ML20199A465

Application for Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively.Amends Would Revise TS Surveillance Sections 4.6.1.1.c,4.6.1.2.a,4.6.1.2.b & Bases to Allow Performance of 10CFR50 App J,Type a Testing

ML20199A465
Person / Time
Site:	Byron, Braidwood
Issue date:	11/07/1997
From:	Hosmer J COMMONWEALTH EDISON CO.
To:	NRC
Shared Package
ML20199A469	List: ML20199A465 ML20199A475
References
NUDOCS 9711170192
Download: ML20199A465 (8)

Text

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, November 7,1997 United States Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Application for Amendment to Appendix A, Technical Specifications, for Facility Operating Licenses:

Ilyron Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF 37 and NPF 66 NRC2ncket Nos 50-4fiand 50-455 liraidwood Nuclear Power Station, Units I and 2 Facility Opetating Licenses NPF-72 and NPF-77 NRG]ncket Nos. 50-456 and 50-412 Containment Leak Rate Testing

Reference:

Letter from J.B. Ilosmer (Commonweahh Edison) to N5 C Document Control Desk dated September 10,1997 Pursuant to Title 10, Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90),

Commonwealth Edison Company (Comed) proposes to amend Appendix A, Technical Specifications, fa-F..cility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF-77 for Ilyron Nuclear Power Station, Units I and 2, and Braidwood Nuclear Power Station, Units I and 2, respectively. Comed proposes to revise Technical Specifications (TS)

Surveillance Sections 4.6.1.1 c,4.6.1.2 a,4.6.1.2.b and the bases to allow Comed to perform 10CFR50, Appendix J. Type A testing of Byron Unit 2 and liraidwood Unit 2

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containments at least once per 10 years based on a single successful Type A test, rather than two successful Type A tests. As discussed in the referenced letter, NRC staff and Comed agreed to use the license amendment process to modify the Type A test frequency. This modificatien of the test frequency requests one exemption from the requirements in NEl 94-01, that is, an exemption from performing two consecutive successful Type A tests prior to extending a testing interval.

4 This amenament request applies to Byron Unit 2 and liraidwood Unit 2 only. It is being docketed for Byron and liraidwood Units 1 and 2 because the Technical Specifications for

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Curreatly containment leakage testing is done in accordance with Regulatory Guide 1.163,

" Performance-Basv Containment Leak-Test Program" September 1995, and 10 CFR 50, l

Appendix J, Option B. The requirements for Type A testing are stated in the Nuclear Energy Institute (NEI) document NEl 94-01, " Industry Guideline for impleinenting Performance-Based Option of 10 CFR 50, Appendix J," which is endorsed by Regula'ory Guide 1.163. Section 9.2 3 of the NEl document allows Type A testing to be petformed 971117019297g7,.g

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U.S. Nuclear Regulatory Commission 2-November 7,1997

' at least once per 10 years based on acceptable performance history which is dermed as completion of two consecutive periodic Type A tests. Currently,Ilyron Unit 2 and liraidwood Unit 2 do not meet the acceptable performance requirements. Ilowever, Comed believes that Type A test requirements for Ilyron Unit 2 and Braidwood Unit 2 have been demonstrated through prior testing and maintenance practice.

The proposed changes in this license amendment request have been reviewed and~

approved by both On site and Off site Review in accordance with Comed procedures. A detailed description and a safety analysis of the proposed changes are presented in Attachment A. The proposed changes to the TS are presented in Attachment B 1 for Ilyron and Attachment 112 for llraidwood. Comed has reviewed this proposed license amendment in accordance with 10 CFR 50.92(c) and had determined that no significant hazards consideration exists. This evaluation is documented in Attachment C. An Environmental Assessment has been completed and is contained in Attachment D.

Comed is notifying the State ofillinois ofits application for this license amendment by copy of this letter and its attachments to the designated State Omcial.

Comed respectfully requests the NRC review and approve this license amendment request by February 7,1998, to support outage scheduling activities at both Ilyron and 11raidwood.

To the best of my knowledge, the statements contained in this document are true and correct.

Please address any comments or questions regarding this matter to D.J, Chrzanowski, Nuclear Licer. sing Administrator, at 630/663-7205.

Sincerely, A SNrv,

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John B. Ilosmer Engineering Vice President Attachments cc:

A.11. Beach, Regional Administrator Rill G. Dick, Byron /Ilr:idwood Project Manager-NRR C. Phillips, Senior Resident inspector-Braidwood Sen:or Resident inspector Ilyron Omce of Nuclear Safety IDNS l

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, STATE OF ILLINOIS

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COUNTY OF DUPAGE

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IN Tile MA' ITER OF

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COMMONWEALTil EDISON (COMED) COMPANY

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Docket Numbers BYRON STATION UNITS I & ?,

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50-454 & 50-455 AND

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BRAIDWOOD STATION UNITS 1 & 2

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50-456 & 50-457 AFFIDAVIT I affirm that the content of this tiansmittal is true and correct to the best of my knowledge,information and belief.

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John D. Ilosmer Engineering Vice President Subscribed and swom to before me, a Notary Public in and 0~OFMid}'i[^~~~^^

7 E O JAc ust Ne T e mouamouc sun o,v4Ns for the State above named, this day of p

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DESCRIPTION OF THE PROPOSED CHANGE Comed proposes to revise TS Surveillance Requirement 4.6.1.1.c,4.6.1.2 a,4.6.1.2 b, and the Bases to note that requirements are modified by approved exceptions. The 11ases for 3/4.6.1.2 will describe the approved exceptions. Specifically, Unit 2 is exempt from the requirement in NEl 94 01 to perform two consecutive successful Type A tests prior to i

extending the testing interval until September 10,2003 for Byron and November 9,2004 for Braidwood. The specific changes are shown in Attachments B land B 2 for Byron and Braidwood, respectively.

I F.

BASES FOR THE PROPOSED CHANGE Comed believes that Type A test requi ements for Byron Unit 2 and Braidwood Unit 2 have been demonstrated through prior testing and maintenance practice. In particular:

IypsATuthtfonnacclibion The first Byron Unit 2 Type A test following the successful pre operational test was conducted in September 1990. During this test, a steam generator manway i

leaked. The leakage was isolated by closing the Main Steam Isolation Valve and i

pressurizing the secondary side of the steam generator to approximately one psig below the containment test pressure. Initially the test was considered successful, since the steam generator pressure is greater than containment pressure expected during post accident conditions. The manway leak was not considered a post-accident leakage path, llowever, following consultation with Nuclear Reactor Regulation, Region til informed Comed that the manway leakage should be treated as valve leakage, and the as found test was considered a failure The next Type A test, conducted in September 1993, was completed successfully on both an as found and a performance basis.

The first Braidwood Unit 2 Type A test following the successful pre-operational test was conducted in November 1991. During the operating cycle preceding the test, a steam generator secondary manway leak deseloped. During this test, the leaking steam generator manway accounted for the majority of the leakage observed. The test was also performed in the as-found condition, so no repairs were affected prior to the test. This leakage was isolated by pressurizing the secondary side of the steam generator to approximately one psig below the containment test pressure. The steam generator pressure was continuously monitored to ensure it was maintained below the containment test pressure. This step was done to ensure that air was not introduced into the containment during the test, thus masking other leakage The test was then successfully completed However, as detailed in inspection Report 50-456/91023; 50 457/91021, instrument tolerance for the direct reading pressure gauge utilized to maintain the steam generator pressure was not considered; therefore, the test was consetvatively classified as a failure on both an as-found and performance basis.

The second Type A test, conducted in November 1994, was completed successfully on both an as fot nd and a performance basis.

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Additionaunfoamttior The steamlines leasing the containment are classified as General Design Criterion

$7 penetrations and are, therefore, not subject to Type C localleak rate testing.

This is because the secondary side of the steam generators is not considered part of the reactor coolant pressure boundary nor is there direct communication with the containment atmosphere The secondary manways of the steam generators are periodically removed during refueling outages to support such maintenance activities as inspections and sludge lancing. At the end of each refueling outage, a walkdown of components inside containmcal is performed as the unit approaches the normal operating parameters of $57 'F and 2235 psig in the Reactor Coolant System. Under these conditions, the secondary side of the steam generators is pressurized to greater than 1000 pdg This walkdown identifies any manway leakage prior to commencement of power operation. Leakage would be assessed and repaired as needed prior to returning the unit to service.

During operation, containment parameters are routinely monitored. Any leakage from secondary manways would condense and be transported to the Centainment floor drain sump via the floor drain system. This sump is equipped with a weir plate which is capable of detecting a one gallon per minute increase in sump input within one hour. This ensures that any unexpected input is promptly identified and assessed This assessment would identify ifleakage from the secondary side of the steam gencruors has developed.

The potentia for a leaking steam generator secondary side manway to become a signilicant c antainment bypass path is extremely low for the following reasons:

Leakage into the steam generator secondary does not have a direct path to tie outside, since the steam lines will isolate in a DBA LOCA scenario. Therefore one has to postulate failure of SG safety valves or pORVs in order to achieve any bypass at all. Since these valves would not be operating at pressures low enough to permit inleakage from the containment their failure in the open position can be considered a very remote possibility.

The physical behavior of the steam generator secondary in a post-LOCA situation will also mitigate against inleakage from the containment. In a DBA LOCA, the steam generators willisolate rapidly due to the containment high pressure signal.

Depressurization will be elatively slow and will occur as a result of ECCS flow through the steam generator primary side transiting towards the break. Cooldown and depressurization of the steam generators (in particular the one of the faulted loop) will occur gradually as the recirculated containment sump water, cooled by the RilR heat exchangers, is injected to the primary, picking up core decay heat, and then flowing through the u-tubes. The containment atmosphere,in contrast, is being directly cooled by the action of the RCFCs and containment spray. As a result, the containment pressure can be expected to decrease to near ambient conditions long before the fauhed loop steam generator depressurizes.

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For small break LOCA situations, the primary side pressure is insically set by the steam ger.crator pressure and controlled by the lowest oct safety valve. Steam Generator depressurintion will take even longer, while containment pressure is lower and w.li decrease faster than in the large break scenarios The ofTsite dose calculations are based on technical specification leakage rates for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and then use 50% of the technical specification leakage afler that This represents a very significant conservatism relative to the actual containment pressures predicted for the DB A LOCA scenarios Therefore, it c n be conduded that a leaking str.am generator manway does not represent a viable comainment leakage pathway since it lacks a Psth to the environment and the necessary pressure potential.

As discussed above, Comed believes the testing history and maintenance practices

' (both past and present) demonstrate the intent and purpose of Type A tests are

'v ing met and extending the frequency for the test on the afTected units will not represent an undue risk to health and safety of the public.

The one-time exception to Section 9.2.3 of the NEl document ends on September 10,2003 for Byron Unit 2 and November,2004 for Braidwood Unit 2. These n

dates coirespond to the end of the 10 year interval following the last successful Type A tests. The results of the next Type A test for Byron Unit 2 and Braidwood Umt 2 will determine the interval for subsequent tests. If that test is successful, there will have been two consecutive successful tests, and the test interval may be extended to ten years, in accordance with NEl 94 01 If the test is not successful, CcmEd v.ill be required to perform the next test within 48 months, in accordance with NEl 94 01.

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PROPOSED SAFETY ANAIMSIE OFTilC PROPOSED CIIANGE The function of the containment is to maintain stmetural integrity and leaktightness following a LOCA to ensure tht the release of nlicactive materials from the containment atmosphere will be restricted to those leak paths and associated leak rates assumed ir. the accident analysis. The limitations on containment leakage rates ensure that the values assiimed in the accident analysis are not exceeded at peak accident pressures Since the proposed change doe

• not alter the plant design, only the test frequency, there is no physical impact on containment stmetural integrity and therefore no direct increase in containment leakage. Ahhough decreasing the test frequency can increase the probability that an increase in containment leakare could go undetected for an extended period of time, the risk resulting from this proposed change i, inconsequential as iocumented in NUREG-1493," Performance Dned Containmut Leak Test Program."

' hiaintenance practices include visual inspection of manways for leakage (Vf 2), and inspdion of seating surfaces and gast ets for scratches, corrosion. tamage, and steam cuts ( tso, Quality Control establishes hold points for propc4 gasket align. ment and final manuay bolt torqueing sequence.

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The purpose of the testing requirements of 10 CFR 50, Appendix J is to assure that (a) leakage through the primary reactor containment and systems and components penetrating containment does not exceed allowable leakage rate values as specified in the TS or associated Hases and (b) proper maintenance and repairs of the reactor containment penetrations and isolation valves are made during the sersice life of the containment.

Each of these requirements has been demonstrated both through prior Type A testing and maintenance practices Even though the initial Type A tests at Hyron and Braidwood Unit 2 were conservatively classified as failures, it was demonstrated that the leakage through the primary reactor containment and systems and components penetrating containment did not exceed allowable leakage rate values as specified in the Technical Specifications or associated Bases. A successful Type A Test performed later further demonstrated the integrity of the containment and underlying requirements continued to be met. Existing surveillance activities such as steam generator manway installation procedures, walkdown of components inside containment prior to commencement of powet operation, and sump monitcring for leakage also support these requirements. Additionally, the leakage mechanism responsible for the only failure at both units, i e., secondary manway leakage, is addressed on an 18 month frequency. Therefore, the probability for undetected leakage continues to be small.

In NUREG 1493," Performance Based Containment Leakage Test Program", the NRC concluded that there is minimal impact on the public health and safety when test intervals are extended. With the exception of the test frequency, the actual tests will not change.

Additionally, the proposed license amendment will not reduce the availability of systems and components associated with containment integrity required to mitigate accident conditions. Containment leakage rates, parameters and accident assumptions would not be affected by the proposed license amendment.

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The preceding discussion demonstrates that the underlying purpose of the m!: can be met without the performance of an Type A test during the next refueling outage for each unit, j

and that the health and safety of the public will continue to be protected.

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IMPACTON PREVIOUS SUBMITTALS-

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There is no impact on any presious submittals.

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SCHEDULE REQUIREMENTS Comed respectfully requests that the NRC review and approve this license amendment i

request by February 7,1998, to support outage scheduling activities at Byron and Braidwood.

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