ML20024H618
| ML20024H618 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/03/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20024H617 | List: |
| References | |
| NUDOCS 9106070052 | |
| Download: ML20024H618 (7) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _
'g UNITFD STATES l'
NUCL C AR REGULATORY COMMISSION a
{.
I WA$NWG TON. D, C. 20665 s,..._... /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 93 AD 83 TO FACILITY OPERATING LICENSE NOS. NPF-10 AND NPr*15
, SOUTHERN CAllFORNIA EDISON COMPANY SAN O! EGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA E
THE CITY OF_ ANAHEIM. CALIFORNIA SAN ONOFRE NUCLEAR GENERATING STATION. UNIT NOS. 2 AND 3 DOCKET NOS. 50-361 AND 50-362 i
1.0 INTRODUCTION
By letter dated Apri' 8,1991, Souther n California Edison Company (SCE or the licensee) requested thanges to the Technical Specifications (TS) for Facility Operating License Nose 1PF-10 and NPF-15 that authorize operation of San Onofre Nuclear Generating Station, Unit Nos. 2 and 3 in San Diego County, California.
-The licensee has requested to revise TS 3/4 6.1.2, " Containment Leakage.
Specifically. _the licensee proposed to amend TS Surveillance Requirement-4.6.1.2.a and the associated Bases to permit the third Type A _ test of each 10-year inservice interval to be conducted during a separate plant outage from the 10-year plant inservice inspection.
2.0 EVALUATION Appendix J requires that a set of three Type A tests be performed during each year service period with-the third test being conducted when the plant is
. shut down for the 19-year plant inservice inspection. The proposed TS change would eliminate the requirement of conducting the third Type A test of a 10-year service period during the shutdown for-the 10-year unit inservice-inspection.
The purpose for requiring the third Type A test during shutdown for the 10-year plant inservice _ inspection is to assure that the three Type A tests are not Dunched together during the first 90 months of the 10-year operation-cycle.
Requiring the third Type A test during the 10-year plant inservice inspection assures that-the three Type A tests are evenly spaced over the L
10-year interval.
Tho' licensee has requested an amendment that would eliminate conducting the third: Type A test in a 10-year service period during the unit shutdown-for the 10-year inservice inspection (151). For example, the third Type A test for 9106070002 910603
{DR ADOCK0500gg1-
the first 10-year service period for San Onofre 2 is scheduled for the San Onofre Unit 2 Cycle 6 refueling outage in August 1991.
The licensee contends that, because the 10-year ISI has been extended beyond 1991, (Cycle 7 refueling outage) the inspection is not necessary for the Unit 2 Cycle 6 refueling outage and, therefore, must t>e uncoupled from the third Type A test in each 10-year service period which is required by Appendix J.
(A similar situation exists for Unit 3). To perform a fourth Type A test during the same shutdown as the 10-year plant ISI (Cycle 7 refutling outage) would only satisfy the Technical Specification requirement to perform a Type A test during the same shutdown for the 10-year plant ISI.
$1ditionally, pcrforming a fourth containment ILRT, for the sole purpose of bei.g done during the same outage as the 10-year 151, would not necessarily enhance Je purpose, or provide further assurance of containment integrity, above that which has already been demonstrated. Moreover, the licensee intends to conduct the three Type A tests at 40210 month intervals during each 10-year service period.
Additionally, not extending the inservice inspection would impose hardship on the licensee with little or no increase in the level of quality of safety.
This inspection is not related to containmant integrity requirements of Appendix J.
The purpose of the Appendix J test program is to ensure that leakage through the primary reactor containment and systems and the components penetrating primary containment does not exceed allowable leakage rate values.
The purpose of the ASME Code Section XI inservice inspection program is to ensure that structural integrity of Class 1, 2, and 3 components is maintained in accordance with ASME Code requirements. Therefore, the proposed separation has no safety consequences because the requirements on containment integrity in Aopendix J and the TSs, and on structural integrity of Class 1, 2, and 3 components in the ASME Code are not being changed by the proposed change to TS 4.6.1.2.a.
The staff has considered uhe amendment request for uncoupling the third Type A test of each 10-year service period from the 10-year unit ISI and concludes it is justified on the grounds that the third Type A test within each 10-year service 7eriod and the 10-year ISI may be scheouled seperately, and the safe operation of San Onofre Unit Nos. 2 and 3 does not require that the two tests be conducted in the same outage.
The licensee is still required to conduct
'0-year 101 in accordance with Section XI of the ASME Code, ite, ore, based upon the information presented above, the staff finds the n,.cndment request acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the anendment. The State official l
had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendnents change a requirement with respect to the installation or use of l
a facility component located within the restricted area as defined in 10 CFR l
part 20. The NRC staff has determined that the amendments involve no I
i r
--,-n--,---
~
~ *
~
3-significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the anendments involve no significant hazards consideration, and there has been no public comment on such finding.. Accordingly, the anendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environnental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 C_0NCL US10N The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Connission's regulations, and (3) the issuance of the anendment will not be inimical to the comnon defense and security or to the health and safety of the public.
Principal Contributor: Lawrence E. Kokajko Date: June 3,1991 l-l I
l l
c l
.. ~ - - -. -... -
3/4.6 CONTAINMENT SYSTEMS 3/4.6-1 PRIMARY CONTAINMENT
' CONTAINMENT-INTEGRITY LIMITING CONDITION FOR OPERATION
- 3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY:
MODES t, 2, 3 and 4 ACTION:
Without-primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one 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 />.
SURVEILLANCE REQUIREMENTS
- 4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
At least once per 31 days by verifying that all penet.ations* listed a.
in Sections A, B and C of Table 3.6-1 not capable of being closed ty L
OPERABLE containc.ent automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or automatic valves secured ** in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.
b.
By verifying that each containment air-lock is in compliance with-the -requirements of Specification 3.6.1.3.
L c.
After each closing _of each penetration subject to Type B. testing, except-containment air locks,.if. opened following a Type A or B
-test, by leak rate testing the seal with gas at P 55.7 psig and a
verifying that when the measured leakage rate for these seals is added to the_ leakage rates determined pursuant to Specifica-tion 4.6.1.2.d for all other-Type B and-C penetrations, the combined leakage rate is-less than 0.60 L,.
\\
- Except valves, blind flanges,-and automatic valves which are located inside l
the cantainment and are locked, sealed or otherwise secured in the closed position.
These penetrations shall be verified closed during each COLD l
SHUTDOWN except that such verificati'on need not be performed more often than once per 92 days.
I
~
- Locked, sealed or otherwise. prevented from' unintentional operation.
I SAN ON0FRE-UNIT 3 3/4 6-1 AMEN 0 MENT N0.35 l
l-
~...... -
m._..
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION i
l 3.6.1.2 Containment leakage rates shall be limited to:
a.
_ An overall integrated leakage rate of:
1.
Less than or equal to L 0
containment air per 24 Ro,urs.10 percent by weight of the "
at P,, 55.7 psig, or 2.
Less than or equal to L, 0.05 percent by weight of the containment air per 24 flours at e reduced pressure of P,
g 27.9 psig.
b.
A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,.
APPLICABILITY:
MODES 1, 2, 3 and 4, ACTION:
With either (a) the measured overall integrated containment leakage rate exceeding 0.75L,allpenetr$tionsandvalvessubjecttoTypesBandCtests
-or 0.75 L, as applicable, or (b) with the measured combined leakage rate-for exceeding 0.60 L,, restore the overall integrated leakage rate to less than or equal to 0.75 L combinedleakag$orlessthanorequalto0.75L,asapplicable,andtherateforallpenetr tests to less than 0.60_L, prior to increasing the Reactor Coolant System temperature above 200*F.
SURVEILLANCE REQUIREMENTS L
4.6.1.2 The containment leakage rates shall be demonstrated at_the following test schedule and shall-be determined in conformance with the criteria specified in Appendix J of -10 CFR 50 using the irethods and provisions of ANSI N45.4 - 1972:
(_
a.
Three Type A te',ts (Overall Integrated Containment Leakage Rate) shall be condue:;d at 40 + 10 month intervals during shutdown at-either P service $er(55.7psig)oratP-(27.9psig)-duringeach10 year t
iod.
Prior to the Type A tests a visual inspection shall
-be conducted-in accordance with Specification 4.6.1.6 to demonstrate the containment structural integrity.
SAN ON0FRE-UNIT 3 3/4 6-2 AMENDMENT NO. 83
,o 3/4.6 CONTAINMENT SYSTEMS, l
BASES at 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioacti materials from the containment atmosphere will be restricted to those le'ya akage paths and associated leak 6ates 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 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 measured overall integrated leakage rate is further lim;ttd to less than or equal to 0.75 L, or less than or equal to 0.75 L, as applicable during t
performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
The surveillance testirg for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR 50.
The following exception, however, applies:
The third Type A test of each 10 year inservice interval, need not be conducted when the unit is shutdown for the 10 year plant inservice inspection.
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 that the overall air lock leakage will not become excessive due to seal damage durin{Jtheintervalsbetweenairlockleakagetests.
s a
l SAN ONOFRE-UNIT 3 B 3/4 6-1 AMENDMENT NO. 83 l
i CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE I
The limitations on centainment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 5.0 psig, 2) thg. con-
]
tainment peak pressure does not exceed the design pressure of 60 psig during LOCA or steam line break conditions, and 3) the assumptions used for the initial conditions of the LOCA and safety analysis remain valid.
The maximum peak pressure expected to be obtained from a LOCA or steam line break event is 55.7 psig.
The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 57.2 psig which is less than the design pressure and is consistent with the accident analyses.
3/4.6.1.5 AIR TEMPERATURE The limitation on containment average air temperature ensures 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.
3/4.6.1.6 CONTAINMENT 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 i
the facility.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 55.7 psig in the event of a steam line break tecident.
The measurement of containment tendon lift off force, the tensile tests of the tendon wires or strands, the visual examination of ten-dons, anchorages and exposed interior and exterior surfaces of the contain-ment, the chemical and visual examination-of the sheathing filler grease, and the Type A leakage tests are sufficient to demonstrate this capability.
The surveillance requirements for demonstrating the containment's struc-tural integrity are in compliance with the recommendations of Proposed Revi-sion 3 to Regulatory Guide 1.35, " Determining Prestressing Forces for Inspec-tion of Prestressed Concrete Containments," April 1979; and Proposed Regula-tory Guide 1.35.1, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," April 1979.
SAN ONOFRE-UNIT 3 8 3/4 6-2 4
.