where Ipcm - lx10 5 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 (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 l
within safety analyses assumptions and (2) establish operational procedures to ensure safety analyses assumptions are met.
The 3.1.F_ requirements are waived during low power PilYSICS TESTS to permit measurement of reactor temperature coefficient and other l
physics design parameters of interest.
Special operating precautions will be taken during these PIIYSICS 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.
l l
r
References:
1.
FSAR Figure 3.2.10 I
.n
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--vw, m
u
- - - - - - - - - - - - - -

ML20077H099

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