Text

Proposed STS for Facility
	ML19340A245
Person / Time
Site:	Oconee
Issue date:	06/01/1976
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML19340A246	List: ML19340A245;
References
	NUDOCS 8001310537
	Download: ML19340A245 (14)
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ENCLOSURE 2 4

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i h DEFINITIONS L .I 1 l- i l~ 1 CHAN FUNCTIONAL TEST- .

I 1,11 AC NEL FUNCTIONAL TEST shall be: ;

l a.. Anal channels .the injection of a simulated signal into the L ,

'channe s close to the primary sensor as practicable to verify i

! OPERABIL neluding alarm and/or trip functions.

b. Bistable chann is - the injection of.a simulated signal into the channel sens to verify OPERABILITY including alann and/or !

trip-functions. l l'

CORE ALTERATION .

1 ie 1.12 -CORE ALTERATION shall be the mov t or manipulation of any com-

~ponent within the reactor pressure vessel ith the vessel head removed

.and fuel in the vessel. Suspension of CORE TERATIONS shall not preclude completion of movement of a component to a sa conservative position. l lE ^ ^ -

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1.13 SHUTDOWN MARGIN shall be the' instantaneous ~~aniount o reactivity by which' the reactor is subcritical or would be subcritical from its esent I condjtion assuming:

a. No change in axial power shaping rod position, and b; . All control rod assemblies (safety and regulating) are fuily inserted except for the single rod assembly c? highest reactivity l wo.*.h which is assumed to be fully withdrawr, IDENTIFIED LE 'KAGE_

', 1.14 IDENTIFIC LEAKAGE shall be:-

7 a.- L'eakage(except.CONTROLLEDLEAKAGE)intoclosedsystems,such

-as pump seal or valve packing leaks that are captured and.

3 . conducted to-a sump or collecting tank, or

b. Leakage into the containment. atmosphere from sources that.are

.both specifically located and known either not to interfere.

with the-operation of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or B&W-STS 1-3 ' June 1, 1976

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L l',DEFIM!TIONS l .

c. . Reactor coo'lant system leakage. through a steam generator to the secondary system.

l DENTIFIED LEAKAGE

'1.15 IDENTIFIED (EAKAGE shall be all leakage which _is not IDENTIFIED LE GE or CONTROLLED LEAKAGE.

PRESSURE' BOUNDARY LEAK."f 1.16 PRESSURE BOUNDAR EAKAGE shall be leakage-(except steam generator tube leakage) through a n -isolable fault in a Reactor Coolant System

> component body, pipe wall;or. ssel wall.

I CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that 1 water flow supplied to the reactor coolant pump seals.

QUADRANT POWER TILT 1.18 ~ QUADRANT POWER TILT shall bei the mdimurii differ ce'6etween the power generated in any core quadrani!!(upper or lower co the average power _of all quadrants in that half (upper or - half))and er of the core divided by the average power of all quadrants in tha half

-(upper or lower) of the care and is expressed in percent.

QuadrantPowerTQ$erinanycorequadrant(upperorlower)

L = 100,,x( Average power of all quadrants (upper or lower) ~ II 1

DOSE EQUIVALENT I-131 ;

11.19 ~ DOSE EQUIVALENT-I-131 shall be that concentration of I-131 (uCi/ gram) ~

which alone would produce the same thyroid dose- as the quantity and

, isotopic mixture.of I-131, I-132, I-133, I-134 and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Tabic III of TID-14844, "Calculstion of Distance Factors for Power and Test Reactor Sites."

T - AVERAGE DISINTEGRATION ENERGY' 1.20 T-AVERAGE DISINTEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant-at the time of sampling) of _ the sum of the average beta and.gama energies l

~ B&W-STS 1 June 1,'1976 O.

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-7e- - Ty- ty-gi a s"y y y -des = >=+pg - wFsgiF>* e e- -e 9 Pgr' ayW*

k DEFINITIONS per disintegration (in MeV) for isotopes, other than iodines, with half

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lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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S_ GGERED TEST BASIS

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1.21 STAGGERED TEST BASIS shall consist of:

a, A test schedule for n systems, subsystems, trains or designated n omponents obtained by dividing the specified test interval to n equal subintervals,

b. The esting of one system, subsystem, train or designated compo nts at the beginning of each subinterval.

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i FREQUENCY NOTATION J 1.22 The' FREQUENCY NOTATION specified fc' the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

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AXIAL POWER IMBALANCE ,. .

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.g 1.23 AXIAL POWER IMBALANCE sha 1- be the THERMAL POWER in the top half I l of the core expressed as a percen age of RATED THERMAL POWER minus the l THERMAL POWER in the bottom half o the core expressed as a percentage l 1

of RATED THERMAL POWER for each qua ant.

SHIELD BUILDING INTEGRITY l.24 SHIELD BUILDING INTEGRITY shall exi when:

a. Each door in each access. opening is closed except when the l~

access opening is being used for no 1 transit entry and exit, then at least one door shall be losed.

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b. The shield building filtration system is EJBLE. I i '

c.- The sealing mechanism associated with each p etration(e.g.

welds,'bellowsor0-rings)isOPERABLE.

s REACTOR PROTECTION SYSTEM RESPONSE TIME l 1.25 The REACTOR PROTECTION 3YSTEM RESPONSE TIME shall be tha time interval from when tha monitored parameter exceeds its trip set int at the channel sensor' until power interruption at the control rod dr ve breakers.

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E B&W-STS 1-5

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June 1, 1976

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t i REACTOR COOLANT SYSTFM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM U'lIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through steam gen-erators, and

d. 10 SPM IDENTIFIED LEAKAGE from the Reactor Coolant System and

e. .

) GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of (2230 + 20) psig.

APPLICABILITY _: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least H0T STANDBY within!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 ,

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY 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 .

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be

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within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor at least ance per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , )

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b. Monitoring the containment sump inventory and discharge at l

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , j

c. Measurement of the CONTROLLED LEAKAGE to the reactor coo i pump seals when the Reactor Coolant System pressure is 2230

+ 20 psig at least once per 31 days with the modulating valve ! ,

Tully open, June 1,1976 B&W-STS 3/4 4-15

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" REACTOR COOLANT.' SYSTEM i

i SURVEILLANCE REQUIREMENTS (Continued)

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d. Perfonnance of a Reactor Coolant System water inventory bale.ce at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation.

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t REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY 4

i LIMITING CONDITION FOR OPERATION l

i 3.4.8 The specific activity of the primary coolant shall be limited to: !

a. 1 1.0 pCi/ gram DOSE EQUIVALENT I-131, and

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b. 1 100/fpCi/ gram APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

4 MODES 1, 2 and 3*.

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a. With the specific activitv of the primary coolant > 1.0 uCi/ gram DOSE EQUIVAlt2T I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that operation under these circumstances shall not exceed 10". of -

the uniti's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

b. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during.

one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg <

l (500)*Fuithin6 hours.

c. With the specific cctivity of the primary coolant > 100/f l

pCi/ gram, be in at least HOT STANDBY with T"V9 < (500)*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4 and 5:

a. With the specific activity of the primary _ coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E uCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific ' activity of the primary coolant is I restored to within its limits. A REPORTABLE 0CCURRENCE shall be prepared and submitted to the Commission pursuant to Specificatior. 6.9.1. This report shall contain the results of- the specific activity analyses together with the following information:

With T avg t (500)*F.

B3-STS 3/4 4-20 June 1,=1976 I

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-i f REACTOR COOLANT SYSTEM' l f i

i ACTION: (Continued) l 1 1. Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the '

I first sample in which the limit was exceeded,

2. Fuel burnup by core region,

3. Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded,

4. History of de-gassing operations, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and a

5. The time duration when the specific activity of the primary coolant exceeded 1.0 uCi/ gram DOSE EQUIVALENT I-131.

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SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis l

l program of Table 4.4-4.

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-i l B&W-STS 3/4 4-21 June 1,1976 .

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' TABLE 4.4-4 to PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE E

AND ANALYSIS PROGRAM-h m

MODES IN WHICH SAMPLE

! SAMPLE AND -

TYPE OF MEASUREMENT AND ANAL't3IS REQUIRED ANALYSIS FREQUENCY

! AND ANALYSIS 1 1, 2, 3, 4 l

1. Gross Activity Determination At least once each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

Z{_ 1 Isotopic ~ Analysis for DOSE 1 per: 14 days 2.

EQUIVALENT I-131 Concentration

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Radiochemical for E Determination 1per} months

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a)---Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever 1,2,3,4,5

4. Isotopic Analysis for Iodine .the specific activity exceeds-N Including 1-131, I-133, and I-135 -

--1.0 pCi/ gram DOSE EQUIVALENT

' _ZG131 or 100/E pCi/ gram, and

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b)_ ;_~0ne sample between 2 and 6 1,2,3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months following a THERMAL

- - POWER change exceeding 15 per-

, JLcent of the RATED THERMAL

- POWER within a one hour period.

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I until the specific activity of the primary coolant system is restored within its limits.

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Sample to be taken afte'r a minimum of 2 EFPD.and 20 days of POWER OPERATION have tilapsed sin -

-@ reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months _or_ longer.

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! ' I 4 i i 4 i i . . 6 ,i i i.1- i i 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 i DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific

' Activity >1.Opci/ gram Dose Equivalent 1-131 l

i B&W-STS 3/4 4-23 Jun2 1,1976 l I 1

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' PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be

< 0.10 uCi/ gram DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the specific activity of the secondary coolant system > 0.10 pC1/ gram l DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ar.d in COLD SHUTDOWN within the following 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

i SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.

B&W-STS 3/4 7-7 June 1, 1976

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! TABLE 4.7-2_

1 I SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM i

SAMPLE AND l

TYPE OF MEASUREMENT ANALYSIS FREQUENCY I

AND ANALYSIS __

i Gross Activity Determination At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i l.

Isotopic Analysis for DOSE a) I per 31 days, whenever

2. the gross activity determina-EQUIVALENT I-131 Concentration tion indicates iodine concen-trations greater than 10%

of the allowable limit.

b) I per 6 months, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit.

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REACTOR COOLANT SYSTEM BASES 4 l 4

3/4.4.7 EMIngr

' css on Reactor Coolant System chemistry ensure that The li ctor Coolant System is minimized and reduce the co rrosion of the oolant System leakage or failure due to stress po tential for Reactor co rrosion. Maintaining e chemistry within the Steady State L shown on Table 3.4-1 provi or Coolant System over the life of the structural integrity of the Re ceeding the oxygen, chloride and plant. The associated effects of e dependent. Corrosion studies f

luoride limits are time and tempera contaminant concentration i its, l

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how that operation may be continued having awievels significant in exces olant System. The f

or the specified limited time intervals withoffect restrictions of on e

time interval permitting continued operation within tiv ctions to tate Limits.

the Transient Limits provides time for taking con-corre l

The' surveillance requirements provide ~adequat:? fficiant 'e assura centrations in excess of the' limits will be detected in _

to take corrective action.

E 3/4.4.8 SPECIFIC ACTIVITY _ l t The limitations on the specific activity of the primary ill not coo an 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 sboundary w

riately small fraction of the Part 100 limit follow umed exceed an aper f 1.0 with an steam generats tube rupture accident in conjunction steady state primary-to-secondary steam generatori leakag l

l The values for the limits on specific activity represent in limits based GPM.

upon a parametric evaluation by the NRC These values are conservative in the specific site para-d m locations.

meters of the site, such as site boundary location anThe NRC is fina conditions, were not considered i ite. in Thisthis evalu reevaluation of the specific activity limits of th s s reevaluation may result in higher limits.

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' June 1, 1976

, B 3/4 4-5 B&W-STS il i

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REACTOR COOLANT SYSTEM .

BASES ._

The ACTION statement permitting POWER OPERATION to continue for l imited time periods with the~ primary coolant's specific activity > 1.0 uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-i, accommodates possible iodineOperation spiking phenomenon which with specific ac- may occur following changes in THERMAL POWER.

tivity levels exceeding 1.0 uti/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T to < (500)*F prevents the release of activity should a steam generator *dbe rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be de-Informat!on obtained tected in sufficient time to take corrective action.

on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses follow-ing power cnanges may be permissible if justified by the data cbtained.

h4.9 PRESSURE /TENPERATURE LIMITS omponents in the Reactor Coolant System are designed to with-stand the e 'ects of cyclic loads due to system temperature and pressure changes. Thes cyclic loads are introduced by normal load transients, reactor trips, an tartup and shutcown operations. The various les used for design purposes are provided in categories of load c During heatup and cooldown, the rates of Section ( ) of the ~ AR. ges are limited so that the maximum speci-temperature and pressure c re consistent with the design assumptions fled heatup and cooldown rate clic operation.

and satisfy the stress limits fo During heatup, the thermal gradie in the reactor vessel wall pressive at the inner wall to produce thermal strasses which vary from ed compressive stresses tensile at the outer wall. These thermal in the internal pressure.

tend to alleviate the tensile stresses induced y state conditions i Therefore, a pressure-temperature curve based on st 11 similar (i.e., no thermal stresses) represents a lower bound o ssel is curves for finite heatup rates when the inner wall of the treated as the governing location.

B 3/4 4-6 June 1,1976 l B&W-STS 6

PLANT SYSTEMS _

BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEMS _

The OPERABILITY of the auxiliary feedwater systems ensures that the Reactor Coolant System can be cooled down to less than (305)"F from

' normal operating conditions in the event of a total loss of offsite power.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of (350) gpm at a pressure of (1133) psig ta the entrance of the steam generators. Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of (700) gpm This at a pressure of (1133) psig to the entrance of the steam generators.

capacity is sufficient to ensure that adequate feedwater flow is available

'o remove decay heat and reduce the Reactor Coolant System temperature to

,less than (305)*F where the Decay Heat Removal System may be placed into j aperation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than (305)*F in the The event of a total minimum loss of offsite power or of the nain feedwater system.

water volume is sufficient to maintain the RCS at HOT STANDBY conditions for ( ) hours with steam discharge to atmosphere concurrent with loss of offsite po..er. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical cnaracteristics.

3/4.7.1.4 ACTIVITY j The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small This fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

dose 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 safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the B 3/4 7-2 June 1, 1976

,,B&W-STS c m q

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ML19340A245

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