ML20217E189

Proposed Tech Specs Surveillance Sections & Bases Allowing Util to Defer 10CFR50,App J,Type a Testing of Byron Unit 2 Containment Until Next Refuel Outage in 1999

ML20217E189
Person / Time
Site:	Byron
Issue date:	03/24/1998
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20217E183	List: ML20217E177 ML20217E189
References
NUDOCS 9803300405
Download: ML20217E189 (9)

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ATTACIIMENT B j MARKED UP PAGES FOR l PROPOSED CIIANGES TO APPENDIX A I

TECIINICAL SPECIFICATIONS OF l FACILITY OPERATING LICENSES NPF-37 and NPF-66,

{ BYRON STATION UNITS 1 & 2 l

l REVISED PAGES:

3/4 6-1 3/4 6-2 3/46-3 l B 3/4 6-1 9803300405 980324 "

PDR ADOCK 05000454 P PDR I

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3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT I

CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION Primary CONTAINMENT INTEGRITY shall be maintained.

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 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 />.

SURVElllANCE RE0VIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

• not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in . Table 3.6-1 of Specification 3.6.3 or for containment isolation valves that are open under administrative controls;

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. By performing containment leakage testing in accordance with Regulatory Guide 1.163, September 1995,nand 10 CFR 50, Appendix J, Option B.

N Nd i u c g roved sched v ty e gh j

• Except valves, blind flanges, and deactivated automatic valves which are '

located inside the containment and are locked, sealed or othenvise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more i

often than once per 92 days.

BYRON - UNITS 1 & 2 3/4 6-1 AMENDMENTNO.[

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[0NTAINMENT SYSTEMS CONTAINMENT LEAKAGE tlMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

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a. An overall integrated leakage rate of less than or equal to L, at P,.

b. A combined leakage rate of less than 0.60 L for all penetrations l andvalvessubjecttoTypeBandCtests,w$enpressurizedtoP,.

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

ACTION:

With either the measured overall integrated containment leakage rate exceeding ,

0.75 L or the measured combined leakage rate for all penetrations and valves subjecl to Types B and C tests exceeding 0.60 L , restore the overall  ;

integrated leakage rate to less than 0.75 L, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEltlANCE RE0VIREMENTS  !

4.6.1.2 The containment leakage rates shall be demonstrated in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, l Option B.

a, Type A (Overall Integrated Containment Leakage Rate) testing shall be conducted in accordance with Regulatory Guide 1.163, j '

September 1995,oand 10 CFR 50, Appendix J Option B.

.f- 1 as mecAMed b) as qpecm ed f schm egh l 3/4 6-2 AMENDMENT NO.

BYRON - UNITS 1 & 2

CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)

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b. The reporting requirements and frequency of Type A tests shall be in accordance with Regulatory Guide 1.163, September 1995, and e -- J 10 CFR 50, Appendix J, Option B(Gs mtmMed by as cgprovec{ sched

c. The accuracy of each Type A test shaTTherTertired-by a suppiet::tti test conducted in accordance with Regulatory Guide 1.163, Septembe k * ?IC S 1995, and 10 CFR 50, Appendix J, Option B. h~

d. Type B and C tests shall be conducted in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;

f. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and

g. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be demonstrated during the shutdown for each Type A containment leakage rate test by a visual inspection of these surfaces. This inspection shall be performed at a frequency in accordance with Regulatory Guide 1.163, September 1995, to verify no apparent changes in appearance or other abnormal degradation.

h. The provisions of Specification 4.0.2 are not applicable.

BYRON - UNITS 1 & 2 3/4 6-3. AMENDMENT N0. 94 t

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3/4.6 CONTAINMENT SYSTEMS BASES i 3/4.6.1 PRIMARY CONTAINMENT .

3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive j 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 f 1

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 limited to less than or during performance of the periodic test to account for equal possible to degrada 0.75 L, tion of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates consistent with the requirements of Appendix J of 10 CFR Part 50, Option B, Regulatory Guide 1.163, September 1995, Nuclear Energy Institute document NEI 94-01, and ANSI /ANS-56.8-1994x { % y-1]

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 during the intervals between air lock leakage tests.

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 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.

The maximum increase in peak pressure expected to be obtained from a cold leg double-ended break event is defined as P . The limit of 1.0 psig for initialpositivecontainmentpressurewilllimitthetotalpressuretoP, which is higher than the UFSAR Chapter 15 accident analysis calculated p,eak pressure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably l'ess than the design pressure.of 50 psig.

BYRON - UNITS 1 & 2 B 3/4 6-1 AMENDMENT NO l

~ REVISED PAGES

[ Insert -1) except as modified for Unit 2. An extension for Unit 2 is allowed to perform the Type A test during B2RO8. Subsequent Type A test intervals for Unit 2 will be determined based on test results, in accordance with NEI 94-01.

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ATTACHMENT C EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATION Commonwealth Edison Company (Comed) proposes to revise Technical Specifications (TS) Surveillance Requirements 4.6.1.1.c ,4.6.1.2.a,4.6.1.2.b and the Bases to allow a schedular exemption to defer the test beyond the interval allowed in the Nuclear Energy Institute (NEI) document NEl 94-01, " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," which is endorsed by Regulatory Guide 1.163,

" Performance Based Containment Leak Test Program" to the Fall of 1999 for Byron Unit 2.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

An extension, by a maximurr. of 10 months, of the Type A test interval does not involve a change to any structures, systems, or components, does not affect reactor operations, is not an accident initiator, and does not change any existing safety analysis previously evaluated in the UFSAR. Therefore, there is no significant increase in the probability of an accident previously evalnated.

Several tables of UFSAR Chapter 15," Accident Analyses," provide containment leak rate values used in assessing the consequences of accidents discussed in this chapter. Although an extension can increase the probability that an increase in containment leakage could go undetected for a maximum of 10 months the risk resulting from this proposed change is inconsequential as documented in NUREG-1493," Performance-Based Containment Leakage Test Program" This document indicated that given the insensitivity of reactor risk to containment leakage rate and a small fraction ofleakage paths are detected solely by Type A testing, increasing the time between integrated leak rate tests is possible with minimal impact on public risk. Further, industry experience presented in this document indicated that Type A testing has had insignificant impact on uncertainties involved with containment ok rates.

Based on risk information presented in NUREG-1493, the proposed change does not increase the probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or difTerent kind of accident from any accident previously evaluated.

l l The proposed change does not alter the plant design, systems, components, or reactor operations, only the frequency of test performance. New conditions or l parameters that contribute to the initiation of accidents would not be created as a result of this proposed change. The change does not involve new equipment and existing equipment does not have to be operated in a different manner, therefore there are no new failure modes to consider.

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,4 An extension, by a maximum of 10 months, of the Type A test interval as shown in NUREG 1493 has no impact on, nor contributes to the possibility of a new or different kind of accident as evaluated in the UFSAR. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

With the exception of this 10 month extension of the Type A test interval, the actual tests will not change. Quantitative risk studies documented in NUREG-1493 regarding extended testing intervals demonstrated that there was minimal impact on the public health and safety. Reducing the frequency and allowing for a greater test interval, as stated in the NUREG resulted in an " imperceptible" increase in risk to public safety. Further, a table in this NUREG regarding risk impacts due to a reduction in testing frequency illustrates that there was also -

minimal difference in risk to the public safety when the test frequency was relaxed.

The proposed change will not reduce the availability of systems and components associated with containment integrity that would be required to mitigate accident conditions nor are any containment leakage rates, parameters or accident assumptions affected by the proposed change.

The proposed change does not involve a significant reduction in a margin of safety, based on the above information.

Based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.

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r ATTACHMENT D i ENVIRONMENTAL ASSESSMENT Comed has evaluated this proposed operating license amendment against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Comed has determined that this proposed license amendment meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) s '

and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards consideration.

As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards consideration.

(ii) there is no significant change in the types or significant increase in the amounts of any efIluent that may be released offsite.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment will not result in changes in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposed amendment result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

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