ML20236M271
| ML20236M271 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/05/1987 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20236M268 | List: |
| References | |
| NLS-87-132, NUDOCS 8708110002 | |
| Download: ML20236M271 (8) | |
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l ENCLOSURE 1 TO SERIAL: NLS-87-132 PROPOSED TECHNICAL SPECIFICATION PAGES BSEP-1 CONTAINMENT INTEGRATED LEAK RATE TESTS l
h5$$0h PDk4 (5224 BAT /lah) ni
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4 UNIT 1
SUMMARY
LIST OF REVISIONS P_a&e Description of Changes 3/46-3 Changed ANSI N45.4-1972 to ANSI /ANS 56.8-1981.
B3/4 6-1 Added paragraph describing exemption to Appendix 3.
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(5224 BAT /lah )
(BSEP-1-111)
CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued) c.
The leakage rate to less than or equal to 11.5 scf per hour for any one main steam line isolation valve, prior.to increasing reactor coolant system temperature above 212*F.
1 SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972, except that leakage rates for Type A tests shall be calculated using the Mass-Point method as specified.in ANSI /ANS 56.8-1981*.
The primary containment leakage rates shall be demcastrated at the following test schedule:
1 a.
Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 + 10 month intervals during shutdown at P,,
49 psig, or P, 25 psig, during each 10 year service period. The
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t third test of each set shall be conducted during the shutdown for the j
10 year plant inservice inspection.
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If any periodic Type A test fails to meet either 0.75 L, or 0.75 L,
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the test schedule.for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fall to meet 0.75 L, or 0.75 L, a Type A test shall be performed at each l
plantshutdownforrefbelingorevery18 months,whicheveroccurs l
first, until two consecutive Type A tests meet 0.75 L, or 0.75 L, at e
which time the above test schedule may be resumed.
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The accuracy of each Type A test shall be verified by a supplemental test which:
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Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L, or 0.25 L.
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Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
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Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalant to at least 25 percent of the total measured leakage rate at P,,
49 psig, or P, 25 psig.
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- Exemption from Appendix J of 10CFR50.
l BRUNSWICK - UNIT 1 3/4 6-3 Amendment No.
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(BSEP-1-111) 1 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIHARY CONTAINMENT INTEGRITY Primary CONTAINMENT 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 accident analyses. This l
restriction, in conjunction with the leakage rate limitation, will limit the l
site boundary radiation doses to within the limits of 10 CFR Part 100 during l
accident conditions.
1 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total l
containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49 psig, P. As an added conservatism, the measured overall integrated leakag,e rate is further limited to less than or equal to 0.75 L, or 0.75 L, as applicable, during performance of the periodic tests to account forpensibledegradationofthecontainment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of l
the valves; therefore, the special requirement for testing these valves.
Exemptions from the requirements of 10 CFR Part 50 have been granted for i
l main steam isolation valve leak testing, testing of airlocks after each l
opening, and leakage calculation methods.
Appendix J, paragraph III.A.3 requires that all Type A (Containment Integrated Leak Rate) tests be performed in accordance with ANSI N.5.4-1972,
" Leakage Rate Testing of Containment Structures for Nuclear Reactors."
ANSI N45.4-1972 requires that leakage calculations be performed using the Point-to-Point or Total Time method.
ANSI N45.4-1972 has been revised to a new standard, ANSI /ANS 56.8-1981, " Containment System Leakage Testing," which incorporates the Mass-Point method for leakage calculations. Type A tests will be performed in conformance with ANSI N45.4-1972 but will use the Mass-Point method for calculation of leakage rates as described in ANSI /ANS 56.8-1981.
3/4.6.1.3 PRIKARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
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i BRUNSWICK - UNIT 1 B 3/4 6-1 Amendment No.
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ENCLOSURE 2 TO SERIAL: NLS-87-132 l
PROPOSED TECHNICAL SPECIFICATION PAGES l
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CONTAINMENT INTEGRATED LEAK RATE TESTS i
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UNIT 2
SUMMARY
LIST OF REVISIONS Page Description of Changes 3/46-3 Changed ANSI N45.4-1972 to ANSI /ANS 56.8-1981.
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B3/4 6-1 Added paragraph describing exemption to Appendix 3.
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11 (5224 BAT /lah)
(BSEP-2-112)
P CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued) c.
The leakage rate to less than or equal to 11.5 scf per hour for any i
j one main steau line isolation valve, prior to increasing reactor coolant system temperature above 212*F.
SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the l
methods and provisions of ANST N45.4-1972, except that leakage rates for Type A tests shall be calculated using the Mass-Point method as specified in i
ANSI /ANS 56.8-1981*.
The primary containment leakage rates shall be demonstrated at the following schedule:
Three Type A Overall Integrated Containment Leakage Rate tests shall a.
be conducted at 40 10 month intervals during shutdown at P 49 psig, or P, 25 psig, during each 10 year service period.,,The t
third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspectio b.
If any periodic Type A test fail
.o meet either 0.75 L, or 0.75 L '
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the test schedule for subsequent i,-e A tests shall be reviewed and approved by the Comniission.
If two consecutive Type A tests fail to meet 0.75 L, or 0.75 L, a Type A test shall be performed at each g
plant shutdown for refueling or every 18 months, whichever occurs first, until two consecutive Type A tests meet 0.75 L, or 0.75 L, at t
which time the above test schedule may be resumed.
c.
The accuracy of each Type A test shall be verified by a supplemental l
test which:
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1.
Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L, or 0.25 L.
t 2.
Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
3.
Requires the quantity cf gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage P,,
49 psig or P, 25 psig.
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- Exemp; ion from Appendix J of 10CFR50.
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BRUNSWICK - UNIT 2 3/4 6-3 Amendment No.
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(BSEP-2-112) 3/4.6 CONTAINMENT SYSTEMS BASES i
3/4.6.1 PRIMARY CONTAINMENT 3 /4. 6.1 '.1 PRIMARi C3NTAINMENT INTEGRITY Primary CONTAINMENT INTECRITY ensures that the release of radioactive materials from the cricainment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
1 3/4.6.1.2 PRIHARY CONTAINMENT LEAKAGE The limitations on primary 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 of 49 psig, P. As an added conservatism, the measured overall integrated leakag,e rate is further limited to less than or equal to 0.75 L, or 0.75 L, as applicable, during performance of the periodic tests to account forpossibledegradationofthecontainment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valves has j
indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore, the special requirement for testing these valves.
Exemptions from the requirements of 10 CFR Part 50 have been granted for main steam isolation valve leak testing, testing of airlocks af ter each opening, and leakage calculation methods.
Appendix J, paragraph III.A.3 requires that all Type A (Containment Integrated Leak Rate) tests be performed in accordance with ANSI N45.4-1972,
" Leakage Rate Testing of Containment Structures for Nuclear Reactors."
ANSI N45.4-1972 requires that leakage calculations be performed using the i
Point-to-Point or Total Time method. ANSI N45.4-1972 has been revised to a new standard, ANSI /ANS 56.8-1981, " Containment System Leakage Testing," which incorporates the Mass-Point method for leakage calculations. Type A tests will be performed in conformance with ANSI N45.4-1972 but will use the Mass-Point method for calculation of leakage rates as described in ANSI /ANS 56.8-1981.
3/4.6.1.3 PRIHARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
BRUNSWICK - UNIT 2 B 3/4 6-1 Amendment No.
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