ML20082U097
| ML20082U097 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/11/1991 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20082U100 | List: |
| References | |
| NPF-37-A-042, NPF-66-A-042, NPF-72-A-031, NPF-77-A-031 NUDOCS 9109190293 | |
| Download: ML20082U097 (18) | |
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COMMONWEALTH EDISON COMPANY _
l DOCKET NO. STN 50-454 8YRON STATION, UNIT NO. 1 AMENDMENTTOFACILITYOPERATINGLICENQ Amendment No. 42 License No NPF-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated October 4, 1988, as supplemented on August 14, 1989, March 20 and April 8, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No NPF-37 is hereby amended to read as follows:
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. (2) Technical Specifications The Technical Specifications contrined in Appendix A as revised through Amendment No. 42 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate 1
the f acility in accordance with the Technical Specifications and the j
Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance, FOR T NUCLEAR REGULATORY COMMISSION to r Richard J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to t;1e Technical Specifications Date o,' Issuance: September 11, 1991 l
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BYRON STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. NPF-66 1.
The_ Nuclear. Regulatory Commission (the Commission) has found that:
A.
The application for amendrent by-Commonwealth Edison Company (the licensee) dated: October 4,1088, as supplemented on -August 14, 1989, March 20 and April 8.-1991, complies with the standards and requirements of the Atomic Energy Act of'1954, as amended (the Act) and the Commission's rules and regulations setuforth in 10 CFR Chapter 1; B.
The facility'willl operate in conformity with the application.
.the provisions of the Act, and the rules and regulations of the
-Commission; C.-
There is reasonable assurance (i) that the activities authorized by this amendment can be. conducted without endangering the health Land safety of the public, and (ii)-that such activities will be
' conducted in compliance with--the-Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of.the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is. amended by changes to.the Technical Specifi-cations as indicated in the attachment to this license amendaent, and paragraph 2.C.(2)ofFacilityOperatingLicenseNo.NPF-66'ishereby amended to read as follows:-
b
. (2) Technical Specifications The-Technical Specifications contained in Appendix A (NOREG-1113),
as revised through Amendment No. 42 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license.
-Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR T NUCLEAR REGULATORY COMMISSION
/
a 4nr Richard J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - lil/lV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Tectnical Specifications Date of Issuance:
September 11, 1991 4
4
ATTACHMENT TO LICENSE AMENDMENT NOS. 42 AND 42 FACILITY OPEftATiNG LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages as indicated by an asterisk are provided for convenience.
Remove Pages insert Pages 3/4 6-5 3/4 6-5
- 3/4 6-6
- 3/4 6-6 B 3/4 6-1 B 3/4 6-1
- B 3/4 6-2
- B 3/4 6-2 i
m.
i CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS i
4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
a.
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by (1) Verifying that the door seal leakage is less than 0.0024La (1.11 SCFH) when the volume between the door seals is pressurized to greater than or equal to 3 psig by means of a permanently installed continuous pressurization and leakage monitoring sys-tem, or (2) Verifying that the door seal leakage is less than 0.01La (4.63 I
SCFH) as determined by precision flow measurements when measured
(
for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to 10 psig-i b.
By conducting overall air lock leekage tests at not less than P,,
44.4 psig, and verifying the overall air lock leakage rate is within its limit:
1)
At least once per 6 months,* and l
2)
Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**
c.
At.least once per 6 months by verifying that only one door in each air lock can be opened at a time.
d.
At least once per 6 months by verifying thit the seal leakage is less than 0.01La (4.63 SCFH) as determined by precision flow measurements when measured for at least 30 seconds with the volume between tF' seals at a constant pressure of greater than or equal to 10 psig; l
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- The previsions of Specification 4.0.2 ar: 'ot applicable.
l
- This repre ents an exemption to Appendix J of 10 CFR Part 50, Paragraph III h
D.2(b)(ii).
i BYRON - UNITS 1 & 2 3/4 6-5 Amendment No. 42
l CONTAINMENT SYSTEMS 4
INTERNAL PRESSURE
?
LIMITING CONDITION FOR OPERATION l' '
3.6.1.4 Primary containment internal pressure shall be maintained between
-0.1 and +1.0 psig, j
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
With the containment internal pressure outside of the limits above, restore i
the internal pressure to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
BYRON - UNITS 1 & 2 3/4 6-6 t
e 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1-PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAIWMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation-doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P.
As an added conservatism, the a
measured overall integrcted leakage rate is further limited to less than or equal to 0.75 L or 0.75 L, as applicable, during performanca of the periodic a
t test to account for possible degradation of the containment leakage barriers between-leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50.
3/4.'6.1.3 CONTAINMENT AIR LOCKS The limitations on c'osure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
The use of precision flow measurements of Specift:ation 4.6.1.3.a(2) must be used whenever the L
continuous monitoring capability in the control room is lost.
3/4.6.1.4 INTERNAL PRESSURE l-The limitations on containment internal pressure ensure that: (1) the l
containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.
l The maximum increase in peak pressure expected to be obtained from a cold leg double-ended break event is 44.4 psig.
The limit of 1.0 psig for initial positive containment pressure will limit the total pressure to 44.4 psig, which is higher than the FSAR Chapter accident analysis calculated peak pres-sure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the design pressure of 50 psig.
BYRON - UNITS 1 & 2 8 3/4 6-1 Amendment No. 42
CONTA1NMENT SYSTEMS BASES i
3/4.6.1.5 AIR TEMPERATURE l
i The 14.nitations on containment average air temperature ensure that the overall c9ntainment average air temperature does not exceed the initial temperatJre Condition assumed in the accident analysis for a steam line l
break accident.
Measurements shall be made at all of the listed running fan i
locations, whether by fixed or portable instruments, to determine the average air emperature, 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structurd integrity of the containment I
will be maintained comparable to'the originai design standards for the life of j
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 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 i
of tendons, anchorages and exposed interior and exterior surfaces of the
- 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 Rev. 3 to Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tencons 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.
The required Special Reports from any engineering evaluation of containment abnormalties shall include a description of the tendon condition, the condition f
of the concrete (especially at tendon enchorages), the inspection procedure, J
the tolerances on cracking, the results of the engineering evaluation and the j
corrective actions taken.
r 3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM i
The 48-inch containment r r;;e supply and exhaust isolation valves are required to be sealed closed spower removed) during plant operations since these i
valves have not been demonstrated capable of closing during a LOCA or steam line break accident.
Maintaining these valves sealed closed during plant operation 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 i
accordance with Standard Review Plan 6.2.4 which incluces mechanical devices to D
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 BOUNDARY dose guideline values of 10 CFR Part 100 would not BYRON - UNITS 1 & 2 B 3/4 6-2 t
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7 COMMONWEALTH EDISON COMPANY DOCKET-NO. STN 50 456-BRAIDWOOD STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 31 License No.-NPF-72
- 1. - LThe Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated October 4, 1988, as supplemented on' August 14, 1989, March 20 and April 8, 1991, complies with-the standards.
and requirements of the Atomic Energy Act of 1954,-as amended (the Act) and the Comission's rules and regulations set forth in It CFR Chapter I; B.-
'The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comissiori; C.
There:1s reasonable assurance (i) that the activities authorized by.this amendment can be conducted without endangering the health-rand: safety of--the public, and (ii)-that such activities will be l
- conducted-in compliance.with the Comission's regulations; l
D.
The issuance of this amendment will not be inimical to the comon
-defense and security or to the health and safety of the public;-
and-e E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
-2.-
Accordingly, the license is amended by changes to the lechnical Specifi- -
cations as-indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility _ Operating License No. NPF-72 is hereby amended to read as follows:
O
w l 1 (2) -Technical Specifications, The Technical Specifications contained in Appendix A as revised through Amendment No. 31 and_the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications an';
the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION fY kk-a for Richard J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - III/IV/V Of,* ice of Nuclear Reactor Regulation Attachtent:
Changes to the Technical Specifications Date of Issuance:
September 11, 1991 s
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendmetit No. 31 License No. NPF-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated October 4, 1988, as supplemented on August 14, 1989, March 20 and April 8, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulatiors; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all apr?icable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
P
6.
2 (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 31 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective ar, of the date if its issuance.
FOR T NUCLEAR REGULATORY COMMISSION LA W
Richard J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - I!!/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 11, 1991
ATTACHMENT TO LICENSE AMENDMENT NOS. 31 AND 31 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET N05. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A"_Teciinical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. Overleaf pages as indicated by an arterisk are provided for convenience.
Renove Pages insert Fages 3/4 6-5 3/4 6-5
- 3/4 6-6
- 3/4 6-6 B 3/4 6-1 B 3/4 6-1
- B 3/4 6-2
- B 3/4 6-2 x
k CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. -
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by (1) Verifying that the door seal leakage is less than 0.0024La (1,11 SCFH) when the volume between the door seals is pressurized to greater than or equal to 3 psig by means of a permanently installed continuous pressurization and leakage monitoring sys-tem, or (2) Verifying that the door seal leakage is less than 0.01La (4.63 SCFH) as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to 10 psig; b.
By conducting overall air lock leakage tests at not less than P,,
44.4 psig, and verifying the overall air lock leakage rate is within its limit:
1)
At least once per 6 months,* and 2)
Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability **
c.
At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
d.
At least once per 6 months by verifying that the seal leakage is less than 0.01La (4.63 SCFH) as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to 10 psig;
- The provisions of Specification 4.0.2 are not applicable.
- This represents an exemption to Appendix J of 10 CFR Part 50, Paragraph III D.2(b)(ii).
BRAIDWOOD - UNITS 1 & 2 3/4 6-5 Amendment No. 31
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' CONTAINMENT SYSTEMS INTERNAL PRESSURE-LIMITING CONDITION FOR OPERATION-
' 3. 6. L 4 Primary containment internal pressure shall be maintained between
-0.1 and +1.0 psig.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
With the containment internal pressure outside of the limits above, restore tha internal pressure to within the limits within I hour 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 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 />.
T SURVEILLANCE REQUIREMENTS 4.6.1.4.The primary containment internal pressure sh'all be determined to be within~the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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=BRAIDWOOD'- UNITS 1 & 2 3/4 6-6 l
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3/4.6 CONTAINMENT SYSTEMS i
r BASES-3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1' CONTAINMENT INTEGRITY Primary CONTAINMENT ' INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be rastricted to those leakage e
paths and associated leak rates assumed in the safety analyses.
This i
restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 CONTAINMENT LEAKAGE ~
i The-limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P.
As an added conservatism, the 3
- measured overall integrated leakage rate is further limited to less than or equal to 0.75.L r O E L, as applicable, during performance of the periodic a
g test to account for possible degradation of the containment leakage barriers i
between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50.
3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provides assurance ttet.
[
the overall air -lock leakage will not become excessive due to seal damage i
during the intervals between air lock leakage tests.
The use of precision flow measurements of Specification 4.6.1.3.a(2) must be used whenever the continuous monitoring capability in the control room is lost, i
i 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design n,ative pressure differential with respect to the outside atmosphere of 0.1 psy, and (2) the t
- containment peak pressure does not exceed the design pressure of 50 psig during steam:line break' conditions.
I The maximum increase in peak pressure expected to be obtained from a cold leg double-ended break event is 44.4 psig.
The limit of 1.0 psig for initial positive containment pressure will limit the total pressure to 44.4 psig, which is higher than the FRAR Chapter accident analysis calculated peak pres-sure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the design pressure of 50 psig.
t BRAIDWOOD - UNITS 1 & 2 B 3/4 6 Amendment No. 31 I
CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial 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.
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 colu leg double ended break accident.
The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the 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 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 Prestressed Concrete Containments," April 1979.
The required Spacial Reports from any engineering evaluation of containment abnormalties shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evaluation and the corrective actions taken.
-3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed (power removed) during plant operations since these valves have not been demonstrated capable of cloi,Ing during a LOCA or steam line break accident.
Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be rel'ased e
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.
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 BOUNDAkY dose guideline values of 10 CFR Part 100 would not BRAIDWOOD - UNITS 1 & 2 B 3/4 6-2
,.