ML20056C355

Proposed TS 3/4.6.1.7, Containment Purge Ventilation Sys

ML20056C355
Person / Time
Site:	Byron, Braidwood
Issue date:	05/13/1993
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20056C354	List: ML20056C353 ML20056C355
References
NUDOCS 9305190390
Download: ML20056C355 (27)

Text

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ATTACH _ MENT B P_BOPO_ SED _GHANGES_TOAPfENDIX A IECHNICALSP_ECIFICAIlONSEQB FACILITY OPERATING _LIGENSES NPF-3LNPF-66. NPF-72 AND NPF-77 Revised Pages 3/4 6-11 3/4 6-12 B 3/4 6-2

• B 3/4 6-3 l

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• No changes. Included for continuity.

ZNLD/2605/4

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9305190390 930513 PDR ADDCK 05000454 P PDR Q 'M

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4 l CONTAINMENT SYSTEMS ,

' i CONTAINMENT PURGE VENTILATION SYSTEM i

i LINITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valves st.all be OPERABLE and: i

a. Each 48-inch containment shutdown purge supply and exhaust isolation l valve' shall be closed and power removed, and

b. The 8-inch containment purge supply and exhaust isolation valve (s)

m =y 5: :;:n Sr up : 1000 h:;r: d:rin; :--; % w ; ::i :: r& -

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nr: 'an :n; "n: h :p:n :t,:n: tin.

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

ACTION:

a. With a 48-inch containment purge supply and/or exhaust isolation '

yaive open and/or powered,- close and remove . power to dsolate. tne penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. _ _

b. With the 8-inch containment purge supply and/or exhaust isolation ,

valve (s) open fortmecc th:n 1000 h= r: 3 rin; ; n kr.hr yar, close l )

the open 5-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, i otherwise 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 i COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. '

c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi- l cations 4.6.1.7.3 and/or 4.6.1.7.4, restore the inoperable valve (s) '

to CPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise be in at least HOT STANOBY 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 COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, rensoes oh<- Oh gwew. w $'gec <.t h w 3 .lo. % a b t.

BRAIDWOOD - UNITS 1 & 2 3/4 6-11 -

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! CONTAINMENT SYSTEMS 1

4 i SURVEILLANCE REQUIREMENTS l i i

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4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve (s)  ;

shall be verified closed and power removed at least once per 31 days. -

i 4.6.1.7.2 TM Or hth: ti = 1520 011 "-i d . n :Mf = t ;;.;; :g;1y =d/:r 1  :.2:rt Schth: ::h n M n M: :;x t r' q : :: h.-22 y::r :Ml' 5:

l LvceA 1 0 C t :'=d :t u nt a s Ex 7 ty:. l i 1 4.6.1.7.3 At least once per 6 months on a STAGGERED TEST 8 ASIS, the inboard j and outboard valves with resilient material seals in each closed 48-inch ,

1 containment purge supply and exhaust penetration shall be demonstrated OPERABLE

! by verifying that the sensured leakage rate is less than 0.05 L, when pressurized l j to at least P ,, 44.4 psig. l J l i

4.6.1.7.4 At least once per 3 months, each 8-inch containment purge supply and'  ;

j' exhaust isolation valve with resilient material seals shall be demonstrated i j 'uPEnnBLE oy veruying:tnat .tne measureadeaaage rataJs.less tnan 0.01 L, wnen pressurized to at least P,, 44.4 psig. t j  !

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CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the i overall containment average air temperature does not exceed the initial i temperature condition assumed in the accident analysis for a steam line  !

break accident. Measurements shall be made at all of the listed running fan ,

locations, whether by fixed or portable instruments, to determine the average  !

air temperature. 1 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment  !

will be maintained comparable to the original design standards for the life of  !

the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 44.4 psig in the event of a cold leg i double-ended break accident. The measurement of containment tendon lift-off j force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the i containment, and the Type A leakage test are sufficient to demonstrate this capability.

The Surveillance Requirements for demonstrating _the containment's structural integrity are in compliance with the recommendations of proposed i Rev. 3 to Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons .

i in Prestressed Concrete Containment Structures," April 1979 and proposed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of  :

Prestressed Concrete Containments," April 1979. l r

The required Special Reports from any engineering evaluation of containment  !

abnormalties shall include a description of the tendon condition, the condition i of the concreta (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evaluation and the  !

corrective actions taken. t 3/4.6.1.7 CONTAlleqENT PURGE VENTILATION SYSTEM ,

)

The 48-inch containment purge supply and exhaust isolation valves are i required to be sealed closed (power removed) during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line i' break accident. Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be released 1 via the Containment Purge System. To provide assurance that the 48-inch contain-  ;

ment valves cannot be inadvertently opened, the valves are sealed closed in  !

accordance with Standard Review Plan 6.2.4 which includes mechanical devices to i seal or lock the valve closed, or prevents power from being supplied to the valve operator.

The use of the containment purge lines is restricted to the 8-inch purge supply and exhaust isolation valves since, unlike the 48-inch valves, the 8-inch valves are capable of closing during a LOCA or steam line break accident.

Therefore, the SITE B0UNDARY dose guideline values of 10 CFR Part 100 would not 8 3/4 6-2 'C BRAIDWOOO - UNITS 1 & 2

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CONTAINMENT PURGE VENTILATION SYSTEM (Continued)

! be exceeded in the event of an accident during containment ; rging :;;r:ti:n.

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MSerE W 2 i2 Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit of Specification 3.6.1.2.b. shall not be exceeded when the leak!ge rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERA 8ILITY;of the Containment

• Spray System: ensures ~that canteimment depressurization and cooling capability wil1~be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and v Containment Cooling System are redundant to each other in providing post-accident cooling of the containment atmosphere. However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume Ilmit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3% ensures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons.

BRAIDWOOO - UNITS 1 & 2 8 3/4 6-3

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! l InserL1  ;

4 ...shall be closed except that these valves may be open for purge system operation for  :

l reasons such as containment pressure control, reduction of airborne activity, respirable j air quality considerations for personnel entry, surveillance tests that require the valve (s) i

. to be open, and for other safety-related purposes.

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i lDaftrL2 l l j Each 8-inch containment purge supply and exhaust isolation valve shall be verified to

be positioned in accordance with Specification 3.6.1.7b at least once per 31 days.

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t InserL3 a

.. PURGING or VENTING operation. Only safety-related reasons such as containment  ;

pressure control or the reduction of airborne activity to facilitate personnel access for l surveillance or maintenance activities should be used to justify opening these isolation valves. Other safety-related reasons would include maintaining limits set out in the i Technical Specifications, Updated Final Safety Analysis Report, Code of Federal  !

t Regulations, Regulatory Guides, Safety Evaluation Reports or other licensing basis  !

, documentation.

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ZNLD/2605/5

CONTAINMENT SYSTEMS CONTAINMENT PURGE VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valves shall be OPERABLE and: I

a. Each 48-inch containment shutdown purge supply and exhaust isolation valve shal1 be closed and power removed, and

b. The 8-inch containment purge supply and exhaust isolation valve (s)

I , may t;; ; pen f;; up t; 1000 heur; during a calendir y ;r pr;vid d n;

! lNSEE } acre than ene line is open at en; time.

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

ACTION:

a. With a 48-inch containment purge supply and/or exhaust isolation valve open and/or powered, close and remove power to isolate the penetretion(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise 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 COLD SHUTDOWN within the following 30-hours. g .

7gy 4g

b. Withthe8-inchcontainmentpurgefsupplyand/orexhaustisolation valve (s) open for nHw: th;n 1000 \;;r: dur'r.g : : m Ms year, close the open 8-inch valve ( orisolatethepenetration(s)within4 hours,l otherwise 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi-cations 4.6.1.7.3 and/or 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

BYRON - UNITS 1 & 2 3/4 6-11

CONTAINMENT SYSTEMS SURVEI(LANCE REQUIREMENTS 4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve (s) shall be verified closed and power removed at least once per 31 days.

I 4.6.1.7.2 The cumulat;;; ti; that all 0-inch containment pu ge supply end/cr I I

"*cxh:::t i:01stier value; have 5 r Op n during a :si:nder year shell be lMfERTd determined et least once pe, 7 d.y .

4.6.1.7.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard valves with resilient material seals in each closed 48-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 L when a pressurized to at least P,, 44.4 psig.

4.6.1.7.4 At least once per 3 months, each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L awhen pressurized to at least P,, 44.4 psig.

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l BYRON - UNITS 1 & 2 3/4 6-12

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CONTAINMENT SYSTEMS incidad (or con % h on i BASES "ON5 to M P3" i

l 3/4.6.1.5 AIR TEMPERATURE 1

i The limitations on containment average air temperature ensure that the j overall containment average air temperature does not exceed the initial i temperature condition assumed in the accident analysis for a steam line i break accident. Measurements shall be made at all of the listed running fan 1 locations, whether by fixed or portable instruments, to determine the average

air temperature.

) 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY

! This limitation ensures that the structural integrity,of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 44.4 psig in the event of a cold leg i double-ended break accident. The measurement of containment tendon lift-off i'

force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the ,

i containment, and the Type A leakage test are sufficient to demonstrate this i capability.

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j The Surveillance Requirements for demonstrating the containment's

structural integrity are in compliance with the recommendations of proposed l Rev. 3 to Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons

in Prestressed Concrete Containment Structures," April 1979 and proposed

Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of i Prestressed Concrete Containments," April 1979.

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The required Special Reports from any engineering evaluation of containment >

i abnormalties shall include a description of the tendon condition, the condition

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of the concrete (especially at tendon anchorages), the inspection procedure, j the tolerances on cracking, the results of the engineering evaluation and the

! corrective actions taken.

• 3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM l

The 48-inch containment purge supply and exhaust isolation valves are j required to be sealed closed (power removed) during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line

! break accident. Maintaining these valves sealed close.d during plant operation i ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 48-inch.contain-ment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to

seal or lock the valve closed, or prevents power from being supplied to the valve operator. t The use of the containment purge lines is restricted to the 8-inch purge >

]'

supply and exhaust isolation valves since, unlike the 48-inch valves, the

8-inch valves are capable of closing during a LOCA or steam line break accident.

Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not i

BYRON - UNITS 1 & 2 B 3/4 6-2 i

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i BASES l Inscri 3 i CONTAINMENT PURGE VENTILATION SYSTEM (Continued)

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be exceeded in the event of an accident during containment purging /g:perati n. l

. Oper ti:n with :n: lin: Op:n will b: limited : 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />; during : ::1:nd:r 1 yeen s

j Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit

' of Specification 3.6.1.2.b. shall not be exceeded when the-leak!ge rat 6s

! determined by the leakage integrity tests of these valves are added to the a previously determined total for all valves and penetrations subject to Type B 4 and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS I

i 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment i leakage rate are consistent with the assumptions used in the safety, analyses.

1 l The Containment Spray System and the Containment Cooling System are redundant to each other in providing post-accident cooling of the containment j atmosphere. However, the Containment Spray System also.provides a mechanism i

for removing iodine from the containment atmosphere and therefore the time

' requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

[ 3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH

' volume and concentration ensure a pH value of between 8.5 and 11.0 for the j solution recirculated within containment after a LOCA. This pH band minimizes l the evolution of iodine and minimizes the effect of chloride and caustic stress 2 corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable be'cause of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3%. ensures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons.

a BYRON - UNITS 1 & 2 B 3/4 6-3

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i i ...shall be closed except that these valves may be open for purge system operation for

- reasons such as containment pressure control, reduction of airbome activity, respirable l air quality considerations for personnel entry, surveillance tests that require the valve (s) i to be open, and for other safety-related purposes.

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4 I loseIt2 Each 8-inch containment purge supply and exhaust isolation valve shall be verified to l

be positioned in accordance with Specification 3.6.1.7b at least once per 31 days.

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1 f ... PURGING or VENTING operation. Only safety-related reasons such as containment j pressure control or the reduction of airbome activity to facilitate personnel access for j surveillance or maintenance activities should be used to justify opening these isolation

valves. Other safety-related reasons would include maintaining limits set out in the

! Technical Specifications, Updated Final Safety Analysis Report, Code of Federal Regulations, Regulatory Guides, Safety Evaluation Reports or other licensing basis

documentation.

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ZNLD/2605/5 f

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A'ITACHMENT C SAFETY EVALUATION i

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8-INCH CONTAINMENT PURGE SYSTEM OPERATION SAFETY EVALUATION 3-

1.0 BACKGROUND

The current technical specification requirement governing operation of the 8-inch containment purge supply and exhaust isolation valves allows only one line (supply or exhaust) to be opened at a time and specifies a cumulative time limit that the valves can be opened. The proposed revision to Technical Specification Section 3/4.6.1.7 would allow simultaneous opening of the purge supply and exhaust isolation valves for containment pressure control, ALARA, and respirable air quality considerations for personnel entry or surveillance tests that require the valve (s) to be open. Both the purge supply and exhaust isolation valves close on a containment isolation signal as before.

2.0 LICENSING BASIS Title 10 of the Code of Federal Regulations, Section 50.59 (10 CFR 50.59) allows the holder of a license authorizing operation of a nuclear power ,

facility the capacity to initiate certain changes, tests, and experiments ,

not described in the Final Safety Analysis Report (FSAR). Prior Nuclear l Regulatory Comission (NRC) approval is not required to implement the i modification provided that the proposed change, test, or experiment does I not involve an unreviewed safety question or result in a change to the plant technical specifications incorporated in the license. While the proposed change to the 8-inch Purge System involves a change to the Byron and Braidwood technical specifications and requires a licensing amendment request, this evaluation will be performed using the method outlined under 10 CFR 50.59 to provide the bases for the determination that the proposed change does not involve an unreviewed safety question. In addition, an evaluation will demonstrate that the proposed change does not represent a significant hazards consideration, as required by 10 CFR 50.91 (a) (1) and will address the three test factors required by 10 CFR 50.92 (c).

3.0 EVALUATIONS The Westinghouse evaluations are limited to the following areas:

radiological consequences, LOCA and LOCA-related accidents, non-LOCA 1 transients, and M/E releases and containment analyses. The following  !

impacts were considered: i

- Impact on dose calculations due to an additional release path. '

- Impact on initial containment conditions for postulated accidents.

Impact on postulated transient results with valves open.

The results of these evaluations are discussed in the following sections.

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- . 3.1 Radiological Ccnscquences Evaluation '

The offsite doses that would result if the 8-inch containment purge i systcm t:as in operation at the initiation of a large break LOCA were i cal'culated in accordance with Branch Technical Position CSB 6-4 i guidelines. The large break LOCA is considered to bound other

{ postulated accidents.

! The radiological consequences of the releases through the 8 inch j containment purge system at the beginning of a hypothetical LOCA are summarized below, i

j Station Site Boundary Low Poculation Zone i

! Byron Thyroid Dose: 9.1 rem 0.27 rem l Gama-body: 0.017 rem 0.0005 rem i

i Braidwood Thyroid Dose: 12.3 rem 1.13 rem i Gama-body: 0.023 rem 0.0021 rem These doses are based on the following assumptions: )

I 1. Primary coolant iodine concentration is at the maximum Tach Spec  !

l value of 60 uti/g Dose Equivalent I-131 (pre-existing iodine l j spike).

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2. Purge valves are isolated at 7.62 seconds into the accident, t

i 3. Only 29% of the containment free volume is used as the mixing volume.

! 4. The flow from the containment to the atmosphere during the time j the miniflow purge system is open is assumed to go through the t supply and exhaust lines and is assumed to increase as a function

of containment pressure during the Double-Ended Cold Leg (DECL)

LOCA.

I 5. The coolant activity released inside containment is based on the DECL mass releases.

The doses, when added to the LOCA dose contributions from containment

leakage and from leakage of recirculated ECCS solution, do not add a

, significant amount to the total. The largest increases are in the

site boundary thyroid dose. The Byron total LOCA dose increases from 115 rem to 124 ren, and the Braidwood total LOCA dose increases from i 155 rem to 167 rem. The dose increases represent a small fraction of 4

the previously calculated dose totals. The increased doses remain within the limits of 10 CFR 100.

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s . 3.2 LOCA and LOCA R21atcd Evaluations 3.2.1 :Large Break LOCA -- (FSAR Chapter 15.6.5) 1 The current large break LOCA licensing basis analysis for

Byron /Braidwood was performed using tne models described in Reference

! (3) and are presented in Reference (2). The Byron /Braidwood Stations Updr.ted FSAR Chapter 6.2 section 6.2.1.5 contains an evaluation for l the effect on large break LOCA calculated peak cladding temperature (PCT) with containment purge. This evaluation already accounted for .

. both valves being open prior to the LOCA and assumed a valve closure l

! time of 5.0 seconds. The impact on containment pressure resulting

! from the loss of air or steam is less than 0.05 psi. This evaluation 1

was based on results from an older LOCA analysis performed for l Byron /Braidwood using the February 1978 version of the Westinghouse ECCS evaluation model. While the current LOCA analysis for Byron /Braidwood is now based on the newer 1981 + BASH model, the 1 evaluation appearing in the Updated FSAR is still valid and bounds i any exp;ted effect of containment purge on the calculated peak l cladding temperature. This conclusion is based on Westinghouse internal sensitivity studies which have demonstrated the conservatism in the older.models. Given that the assumption for purge valve closure in 5.0 seconds does not change, the large break LOCA analysee

calculated consequences increase peak cladding temperature by a negligible amount (l'F). Thus, the analysis consequences and 1 assumptions are not changed by the proposed change to the Byron and Braidwood Technical Specifications. The proposed change will result in a l'F PCT margin penalty for the Byron and Braidwood Stations.

3.2.2 Small Break LOCA - (FSAR Chapter 15.6.5) i j The current small break LOCA licensing basis analysis for the Byron /Braidwood Stations (Reference 2) was performed with the model described in References (4) and (5). Small break LOCA analyses as performed by Westinghouse, do not model the containment or take j credit for containment pressure above atmospheric pressure. Thus,  ;

the calculated small break PCT is not affected by containment purge )

or changes to the containment purge Technical Specification. Thus, i 1 implementation of the new Technical Specification for 3/4.6.1.7 1

allowing both the purge supply and exhaust valves to be open simultaneously will not result in any hypothesized small break exceeding the 10 CFR 50.46 criteria. j 3.2.3 Blowdown Reactor Vessel and Loop Forces - (UFSAR Chapter 3) i The licensing basis LOCA hydraulic forces analysis results found in Reference (1) calculate the peak loads occurring within the first 0.5

, seconds of the transient. 'During this period the break flow is .

critically limited and is therefore unaffected by changes in  !

, containment pressure. Thus the occurrence of the peak loads is well before any containment back pressure effects could change the calculated break flow and therefore the consequences associated with LOCA hydraulic forces. Therefore, the proposed change to the 4

,e . Technical Specificaticns covaring containment purge at the l Byren/Braidwood Staticns, cill not affset the consequences or j ccoclusiens of the LOCA hydraulic fcrees analysis found in the j Byron /Braidwood Stations Updated FSAR.

7 l 3.2.4 Post-LOCA Long-Term Core Cooling (FSAR Chapter 15.6.5) i

{ Reference (6) presents- the Westinghbuse licensing commitment to keep j the reactor core subcritical with all control rods out (ARO)

! following a hypothetical large break LOCA and crediting only the l boron provided by the ECCS. Reference (6) is cited in the

! Byron /Braidwood Updated FSAR Chapter 15.6.5 and is part of the

! licensing basis for the Byron /Braidwood stations. Reference (7) has

! provided clarification to the utilities on the Westinghouse licensing

. position. Since the Post-LOCA subcriticality is based on large break i requirements and further assumes that the contents of the RWST have 'i been injected and reside in the RCS/ Containment sump Post-LOCA, this  !

licensing requirement is not sensitive to assumptions for' containment  ;

pressure during a hypothetical LOCA. Thus, the proposed change-to  !

the Technical Specification covering containment purge at  ;

j Byron /Braidwood Stations has no effect on this requirement. j

) 3.2.5 Hot Leg Switchover to Frevent Potential Boron Precipt 4 tion )

i (FSAR Chapter 6.3).  ;

1 )

i The calculationt performed to determine the time (Post-LOCA) at which  !

j the boron concentration in the reactor vessel would exceed the  ;

solubility limit do not explicitly model the reactor containment and  ;

are independent of assumptions for containment pressure. Thus, the '

j -

proposed change to the Technical Specification covering containment

pun;e at Byron /Braidwood Stations has no effect on this licensing ,

requirement. I 3.2.6 Loss of Heat Sink Emergency Response Guidance A Loss of Heat Sink (LONS) eve 3t is not a design basis accident for PWRs, but following the accident

at Three Mile Island the NRC required the utilities to improve operator training and guidance. As part of these required improvements, Westinghouse performed analyses of inadequate core cooling scenarios in order.to provide improved emergency operating procedures. An inadequate core cooling condition could be developed by a loss of all secondary feedwater following a reactor trip. Analyses of this scenario are presented in Reference (8). A LOHS event behaves similarly to a small break LOCA in that high RCS pressure exists during the core uncovery period and persists through the recovery period. Simil'ar to the small break LOCA analysis assumptions, credit is not taken in the analyses for containment pressure higher than atmospheric pressure and therefore, changes in containment purge will not affect the conclusions drawn from Reference (8). Therefore, the LORS recovery technique of." Bleed and Feed" initiated upon symptoms of Steam Generator Dry-out will remain effective in preventing inadequate core cooling.

3.3 N:n-LOCA Evaluaticn The.ntn-LOCA safety analyses pras nted in Chapter 15 of th2 UFSAR are notPadversely affected by changing the method of containment purge.

This activity does not affect normal plant operating parameters, accident mitigation capabilities, the assumptions used in the  !

non-LOCA transients, or create conditions more limiting than those 1

enveloped by the current non-LOCA analyses. Therefore, the j conclusions presented in Chapter 15 of the UFSAR remain valid and the

! proposed evaluation for continued operation does not constitute an unreviewed safety question.

i 3.4 Mass and Energy Release / Containment Analyses '

Relaxation of Tech Spec Section 3/4.6.1.7 to allow simultaneous opening of the 8-inch containment purge supply and exhaust isolation

• valves does not adversely affect the short term and the long term i LOCA and MSLB mass and energy releases and the containment analyses.

This change does not affect the normal plant operating parameters, system actuations, accident mitigating capabilities or assumptions important to the containment analyses, or create conditions more .

) limiting than those assumed in these analyses. Therefore, the  !

i conclusions presented in -the SAR remain valid with respect to the i containment analyses.

1,

4.0 CONCLUSION

S The proposed change to the Technical Specifications and the impact to the licensing basis analyses of allowing simultaneous opening of the 8-inch l containment purge supply and exhaust isolation valves have been evaluated .

j by Westinghouse. The analyses and evaluations have determined that the proposed operation of the 8-inch containment purge system will have no adverse impact upon the licensing basis analyses. The impact of the 4 proposed change on the large-break LOCA doses has been calculated and the ,

radiological consequences associated with the proposed change do not

significantly change the dose totals. The dose totals remain within the limits of 10CFR100. It is not necessary to specifically limit the time ,

l these valves may be opened during normal operation since the effluent j through these valves is monitored for radioactivity and the amount of time

the valves may be opened is a function of the actual releases. If the radioactivity being released is within the allowable 10 CFR Part 50 Appendix I guidelines for continuous release, the valves do not require a l t -

specific limitation on the duration they are opened. In addition, if the radioactivity being released exceeded allowable limits, the valves would t

receive a signal to close. Also note that the reasons for opening these valves is specifically limited to the conditions stated in the 3 specifications.

' The recommended Technical Specification changes and a No Significant Hazards evaluation are presented as attachments to this safety evaluation.

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Th2 preposed ch[ng2 to allow simultannus Cp;ning of th2 8-inch '

centainment purge system supply and exhaust isolation valves does not represcnt an unreviewed safety questien par the definition and i

requirements of 10CFR50.59 based en the following justification.

1. Simultaneous opening of the 8-inch containment purge supply and sxhaust isolation valves does not increase the probability of an accident previously evaluated in the FSAR. The isolation valves do not initiate any accident discussed in the FSAR. Component and (

system performance will not be adversely affected since equipment and j system design criteria continue to be met. The containment isolation  :

signals which automatically close the 8-inch isolation valves are not i affected by the proposed change.  ;

1

2. The consequences of an accident previously evaluated in the FSAR are  ;

not increased due to simultaneous opening of the 8-inch containment I purge supply and exhaust isolation valves. A conservative analysis  ;

demonstrated that the radiological releases through the purge system  :

would be small. The largest additional dose calculated with  !

conservative assumptions was approximately 9 rem thyroid dose at the Byron Station site boundary and 12 rem thyroid dose at the Braidwood Station site boundary. This additional dose was added to the LOCA doses previously reported in Chapter 15 of the t!FSAR. The new LOCA i doses remain well within the limits specified in 10CFR100.

3. Simultaneous opening of the 8-inch containment purge supply and  !

exhaust isolation valves does not create the possibility of an )

accident which is different than any already evaluated in the FSAR.  :

The proposed change is in the operation, not the design of the ,

equipment. The containment isolation signal received by the 8-inch containment purge System is unchanged and the closure time for the isolation valves are also unaffected by the proposed change. No new .

failure modes have been defined for any system or component important ,

to safety nor has any new limiting single failure been identified.  !

Therefore, the possibility of an accident different than any already ,

evaluated is not created.

4. The proposed change to the 8-inch containment purge system operation will not increase the probability of a malfunction of equipment important to safety previously evaluated in the FSAR. The ,

containment isolation capability is unaffected by this change. i Component and system performance will not be adversely affected since l equipment and system design criteria continue to be met. There is no  !

impact on any other safety system to perform its intended function in mitigating any accidents.

5. Simultaneous opening of the 8-inch containment purge supply and exhaust isolation valves will not increase the consequences of a malfunction of equipment important to' safety previously evaluated in the FSAR. The change in the operation of the valves will not degrade any system performance such that its malfunction will adversely affect another transient. Therefore, no more severe dose consequences will result due to this modification.

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, . . 6. Simultan rus op;ning of the 8-inch containment purge supply and exhaust isolaticn valves will not create the possibility of a malfunction of cquipment important to safety different than any

- already evaluated in the FSAR. The change in the operation of the valves does not affect the original design and performance criteria for the system and has not introduced a new limiting single failure for these systems.

7. The margin of safety as defined in the basis of the Technical Specifications is not significantly reduced by the change in the 8-inch containment purge system operation. All acceptance criteria with respect to fuel, RCS pressure boundary, and containment integrity continue to be met.

Based on the information presented above, it can be concluded that the proposed change to the 8-inch containment purge system as described in this evaluation does not involve an unreviewed safety question as defined in 10 CFR5 0.59.

5.0 REFERENCES

1) BYRON /BRAIDWOOD STATIONS UPDATED FSAR

2) Westinghouse Letter 89CB*-G-0083, J. W. Swogger (W) to Dr. W.

Naughton (CECO), " VANTAGE 5 RELOAD TRANSITION SAFETY REPORT FOR THE BYRON /BRAIDWOOD STATIONS UNIT 1 AND 2", JULY 1989, Letter Dated July 18, 1989.

3) WCAP-10266 REY.2 with Addenda (PROPRIETARY), "THE 1981 VERSION OF THE WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE",

AUGUST, 1986.

4) WCAP-10079-P-A (PROPRIETARY) and WCAP-10080-A (NON-PROPRIETARY)

"NOTRUMP: A NODAL TRANSIENT SMALL BREAK AND GE!!ERAL NETWORK CODE", AUGUST 1985.

5) WCAP-10054-P-A (PROPRIETARY), " WESTINGHOUSE SMALL BREAK LOCA ECCS .

EVALUATION MODEL USING THE NOTRUMP CODE", AUGUST 1985. l

6) WCAP-8339 (NON-PROPRIETARY), " WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL - SUlmARY", JUNE 1974.

7) WESTINGHOUSE TECHNICAL BULLETIN NSID-TB-86-08, " Post-LOCA LONG-TERM COOLING: BORON REQUIREMENTS", OCTOBER 31, 1986

8) WCAP-9914 (PROPRIETARY), "PORY SENSITIVITY STUDY FOR LOFW-LOCA ANALYSES", JULY 1981.

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L AUAC_HMENT D i

EV_ALUATION OF SIGNIRCANT HAZARDS CONSIDERATIONS .

I I

i On behalf of the Commonwealth Edison Company (CECO), the Westinghouse Electric Corporation has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10, Code of Federal Regulations, Part 50, Section 92, Paragraph (c) [10 CFR 50.92(c)], a proposed  ;

anr dment to an operating license involves no significant hazards considerations if -

opeimion of the facility in accordance with the proposed amendment would not: j

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of  :

accident from any accident previously evaluated; or  !

3. Involve a significant reduction in a margin of safety.  ;

CECO has reviewed the Westinghouse Electric Corporation evaluation and concurs ,

with the methodology and conclusions. The basis for this determination of no significant hazards considerations is presented below. l

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ZNLD/2605/6 I i

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F SIGNIFICANT HAZARDS EVALUATION FOR

~ BYRON S"ATION 8-INCH CONTAINMENT PURGE SYSTEM TECHNICAL SPECIFICATION CHANGE Intrcduction:

This analysis is provided in accordance with 10CFR50.92 to demonstrate that the proposed license amendment to revise Technical Specifications for the Containment Purge Ventilation System represents no significant hazards. The regulations state the. if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated, or 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of '

safety, then a determination of no significant hazards can be made.

1 Descrintion of Amendment Reauest: i The purpose of this amendment request is to revise Technical Specification 3/4.6.1.7 to permit the supply and exhaust valves of the 8-inch containment purge system (miniflow purge system) to.be opened simultaneously under certain l conditions. The existing Technical Specificttion states that the 8-inch- I containment purge supply and exhaust isolation valve (s) may be open for up to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> during a. calendar year provided no more than one line is open.at one time. When only the supply or the exhaust line of the miniflow purge system is open, the containment pressure is affected because of mass addition or deletion without.a corresponding input of outside air or release of containment atmosphere.

The change in Tech Specs will allow the miniflow purge system to be used without adversely affecting containment pressure. Simultaneous opening of both the supply and exhaust lines will equalize mass flow into and out of containment. This will improve system efficiency and reduce the number of

hours the miniflow purge system must be operated to create respirable air L conditions inside containment.

Basis for No Sionificant Hazards Detemination:

The proposed change does not affect the containment isolation function of the 1 8-inch containment purge system. The containment isolation signals which automatically close the 8-inch valves are not being altered by this change.

The bounding condition evaluated was miniflow purge supply and exhaust lines both open at the beginning of a postulated loss of coolant accident (LOCA).

Under these conditions, the containment atmosphere would have a path through the 8-inch containment purge system until the containment isolation valves are closed. The containment isolation valves are closed 7-seconds into the accident (2 seconds for signal generation and 5 seconds for valve operation).

,.. The propos::d chang 2s have been cvaluatcd in accordance tsith the Significant Hazards criteria of 20CFR50.92. The results of the evaluation demonstrate that thechangesd3notinvolveanysignificanthazardsasdescribedbelow.

1) A significant increase in the probability or consequences of an accident previously evaluated.

The proposed change was evaluated with respect to the consequences in the first few seconds of a LOCA, i.e. from the beginning of the accident until the miniflow purge system is isolated. A conservative analysis demonstrated that the radiological releases through the miniflow purge system would be very small.

The largest additional dose calculated with conservative assumptions was less than 4 rem thyroid dose at the Byron Station site boundary and less than 5 rem thyroid dose at the Braidwood Station site boundary. This additional dose was added to the LOCA doses previously calculated and reported in Chapter 15 of the UFSAR. The new dose total remains well within the limits of 10CFR100. It should be noted that there is significant margin between the thyroid doses and the 300 rem guideline specified in 10 CFR 100.

Simultaneous opaning of the 8-inch containment purge supply and exhaust isolation valves does not increase the probability of an accident previously evaluated. Component and system performance will not be adversely affected since equipment anri system design criteria continue to be met. The isolation valves do not initiate any accident discussed in the FSAR. The containment isolation signals which automatically close the 8-inch isolation valves are not affected by the proposed change.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created.

The proposed change is in the operation of the system and not in the design of the equipment. The containment isolation signal received by the 8-inch containment purge system is unchanged and the closure time assumed

! for the isolation valves are also unaffected by this change.

a y' The change in operation of the 8-inch containment purge system does not create new or different accidents. The initiation of postulated accidents

,\ does not depend on the 8-inch containment purge system. The proposed operation does not create a different accident from any previously evaluated because the purge system will continue to be isolated by the containment isolation signal.

~

3) Involve a significant reduction in a margin of safety.

The margin of safety as defined in the basis of the Technical Specificatit,ns is not significantly reduced by the change in the miniflow purge system operation. All acceptance criteria with respect to fuel, RCS pressure boundary, and containment integrity continue to be met, van % M W N W Y $ $ hr$ k h hh y

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ATTACIBEENT D ENVIRGSGBTAL ASSESSMENT STATEMENT Commonwealth Edison Company (CECO) has evaluated the proposed unendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.

It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided for under 10 CFR 51.22(c)(9). This determination is based on the following:

1. These changes are being proposed as an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to the operation of the mini-purge system. The proposed.

amendment would allow the concur. rent opening of the mini-purge supply and exhaust lines for the purposes of containment pressure control, reduction of airborne activity, respirable air quality considerations for personnel entry, surveillance tests that require the valve (s) to be open, and other safety-related purposes. There is a slight increase in the potential offsite dose as a result of this change due to an additional release pathway being open from containment at the onset of an accident. This slight increase in potential offsite dose has been evaluated by the Westinghouse Electric Corporation and determined not to be significant and well below the 10 CFR 100 limits,

2. the amendment involves no significant hazards consideration,

3. there is no significant increase in the amounts of any eff2uents that may be released offsite, and

4. there is no significant increase in individual or cumulative occupational radiation exposure.

. = .

i 6[TACHMENT E ENVIRQNMENTAL ASSEMENT STATEMENT Commonwealth Edison Company (CECO) has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided for under 10 CFR 51.22(c)(9). This determination is based on the following:

1. These changes are being proposed as an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to the operation of the mini-purge system. The proposed amendment would allow the concurrent opening of the mini-purge su:) ply and exhaust lines for the purposes of containment pressure control, rec uction of airborne activity, respirable air quality considerations for personnel entry, surveillance tests that require the valve (s) to be open, and other safety-related purposes.

There is a slight increase in the potential offsite dose as a result of this change due to an additional release pathway being open from containment at the onset of an accident. This slight increase in potential offsite dose has been evaluated by the Westinghouse Electric Corporation and determined not to be significant and well below the 10 CFR 100 limits,

2. the amendment involves no significant hazards consideration,

3. there is no significant increase in the amounts of any effluents that may be released offsite, and

4. there is no significant increase in individual or cumulative occupational radiation exposure.

l ZNLD/2605/7 1

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ATTACHMENT F Simpified Drawing - Containment Purge System i

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