ML20136G247

Proposed Tech Specs Re SGTR Evaluation

ML20136G247
Person / Time
Site:	Catawba
Issue date:	03/07/1997
From:	DUKE POWER CO.
To:
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ML20136G228	List: ML20136G224 ML20136G247
References
NUDOCS 9703170249
Download: ML20136G247 (15)

Text

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PLANT SYSTEMS 3.7.1.5 NOT USED STEAM GENERATOR POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION g

l Fod I

3.7.1.

-Three team generator power-operated relief valves (PORVs) and

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associat d r te manual controls, including the safety-related gas supply 1

systems, s all be OPERABLE.

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

i i

ACTION:

i I

a.

With one less than the required steam generator PORVs OPERABLE,

)

l-restore the inoperable steam generator PORV to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat j-removal.

1 b.

With two less than the required steam generator PORVs OPERABLE, j

restore at least one of the inoperable steam generator PORVs to r

OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least H0T STANDBY within i

the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

[

and place the required Residual Heat Removal loop in operation for decay heat removal, i

j SURVEILLANCE REQUIREMENTS l

4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated i

OPERABLE:

i a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that at least one of the two j

nitrogen bottles associated with each PORV has a pressure greater l

than or equal to 2100 psig, and l

b.

At least once per 18 months and prior to startup following any refueling shutdown by verifying that all steam generator PORVs will i

operate through one cycle of full travel using remote manual l

controls and safety-related gas supply.

L

{ +

• When steam generators are being used for decay heat removal.

?

I 9703170249 970307 1

PDR ADOCK 05000413

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P PDR I

CATAWBA - UNIT 1 3/4 7-9 Amendment No.

148

1 PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES i

4 The Surveillance Requirement for the Main Steam power-operated relief valves (PORVs) nitrogen supplies ensures that the PORVs will be available to mitigate the cor. sequences of a steam generator tu t e accident concurrent with loss of offsite power. This assu pg PORV on the-ruptured steam generator is unavailable, and tha<

...e c...c r o are used to N

"A hk N$

em m

esdi4ed e %e e<eni Jhr+ m:4e. ep*<*bu I4 u,utv41tabw r

w ro sb OPERABLE is the requirement that the associated PORV block valves upstream be j

o)en or OPERABLE.

Should an associated PORV block valve be closed and inoper-a)1e, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status.

Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION i

The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maxi-mum allowable fracture toughness stress' limits.

The limitations of 70'F and 200 psig are based on a steam generator RTNDT of 60'F and are sufficient to prevent brittle fracture 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

CATAWBA - UNIT 1 B 3/4 7-3

l PLANT SYSTEMS STEAM GENERATOR POWER OPERATED RELIEF VALVES J

j LIMITING CONDITION FOR OPERATION hoW 3.7.1.

-Three s am generator power-operated relief valves (PORVs) and associ ed rem manual controls, including the safety-related gas supply systems, s k e OPERABLE.

i APPLICABILITY: MODES 1, 2, 3, and 4.*

1 ACTION:

3 a.

With one less than the required steam generator PORVs OPERABLE, restore the inoperable steam generator PORV to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat removal.

b.

With two less than the required steam generator PORVs OPERABLE, restore at least one of the inoperable steam generator PORVs to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within l

the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat removal.

4 SURVEILLANCE REQUIREMENTS.

4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that at least one of the two nitrogen bottles associated with each PORV has a pressure greater than or equal to 2100 psig, and b.

At least once per 18 months and prior to startup following any refueling shutdown by verifying that all steam generator PORVs will operate through onc cycle of full travel using remote manual controls and safety-related gas supply.

1

• When steam generators are being used for decay heat removal.

f CATAWBA - UNIT 2 3/4 7-10 Amendment No.

142

PLANT SYSTEMS BASES l

3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES

{

The Surveillance Requirement for the Main Steam power-operated relief j

valves (PORVs) nitrogen supplies ensures that the PORVs will be available to mitigate the consequences of a steam generator tube ure accident concurrent with loss of offsite power. This ass pt t e PORV on the t..: 3 g ruptured steam generator is unavailable, and tha i

....r wo are used to em ratu Load h

mist Po945 d i

s maka A W uw h+ vasc cren% is uuvad4tt.

OP E is the t

0 ock valves upstream be open or OPERABLE. Should an associated PORV block valve be closed and inoper-able, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status.

Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maxi-mum allowable fracture toughness stress limits.

The limitations of 70*F and 200 psig are based on a steam generator RT of 60*F and are sufficient to ET prevent brittle fracture.

3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

CATAWBA - UNIT 2 B 3/4 7-3

PLANT SYSTEMS 3.7.1.5 NOT USED STEAM GENERATOR POWER OPERATED REllEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 Four steam generator power-operated relief valves (PORVs) and ascociated remote manual controls, including the safety-related gas supply systems, shall be OPERABLE.

Al+LICABILITY: MODES 1, 2, 3, and 4.*

ACTION:

a.

With one less than the required steam generator PORVs OPERABLE, restore the inoperable steam generator PORV to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat removal.

b.

With two less than the required steam generator PORVs OPERABLE, restore at least one of the inoperable steam generator PORVs to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for i

decay heat removal.

SURVEILLANCE RE0VIREMENTS 4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that at least one of the two nitrogen bottles associated with each PORV has a pressure greater than or equal to 2100 psig, and b.

At least once per 18 months and prior to startup following any refueling shutdown by verifying that all steam generator PORVs will operate through one cycle of full travel using remote manual controls and safety-related gas supply.

• When steam generators ne being used for decay heat removal.

CATAWBA - UNIT 1 3/4?-9 Amendment No.

-- -- a

._._____ _ _._._._._ _._._. _ _ _ _... _ m _.

PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES The Surveillance Requirement for the Main Steam power-operated relief valves (PORVs) nitrogen supplies ensures that the PORVs will be available to mitigate the consequences of a steam generator tube rupture accident concurrent with loss of offsite power. This assumes that the PORV on the ruptured steam generator is unavailable, and that at least two are used to 1

cool the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator. Local operation of the steam line PORVs is credited in the event that remote operation is unavailable.

Concurrent with the requirement that a specific number of PORVs be.

OPERABLE is the requirement that the associated PORV block valves upstream be open or OPERABLE. Should an associated PORV block valve be closed and.inoper-able, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status.

Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION l

The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maxi-mum allowable fracture toughness stress limits. The limitations of 70*F and 200 esig are based on a steam generator RTNDT of 60*F and are sufficient to prevent brittle fracture.

3/4.7.3 COMP 0NENT C0OLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

The redundant _ cooling capacity of this system, assuming a single failure, is consistent'with the assumptions used in the safety analyses.

CATAWBA - UNIT 1 B3/47-3

~

PLANT SYSTEMS STEAM GENERATOR POWER OPERATED RELIEF VALVES i

LIMITING CONDITION FOR OPERATION 3.7.1.6 Four steam generator power-operated relief valves (PORVs) and l

associated remote manual controls, including the safety-related gas supply 4

systems, shall be OPERABLE.

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

ACTION:

i a.

With one less than the required steam generator PORVs OPERABLE, restore the inoperable steam generator PORV to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat removal.

b.

With two less than the required steam generator PORVs OPERABLE, restore at least one of the inoperable steam generator PORVs to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required Residual Heat Removal loop in operation for decay heat removal.

SURVEILLANCE RE0VIREMENTS 4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that at least one of the two nitrogen bottles associated with each PORV has a pressure greater I

than or equal to 2100 psig, and i

b.

At least once per 18 months and prior to startup following any refueling shutdown by verifying that all steam generator PORVs will i

operate through one cycle of full travel using remote manual I

controls and safety-related gas supply.

• When steam generators are being used for decay heat removal, j

l CATAWBA - UNIT 2 3/47-10 Amendment No.

PLANT SYSTEMS BASES i

3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES The Surveillance Requirement for the Main Steam power-operated relief valves (PORVs) nitrogen supplies ensures that the PORVs will be available to mitigate the consequences of a steam generator tube rupture accident concurrent with loss of offsite power. This assumes that the PORV on the ruptured steam generator is unavailable, and that at least two are used to cool the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator. Local operation of the steam line PORVs is credited in the event that remote operation is unavailable.

Concurrent with the requirement that a specific number of PORVs be OPERABLE is the requirement that the associated PORV block valves upstream be open or OPERABLE. Should an associated PORV block valve be closed and inoper-able, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status.

Additionally, if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply in addition to Technical Specification 3.7.1.6.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maxi-mum allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are based on a steam generator RT of 60'F and are sufficient to m

prevent brittle fracture 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during nonnal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

l CATAWBA - UNIT 2 B3/47-3

~

Q 15.6 Decrease in Reactor Coolant inventory Catawba Nuclear Station consequences resulting from failure of the 3 inch CVCS letdown line is not expected to exceed a small fraction of the 10CFR 100 dose limits.

15.6.3 STEAM GENERATOR TUBE FAILURE 15.6.3.1 Identification of Causes and Accident Description The accident exarJned is the complete severance of a single steam generator tube. The accident is asst'med to take plaw. power with the reactor coolant contaminated with fission products corresponding to coutinuous operation with a limited amount of defective fuel rods. The accident leads to an increase in contanination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a ec, incident loss of offsite power or failure of the condenser steam dump system, discharge of activity

}

to the atmosphere takes place via the steam generator safety and/or power operated relief valves.

In view of the fact. hat the steam generator tube material is highly ductile loconel-600, the assumption of a complete severance is considered somewhat conservative. The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the Steam and Power Conversion System is subject to continual surveillance and an accumulation of minor leaks which exceeds the limit established in the Technical Specifications is nct permitted during the unit operation.

The operator is expected to determine that a steam generator tube rupture has occurred and to identify and isolate the affected steam generator on a restricted time scale in order to nunumze contamination of the secondary system and ensure termination of radioactive release to the atmosphere from the affected unit. The recovery procedure can be

'ed out oc a time scale which ensures that break flow to the secon t

' t 4e wate eve sffected steam gnerator rises into the main steam y

4 pipin (Maren* 5 to drene: 32 cn pc.; If 5,.' Sufficien. indications and controls are provided to en ble the' rator t arry t these etio ati actonly.

Immediately apparent symptoms of a tube rupture accident, such as falling pressurizer pressure and level and increased charging pump flow, are also symptoms of small steam line breaks and loss of coolant accidents. It is therefore important for the operator to determine that the accident is a rupture of a steam generator tube,in order to carry out the correct recovery procedure. The accident under discussion can be identified by the following method. In the event of a complete tube rupture, the reactor coolant system 3

pressure decreases and the condenser air ejector radiation monitor (if aligned) exhibit abnormally high readings. If the containment pressure, containment radiation, and containment recirculation sump level exhibit normal readings, then a steam generator rupture is diagnosed to have occurred.

4 Note that break sizes smaller than ccmplete severance of a tube, with less bre.k flow from primary to secondary, exhibit a slower rise in steam generator water level, and an increased time interval for actuation 3

of the condenser air ejector radiation monitor. Therefore, more time may be available to the operator to 4

diagnose the accident and take steps to isolate the ruptured steam generator.

If normal operation of the vanous plant control systems is assumed, the following events are initiated by a tube rupture:

1. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side, steam flowlfeedwater flow mismatch occurs as feedwater flow to the affected steam generator is reduced as a result of primary coolant break flow to that generator.

15-98 (30 NOV 1995)

_~,- -

_ ~ -

s f

6 Catawba Nuclear Station 15.6 Decrease in Reactor Coolant Inventory 4

2. He decrease in RCS pressure, due to continued loss of reactor coolant inventory, leads to a reactor trip signal on low pr;ssurizer pressure or overtemperature AT. The resultant plant cooldown following reactor trip leads to a rapid decrease in pressurizer level. A safety injection sign by low pressurizer pressure, follows soon after reactor trip. The safety injectiori signal automaticall -

terminates normal feedwater supply and initiates auxiliary feedwater addition.

3

3. The condenser air ejector radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system, and will automatically tenninate steam generator blowdown.

4. The reactor trip automatically trips the turbine and, if offsite power is available, the steam dump valves open, permitting steam dump to the condenser. In the event of a coincident station blackout 4

(loss of offsite power), as assumed in the analyses presented in this section, the steam dump valves automatically close to protect the condenser. The steam generator pressure rapidly increases resulting i

in steam dischtrge to the atmosphere through the steam generator safety and/or power operated relief j

valves. Steam flow as a function of time is constant initially until reactor trip. His is followed by turbine trip which results in a large decrease in flow, but a rapid increase in steam pressure to th-i safety valve setpoint.

1 j

5. Following reactor trip, the continued action of the auxiliary feedwater supply and borated safety i

injection flow (supplied from the RWST) provide a heat sink which absorbs the decay heat.

6. Safety injection flow results in increasing pressurizer water level, the rate of increase depending upon the amount of aunhary equipment operating.

2

7. In order to stop the leakage from the Reactor Coolant System to the ruptured steam generator, the 2

operator uses the intact steam generators to reduce the temperature of the primary coolant. This is i

2 accomplished ur,ing steam dump to the condenser or, in the absence of offsite power, the PORVs on 2

the steam lines of the int eam geneysors96hve RYrj rov' ed r

t am 2

n Air (

Syst the absence of the \\ l syste, nitro en cylinders in the Do ouse 2

m 4%une 4m. regge rpe,ah esould be use to operate these valves. Lcui cW. tag to.Mrs t

cru vA4 2

e reduction ' primary oolant tem rature enables the Reactor Coolant Sv m to m

2 u

e as e a

o lant ress

's reduce o appr x1hateh-tri'at o e rup d

2 steam generator. The pressure reduction e unmates riving orce for the primary-to-secondary 2

leakage. The reduction is accomplished using normal pressurizer spray. For a case in which the VI 2

system is unavailable (this system also provides motive force for the normal pressurizer spray valves 2

and the pressurizer PORVs), the operator aligns the cold leg accumulator nitrogen gas as a motive l

2 force for either of two pressurizer PORVs.

1 A steam generator tube failure is classified as an ANS Condition IV event, a limiting fault. See

" Classification of Plant Conditions" on page 15-1 for a discussion of Condition IV events.

15.6.3.2 Analysis of Effects and Consequences a

1 Method of Analysis 4

Three separate evaluations are performed for this accident. First, the offsite doses are calculated. Second, the margin to steam generator overfillis detennined. Third, the potential for DNB to cause fuel cladding 4

i 4

failures, which would increase the offsite doses, is evaluated. The separate evaluations sometimes make conflicting assumptions in order to conservatively determine the degree to which the separate acceptance 4

4 criteria are challenged.

I Detailed thermal hydraulic calculations are performed to determine primary to secondary mass release and 4

to determine the amount of steam vented from each of the steam generators, using the RETRAN-02 (30 NOV 1995) 15-99

.__.m i

i' 15.6 Decrease in Reactor Coolant Inventory Catawba Nuclear Statica 4

4 code, described in the introduction to this chapter, and using the methodology in Attachment 2 Reference 4

32 on page 15-119.

i 2

In stimatingqhe, mas hgu the broken tube for dose calculation purposes, the 1

Ilow g assumptions e mad.

}

1. D--eter '8p ~eem en mw op44frv ad,,,,

Reeder bri a 1

eur: --~m'!y at tbs-y '3n) m;m, s

.ss of otTsite power occurs at reactor tnp.

I g-the on signal, two high head safety injection pumps are 1

aligned to the safety injection flowpath and two intermediate-head safety inje-tion pumps are I

actuated. These pumps continue to deliver flow until safety injection is manually termmated by the 1

operator.

1

3. After reactor trip, bre flo eac es an equilibrium when it is balanced by incoming safety injection i

i flow as shown in Fi re 1 resultant break flow continues at approximately the same value i

I from plant trip until pr ssur qualized. Operator actions are modeled to terminate break flow.

1 The above assumptions, extremely conservative for the design basis tube rupture, are made to maxtmize 4

doses and do not model all expected operator actions for recovery. Plant characteristics and initial I

s" on page puw EasA4.et e,dewt4hhs art eddeAed to evA*k +k^*M*'**4 ov4C'll.

.h --us of s'--. aenenter credil' is deemen'-d i-Refernee ' en -

even when the steam generator tube rupture event is analyzed with M.!$Nw!20. The result is that, hich are conservative with respect to overfill, including the most limiting single failure, there is margin such that overfill is 4

4 avoided. Thl: meiodology, prepared by di: "'ennghuse Oune-: C--- p, E " de 1-# *e e.d" for ^+-wb2. 2: de~~e-'ed in At'eeh ent 5 te Refenne: 32 en pag id,15--119/tb4 r&delotj u' red l

cate=Ja+iew w4 prepared kst he teGjboMe ownei4 4ronf dLM dotam4ed tk EeCeesaebM M P*P 1%QD l

l j

e DNBR calculation for t accident is rformed with e VIPRE-computer co escribed i

4 duction t ch us tha 1istic esi de '

-in Reference on i

2 page i

DNBR is a concem for this transient because the assumed loss of offsite power causes a l

2 reactor coolant pump coastdown. Because of the loss ofinventory through the ruptured tube, the RCS 4

pressure is significantly lower than the normal operating value when the coastdown occurs. Since the loss 1

2 of offsite power is assumed to occur coincident with reactor and turbine trip, the amount of 2

depressurization prior to the coastdown would be limited by the overtemperature AT t ' function. This B6%ca e rel

'e e e en 2

's mt uce y

te 2

DN of the eatup and deprgsption allowed by this trip function, tgaggtgg rupture 2

coastdown transient from a A pr~ed RCS pressure is bounded by the 2

transient frem -- elev2ted RCS tenpertur - DNBR : Serefers-n^' expMt!y cele"h'ed fer %- t+

2 uptre '--%+, but the neu!! !: Sunded by 'he ~!e"'"!:: fer f=dwater line-break in Section 15.2.S.2,-

(

2

" A "y= ef Etrem ned Cen=quen=:" en p2;;e 15 A3, WM 1

R 4

The result et al hydraulic calculations for dose inputs are shown in the following figures:

i le'4 l

Figure 15404-ak Flow

}

toe 4

1 Fi 15409 Re ctor Coolant System Pressure 1

Figu ISM Rea tor Coolant System Temperature (For Ruptured Loop) j 1

Fi. e 15407 R ctor Coolant System Temperature (For intact loops) 1 Figu 15 Oh P ssurizer Water level 1

Figu 15-tNSt Line Pressure 15-100 (30 NOV 1995) i 4

w t

2 4

Catawba Nuc Station 15.6 Decrease in Reactor Coolant Inventory j

i tV6 1

Fi 15469 St Generator Water Levels 1

The sequence of es ents is presented in Table 15-49.

15.6.3.3 Environmental Consequences The postdated accidents invoking release of steam from the secondary system do not result in a O

signiscant release of radioactivity unless there is leakage from the RCS to the secondary system in the steam generators. A conservative analysis of the postulated steam generator tube rupture assumes the loss of offsite power. This causes the loss of main steam dump capabilities and the subsequent venting of steam from the secondary system to the atmosphere. A conservative analysis of the potential offsite doses j

resulting from this accident is presented assuming primary to secondary leakage.

This analysis incorporates assumptions of 1 percent defective fuel and steam generator leakage of I gpm prior to the

{

postulated accident for a time sufficient to establish equilibrium specific activities in the secondary system.

Three postulated cases are analyzed:

Case 1:

Normal equilibrium Technical Specification iodine concentrations exist at the time of the i

accident.

i Case 2:

here is a pre-existing iodine spike at the time the accident occurs. The reactor coolant 0

concentrations are the maximum permitted for full power operation (60 times the normal O

equilibrium Technical Specification limit).

i i

Case 3:

There is a coincident iodine spike at the time the accident occurs. The iodine concentrations are found by increasing the equilibrium appearance rate in the coolant by l

0 a factor of 500.

j l

3 The primary coolant activity prior to the accident correspond to limits set by Technical Specifications.

The following assumptions and parameters are used to calculate the activity release and offsite dose for the postulated steam generator tube rupture:

3

1. Prior to the accident, an equilibrium activity of fission products exists in the primary system.

2. The accident is initiated by the rupture of a steam generatcr tube, which results in the transfer of 3

approximately 174,000 pounds of reactor coolant into the shell side of the defective steam generator.

3. Offsite power is lost.

3

4. The primary to secondary leakage is.105 gal / min in each of the nondefective steam generators.

0

5. The steam release from the defective steam generator termmates in 65 minutes. He release from the nondefective steam generators terminates in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

6. All noble gases which leak to the secondary side are released.

0

7. The steam generator iodine partition factor is 0.01 during the accident.

8. For Case 1, the primary coolant concentration is at the equilibrium Technical Specification limit.

9. For Case 2, the primary coolant concentration is at the maximum permitted for full power operation 0

(60 times the normal equilibrium Technical Specification limit).

10. For Case,3, the iodine spike occurs at the onset of the accident and continues for the duration of the accident. The iodine concentrations are determined by increasing the equilibrium appearance rate by 0

a factor of 500.

11. Other assumptions are listed in Table 15-31.

I i

l (30 NOV 1995) 15 101

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16.10 STEAM AND POWER CONVERSION SYSTEM STEAM VENT TO ATMOSPHERE 4

l 16.10-1 COMMITMENT l

t L

F048 l

Thr c c - team generator PORV safety-related gas supply system e OPERABLE with both nitrogen bottles per S/G PORV pressurized to greater than or equal to 2100 psig.

APPLICABILITY:

Modes 1, 2, 3, and 4*

REMEDIAL ACTION:

1 a.

With one nitrogen bottle on one or more S/G's l

less than 2100 psig, immediately start corrective action to return the nitrogen supply to OPERABLE.

Work to return the nitrogen supply to OPERABLE status should continue without interruption.

b.

With two nitrogen bottles on one or more S/G's less than 2100 psig consider the PORV(s) inoperable and j

refer to Technical Specification 3.7.1.6 for the i

required ACTION.

TESTING REQUIREMENTS:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that both l

nitrogen bottles per S/G PORV has a pressure greater i

than or equal to 2100 psig.

1

REFERENCES:

1 1)

Design Basis Specification for the Catawba Main Steam, Main Steam Vent to Atmosphere and Main Steam Bypass to 1

Condenser System, Section 20.3.4 2)

PIR 0-C90-0304 3)

Branch Technical Position RSBS-1 4)

CNC-1223.43-01-0011, rev. 1 i

When Steam Generators are being used for decay heat removal.

16.10-1 2/91

o

$ 0 i

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ENVIRONMENTAL ASSESSMENT / IMPACT STATEMENT Pursuant to 10CFR51. 22 (b), an ' evaluation of this proposed j

amendment has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10CFR51.22 (c) (9) of the regulations.

3 proposed amendment to the Catawba Technical j

The 4

Specifications increases the numbers of PORVs required by i

one (1).

This change will not cause any additional volume or type of effluent available for any adverse environmental impact or personnel exposure.

Since it has been determined that there is:

1.No significant hazards consideration;

)

2. No significant change in the

types, or significant increase in the amounts, of any effluents that may be i-released offsite; and

3. No significant increase in-individual or cumulative j

occupational radiation exposures involved;

)

i j

the proposed amendment to the Catawba Technical Specifications and the UFSAR meets the criteria of I

I 10CFR51. 22 (c) ( 9) for categorical exclusion from an environmental assessment / impact statement.

a l

f

?

.ll

e i

REFERENCES 1)

H.

B.

Tucker Letter to USNRC, December 7, 1987, " Steam Generator Tube Rupture Analysis."

2)

WCAP-10698, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill."

WCAP-10698, Supplement 1, Evaluation of Offsite Radiation Doses for l

a Steam Generator Tube Rupture Accident."

l 3)

NUREG-0800,

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Reactors,"

Branch Technical Position RSB 5-1.

4)

M.

S.

Tuckman Letter to

USNRC, August 27,

1996,

" Supplement to Replacement S/G Proposed TS Amendment" l

5)

USNRC, Safety Evaluation Report by the Office of Nuclear Reactor Regulation Relating to Steam Generator Tube Rupture, May 14, 1991.

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