Text

Revised Proposed Tech Specs,Adding Addl Clarifications & Improvements Identified by Plant Personnel & Senior NRC Resident Inspector
	ML20236K505
Person / Time
Site:	Prairie Island
Issue date:	07/27/1987
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20236K493	List: ML20236K491; ML20236K496; ML20236K505;
References
	NUDOCS 8708070143
	Download: ML20236K505 (18)
	v • d • e

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Exhibit B Prairie Island Nuclear Generating Plant Re"ision 1 to the License Amendment Reauest dated March 17. 1986 Proposed Changes Marked Up on proposed Revision 0 Technical Specification pages l

Mxhibit B consists of the Technical specification pages with the proposed Revision 1 changes marked up on those pages.

If the change affects a pages not revised in Revision 0, then the change is marked up on existing Technical Specification pages. The affect pages are listed below:

TS.3.3-4 TS.3.3-5 TS.3.3-6 TS.3.3-7 (two pages)

TS.3.4-1 TS.3.4-2 Table TS.3.5-4

TS.3.14-3 Table TS.4.1 2B (2 of 2)

TS.4.5-3

B.2.2-1 B.3.3-3 (two pages)

B.3.4-1 B.4.12-2 h

AD P

i TS.3.3-4 I

REV g

/.

d. Manual valves in the above systems that could (if improperly positioned) reduce spray flow below that assumed for accident analysis, shall be blocked and tagged in the proper position.

j During POWER OPERATION, changes in valve position will be under l

direct administrative control.

j d

. I.uts::ti:

1 :c, interisch;, ducts, d _gr;, ::str:1: and pipin;; :::::inted with th: ;b: : : -.p n:::: and ::quir:d for 20:ident renditic : ::: OPEPJ. ELE.

l

/.Thefollowingmotor-operatedvalveconditionsshallexist:

- (1) The Unit ~ 1 operation, containment spray system motor-operated valves MV32096 and MV32097 shall be closed and shall have the motor control. center supply breakers open.>

(2) For Unit 2 operation, containment spray system motor-operated valves MV32108 and MV32109 shall be closed and shall hav, the motor control center supply breakers open.

2.

During STARTUP OPERATION or POWER OPERATION, any one of the fol-lowing conditions of inoperability may exist for each unit provided

~

STARTUP OPERATION is discontinued until OPERABILITY is restored.

f._

If OPERABILITY is not restored within the time specified', 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 . If OPERABILITY is not

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restored within an additional'48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in COLD SHUTDOWN within

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the fo11owing-30 hours.

a. One containment fan cooler unit may be inoperable for 7 days, provided both containment spray pumps are OPERABLE.

b. One containment spray pump may be. inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provided at least two containment fan. cooler units are OPERABLE.

c. Two containment fan cooler units may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provided one containment spray pump is OPERABLE.

i

-d. Two containment spray pumps may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , provided four containment fan cooler units are OPERABLE.

e. The spray additive tank may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

L u.)

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l TS.3.3-5 REV C.

Component Cooling Water System 1.

Single Unit operation

a. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F, unless the following conditions are satisfied. If these conditions cannot be satisfied, except as specified in 3.3.C.1.b below, within one hour initiate the action necessary to place the affected unit in HOT SHUTDOWN, 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(1) The two component cooling pumps assigned to that unit are OPERABLE.

l (2) The.two component cooling heat exchangers assigned to that unit are OPERABLE, l

(3)

.11 ::17::, interl:che, inntrerent tie: 2nd pipir; seeeci-i

^

I' et:d eith th: 25: : :::penentc, 2nd required fer the -

-functioning of th: 07:00; during :::ident ;;ndition:, ::: --

6 RA N -

l

b. During STARTUP OPERATION or POWER OPERATION, any one of the I-following conditions of inoperability may exist provided startup operation is discontinued until OPERABILITY is restored.

If, l

OPERABILITY is not restored within the time specified, 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 . If OPERABILITY is not restored within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in COLD SHUTE 0WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l (1) One of the assigned component cooling pumps may be.

I inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

f (2) One of the assigned component cooling heat exchangers may j

be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

l 4

l 1

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TS.3.3-6 REV' f

2.

Two-Unit Operation r

a. A second reactor shall not be made or maintained' critical nor shall reactor coolant' system average temperature exceed 200'F,

~unless the following conditions are satisfied.' If these cond1-t

-tions cannot be satisfied, except as specified in 3.3.C.2.b below, within one hour initiate the action necessary to place the affected unit in HOT SHUTDOWN, and be in at least HOT i

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

/

"i-following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f' (1) Three component cooling pumps are OPERABLE.

(2) Tour component cooling heat exchangers are OPERABLE.

l

-fa)

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1ver, interle:Fr 2nd piping : rcisted

-with the 2beve'cerpenente, 2nd required fer

-th=

fanctfaa4"g af

k= ryster drdag ecciA=nt mea 4444:ne, :: OPEP>"LE.

b. During STARTUP OPERATIONS or POWER OPERATION either one of the following conditions of inoperability may exist provided STARTUP OPERATION is discontinued until operability is restored. If OPERABILITY is not restored within the time specified, be in at

^

f least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

If OPERABILITY is not restored within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in COLD SHUIDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(1) One of the three component cooling punps may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

I (2) One of the two component cooling heat exchangers associated with each unit may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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.3.

tr" TS.3.3-7

.REV is 3

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D.

Cooling Water System l.

A reactor shall not be made or maintained critical nor shall'reacto; coolant system average temperature exceed 200*F, unless the following conditions are satisfied. '

If these conditions cannot be satisfied, except as specified in 3.3.D.2 beLow, within one hour initiate the

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action necessarto plade the affected' unit in HOT SHUTDOWN, j

y 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

a. Two diesel-driven cooling water pumps 2nd tre :::::

drive :: lin;; unter ; 7: are OPERABIE.

iEterlochs, in:tr;;;nt:risn, pipin;; nnd e /wo No b8' b M

b. All valv:0, fuel til supp,k required fer the functioning cf "

g//g sundr'fvaf#

.cooUtng eter-nycter durin;; 22 id:nt rendi:1 :: nr:

gg opecAa ce

'9PEPJ1LE.

2 2.

During'STARTUP OPERATION or POWER OPERAT ON, the following g, % $4[cyM conditions of inoperability may exist provided STARTUP OPERATION is discontinued until OPERABILITY is restored.

b */ '^'f If these conditions cannot be satisfied, within one hour are d/M44#

initiate the action necessary to place the affected unit in HOT SIPJTDOWN, 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 in COLD SHUTDOWN within the following

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30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , i

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a. One diesel-driven cooling water pump may be inoperable for

7. days (total for both diesel-driven cooling water pumps during any consecutive 30 day period) provided:

l (1) the other diesel-driven pump and its associated diesel generator are OPERABLE.

(2) the engineered safety features associated with the operable i

diesel-driven cooling water pump are OPERABLE; and l

1

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(3) both paths from transmission grid to the plant 4 kV i

p safeguards buses are OPERABLE.

1 j

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s (4) two motor-driven cooling water pumps shall be OPERABLE.

l

b. One of the two required motor-driven cooling water pumps may be inoperable for 7 days provided both diesel-driven cooling water pumps are OPERABLE.

c. One of the two required cooling water headers may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

(1) the diesel-driven pump and the diesel generator associated i

with safety features on the operable header are OPERABLE.

(2) the horizontal motor-driven pump associated with the OPERABLE header and the vertical motor-driven pump are OPERABLE.

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Insert A

-W f

d.

One of the Safeguards Traveling Screens may be inoperable for 6 f

months provided a sluice gate connecting the Emergency and the Cire Water Bay is open (except during periods of testing not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

e.

Both Safeguards Traveling Screens may be inoperable for 7 days

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provided a sluice gate connecting the Emergency and the Circ Water Bay is open.

N A

-6 TS.3.4-1

~

REV l

g, 3.4 STEAM AND POWER CONVERSION SYSTEM Applicability Applies to the operating status of the steam and power conversion system.

Objective To specify minimum conditions of steam-relieving capacity and auxiliary feed-water supp1; necessary to assure that capability of removing decay heat from the reactor, and to limit the concentration of activity that might be released by steam relief to the atmosphere.

She-Specification b

GreeA A.

Safety and Relief Valves A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied. If these conditions cannot be satisfied within one hour initiate the action necessary to place the affected unit in HOT SHUTDOWN, 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 system average temperature below 350*F within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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ltlt /r h sc Y N y

. 5 1.-

Rated relief capacity of ten 4

..m

,,e;es safety valves is

/dId # 0

. 'N available for that reactor, except during testing,

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//40 a rg/ ' //3/ /$

2.

Both steam generttor power-operated relief valves for that je{

l reactor are OPERABLE.

Redwalw e

B.-

Auxiliary feed System 1.

A reactor shall not be ede or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied. 'If these conditions cannot be satisfied, except as specified in 3.4.B.2 below, within one hour initiate the action necessary to place the affected unit in HOT SHUTDOWN, 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 350*F within the'following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

a.

For single unit operation, the turbine-dr1ven pump associated with that reactor plus one motor-driven pump are OPERABLE.

l b.

For two-unit operation, all four auxiliary feedwater pumps are OPERABLE.

l

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TS.3.4-2 REV

'!alves and piping acccciated "ith the 2beve comperente l

f 2.

-ere-OPERABLE cx::pt--that--during-ST.^PTUP OPEPl. TION

-necessery change; ;;y b:--med in :eter :peret-ed-valve.

-posit-ica.

.11 such change ch:11 be under direct ^ 4ad-

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-stret-iv : cent:01.

A minimum of 100,000 gallons of water is available in the condensate storage tanks and ; bachup cuppl-j cf rir:r-

.: ter 10 cv 11:b1: thr ugh th: :: lir.; = t:r ;j:::m.

,e'.

For Unit 1 operation motor operated valves MV32242 and MV32243 shall have valve position monitor lights OPEPABLE and shall be l

locked in the open position by having the motor control center supply breakers manually locked open. For Unit 2, correspond-ing valve conditions shall exist.

E Manual valves in the above systems that could (if one is im-properly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency use.

During POWER OPERATION, changes in valve position will be under l

direct administrative control.

g The condensate supply cross connect valves C-41-1 and C-41-2, l

to the auxiliary feedwater pumps shall be blocked and tagged open. Any changes in position of these valves shall be un direct administrative control.

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2.

During STARTUP OPERATION or POWER OPERATION, any one of he following conditions of inoperability may exist for each unit p vided STARTUP OPERATION is discontinued until OPERABILITY is rest ed. If OPERABI-LITY is not restored within an additien21 'S heurc, place the affected unit (or either unit in the case of a motor driven AFW pump inoperability) 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 350*F within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

A turbine driven AFW pump may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

a.

b.

A motor driven AFW pump may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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TS.3.14-3

, [,,

s' REV D.

Carbon Dioxide System 1.

Except as specified in 3.14.D.3 below, the C0 system protecting g

the relay and cable spreading room area shall e OPERABLE with a l

minimum level of 60% in the CO storage tank.

2 2.

During those periods when the relay and cable spreading room area is normally occupied, automatic initiation of the CO system may 2

be bypsssed. During those periods when the area is normally unoccupied, the CO system shall be capable of automatic initiation 2

unless there are personnel actually in the area.

3.

If specification 3.14.D.1 cannot be met, a continuous fire watch with backup fire suppression equipment shall be stationed in the relay and cable spreading room within one hour. Restore the system to OPERABLE status within 14 days or submit a special l

report to the Commission within 30 days outlining the cause of inoperability and the plans for restoring the system to OPERABLE status.

M Fire Hose Stations y

1.

Whenever equipment protected by hose stations in the following

{,-.

areas is required to be OPERABLE, the hose station (s) protecting that area shall be OPERABLE:

(

a.

Diesel Generator Rocms b.

Safety Related Switchgear Rooms c.

Safety Relatd Areas of Screenhouse d.

Auxiliary Building e.

Control Room f.

Relay & Cable Spreading Room g.

Battery Rooms h.

Auxiliary Feed Pump Room 2.

If Specification 3.14.E.1 cannot be met, within one hour hoses supplied from OPERABLE hose stations shall be made available for l

routing to each area with an inoperable hose station.

Restore the inoperable hose station (s) to OPERABLE status within l

14 days or submit a special report to the Commission within 30 days outlining the cause of the inoperability and the plans and schedule for restoring the stations to OPERABLE status.

(

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Table TS.4.1-2B (Page 2 of 2)

REV '" '""'

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TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Reference

'14. Secondary Coolant Gross Beta-Weekly Gamma act ivity-

15. Secondary coolant Isotopic 1/6 months (5)

Analysis for DOSE EQUIVALENT I-131 concentration

16. Secondary Coolant Chemistry pH 5/ week (6) f ri 5/ week (6)

Sodium 5/ week (6) l (oho NOTES:

1.

Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

2.

To determine activity of corrosion products having a half-life greater than 30 minutes.

3.

See Specification 3.8 for requirements during refueling.

4.

The maximum interval between analyses shall not exceed 5 days.

5.

If activity of the samples is greater than 10% of the limit in Specification g g, p 0.4.A.4, the frequency shall be once per month.

6.

The maximum interval between analyses shall not exceed 3 days.

}

See Specification 4.1.D.

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TS.4.5-3 REV t0 -(7/10/70T

.)

2.

Containment Fan Motors The containment Fan Coil Units shall be run on low motor speed for at le as t 15 minutes at intervals of one month. Motor current shall be measured and compared to the nominal current expected for the test conditions.

i 3

Valves The ref ueling water s torage tank outlet valves shall be a.

tested in p rf rci.4 the punp tcsts.

g e c oro!&M,C4 b.

The accumulator check valves will be checked for operability

-g gghs*

during each refueling shutdown.

d'-

The boric acid tank valves to the saf ety injection system c.

shall be tested at intervals of one month.

d.

The spray chemical additive tank valves shall be cycled by operator action at intervals of one month.

Actuation circuits for cooling water system valves that e.

isolate the non-essential equipment from the system shall be tested monthly.

f.

All motor-operated valves in the SIS, RHR, containment spray, cooling water, and component cooling water system that are designed for operation during the safety injection l

or recirculation phase of emergency core cooling, shall be tested for operability at each refueling shutdown.

g.

The correct position of the throttle valves below shall be verified as follows:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation.

2.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following maintenance on the valve when f

the Safety injection System is required to be operable, and 3.

Periodically at least once per 18 months to the extent not verified in accordance with 1 and 2 above within this time period.

Unit 1 Valves Unit 2 Valves SI-15-6 2SI-15-6 SI-15-7 2SI-15-7 SI-15-8 2SI-15-8 SI-15-9 2SI-15-9 s.

B.2.2-1 REV 4

2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE Bases.

The reactor coolant system (Reference 1) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmos-phere.. In the event of a fuel cladding failure the reactor coolant system is the primary barrier against the release of fission products.

By establishing a system pressure limit, the continued integrity of the reactor coolant system is assured. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III is.110% of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves and fittings under USAS Section B31.1 is 120% of design pressure. Thus, the SAFETY LIMIT of 2735 psig (110% of design pressure) has been established (Reference 2).

The nominal settings of the power-operated relief valves G;^; ec i;;-),

the reactor high pressure trip (200;,;i;) and the safety valves fib 4M-W have been established to assure that the pressure never reaches the reactor coolant system pressure SAFETY LIMIT.

l In addition, the reactor coolant system safety valves (Reference 3) are j

sized to prevent system pressure from exceeding the design pressure by more than 10 percent (2735 psig) in accordance with Section III of the

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ASME Boiler and Pressure Vessel Code, assuming complete loss of load without a direct reactor trip or any.other control, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valves settings.

The acainel setting; of the fiv; sfety valv;; cn ::ch cf th; ;;te i

ete:: lin:: ::: 1075 psig, 1000 psis, 110; y.1, 1120 fois, oud 112^

3 pri;;.

As an assurance of system integrity, the reactor coolant system was hydrotested at 3107 psig prior to initial operation (Reference 4).

l i

i i

References 1.

USAR, Section 4.1 2.

USAR, Section 4.1.3.1 4

3.

USAR, Section 4.4.3.2 4.

USAR, Section 4.1

)

B.3.3-3

,Q REV l

/

3.3 ENGINEERED SAFETY FEATURES Bases continued inoperable during POWER OpERATICN. This is an abnormal operating situation, in that plant operating procedures require that inoperable coni.ainment fan cooler units be repaired as soon as practical. However, because of the difficulty of access to make repairs, it is important on occasion to be able to operate temporarily with only two containment fan cooler units. Two containment fan cooler units can provide adequate cooling for normal operation when the containment fan cooler units are cooled by the chilled water system (Reference 3).

Compensation for this mode of operation is provided by the high degree of redundancy of containment cooling systems during a Design Basis Accident.

One component cooling water pump together with one component cooling heat exchanger can accommodate the heat removal load on one unit, either following a loss-of-coolant accident or during normal plant shutdown. The four pumps of the two-unit facility can be cross-connected as necessary to accommodate temporary outage of the pump.

If, during the post-accident phase, the component cooling water supply were lost, core and containment cooling could be maintained until repairs were effected (Reference 5).

l

(

Normal cooling water supply is from two motor-driven pumps backed up j

by a third motor-driven pump (Reference 6).

In the event of complete l

t V

loss of station power, cooling wate'r is supplied by two diesel-driven pumps which start automatically, each serving half the fan coolers in each reactor. Operation of a single cooling water pump of either type provides sufficient cooling in one unit during the injection and recirculation phases of a postulated loss-of-coolant accident plus sufficient cooling to maintain the second unit in a hot standby condi-

.Esw 0

The component cooling water system and the, cooling water system provide water for cooling components used in normal operation, such as turbine generator components, and rerctor auxiliary components in addition to supplying water for accident functions. These systems are designed to automatically provide two separate redundant paths in each system following an accident. Each redundant path is capable of cooling required components in the unit having the accident and in the oper-ating unit.

There are several manual valves and manually-controlled motor-operated valves in the engineered safety feature systems that could, if one valve is improperly positioned, prevent the required injection of emergency coolant (Reference 7).

These valves are used only when the l

4 reactor is suberitical and there is adequate time for actuation by the reactor operator. To ensure that the manual valve alignment is appro-priate for safety injection during power operation, these valves are tagged and the valve position will be changed only under direct sJ 4

i

Insert B y The Safeguards Traveling Screens and Emergency Cooling Water Supply line f

are designed to provide a supply of screened cooling water in the event I

that an earthquake 1) destroys Dam No. 3 (dropping the water level in the I

normal canal to the screenhouse) and 2) causes the banks bordering the normal canal to the screenhouse to collapse eliminating the river as a source of cooling water. The Safeguards Traveling Screens and Emergency Cooling Water Supply line provide an alternate supply of water to the two diesel-driven and the one vertical motor-driven cooling water pumps.

Their normal supply is from the Circ Water Bay thru one of two sluice gates.

Either one of che two sluice gates or one of the two Safeguards Traveling Screens will adequately supply any of the three cooling water pumps. The Safeguards Traveling Screens are not considered part of the

" engineered safety features associated with the operable diesel-driven cooling water pump" for determination of operability of diesel-driven cooling water pumps.

A

B.3.4-1 REV 3.4 STEAM AND POWER CONVERSION SYSTEMS Bases A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators.

Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.

The ten main steam srfety valves have a total combined rated capability of 7,745,000 lbs/hr. The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten main steam safety valves will be able to relieve the total steam flow if necessary (Reference 1).

In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be assured by availability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the main steam safety valves. One auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedwater pump for each reactor can be made available to the other reactor. L N

The minimum amount of water specified for the condensate storage tank is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shutdown. Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling water system.

The two power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a high energy line rupture outside containment (Reference 2).

Isolation dampers are required in ventilation ducts that penetrate those rooms containing equipment needed for the accident.

The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident

analyses, hdyd '

OP'#-0TIONS> &

O ') STB 9TDP Q y aSca valves h

V p

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O Reference (Ull CM OS Y "C'* $

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{c55 M M (g

2. FSAR, Appendix I g /m, b fg @ Q.

1. USAR, Section 11.9.4 A

B.4.12-2

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REV I

4.12 STEAM GENERATOR TUBE SURVEILLANCE Bases continued plants have demonstrated the capability to reliably detect wastage type defects that,have penetrated 20% of the original 0.050-inch wall

. thickness (Reference 2).

l Whenever the results of any steam gene rator tubing in-service inspection fall into Catgegory C-3, these results will be promptly reported to the Commission prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

Degraded steam generator tubes may be repaired by the installation of sleeves which span the section of degraded steam generator tubing. A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded.

The following sleeve designs have been found acceptable by the NRC Staff:

a.

Westinghouse Mechanical Sleeves (WCAP 10757) f b.

Westinghouse Brazed Sleeves (WCAP-10820) c.

Combustion Engineering Leak Tight Sleeves (CEN-294-XP) l Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval prior to their use in the repair of degraded steam generator tubes. The submittals related to other sleeve designs shall be made at least 90 days prior to use.

References 1.

Testimony of J Knight in the Prairie Icland Public Hearing on 1/28/75 i

i 2.

Testimony of L Frank in the Prairie Island Public Hearing on j

i 1/28/75

)

Exhibit C Prairie Island Nuclear Generating Plant Revision 1 to the License Amendment Recuest dated March 17. 1986 i

Crossreference between Revision 1 Changes and the Exhibit A Safety Evaluations Change:

Applicable Safety Evaluation:

(Pages refer to_new page numbers)

(page numbers refer to Rev 1 of Exhibit A)

1) Deletion of LCOs on Valves, Section 3.3, Item i, page A-28 Interlocks and Piping Section 3.4, Item b, page A-31 (pages TS. 3.3-4, 5,.6, 7,

TS.3.4-2, B.3.4-1)

2) Addition of the Safeguards Item 6, page A-72 Traveling Screens (pages TS.3.3-7, B.3.3-3)

3) Auxiliary Feedwater System Section 3.4, Item d, page A-31

)

Changes Item 3, A-66 (pages TS.3.4-1, 2) l

4) Miscellaneous Changes a) Add Steam Generator Safety Section 3.4, Item e, page A-31 Valve Setpoints to Section 3.4.A (pages TS.3.4-1) and add

" Steam Generator" to the Section title, b) Correction to Table TS.3.5-4.

Section 3.5, page A-33 c) Correction in section labeling Section 3.14, page A-51 (TS.3.14.E).

d) Change pH Control Additive Section 4.1, page A-54 from Ammonia to "pH Control Additive" (Table TS.4.1-2B).

e) Correction of Incorrect Section 4.1, page A-54 Reference to 3.4, A.D in Table 4.1-2B.

l l

f) Change the surveillance on the Section 4.5, page A-55 refueling water storage tank outlet valves in accordance l

with Section 4.2 (page TS.4.5-3).

g) Correct Reference to a Topical Item 3, page A-67 i

Report (Page B.4.12-2) l l

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ML20236K505

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