ML20069F332

Application for Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,revising TS 4.4.5.2, SG Tube Sample Selection & Insp, 3.4.6.2, RCS Operational Leakage & Associated Bases

ML20069F332
Person / Time
Site:	Byron, Braidwood
Issue date:	06/03/1994
From:	Saccomando D COMMONWEALTH EDISON CO.
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20069F335	List: ML20069F332 ML20069F340
References
NUDOCS 9406080217
Download: ML20069F332 (20)

Text

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Commanwrith Edison

1. 1400 Opus Place C Downers Grove, Illinois 60515 June 3,1994 Mr. William T. Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 P

Attention: Document Control Desk

Subject:

Application for Amendment to Facility Operating Licenses:

Byron Station Units 1 and 2 i (NPF-37/66; NRC Docket Nos. 50-454/455) ,

Braidwood Station Units 1 and 2 (NPF-72/77; NRC Docket Nos. 50-456/457)

" Steam Generator Tube Sleeving Followup Commitments"

References:

See Attached l

Dear Mr. Russell:

Reference 1 transmitted Commonwealth Edison Company's (CECO) request to amend - '

Technical Specification 3/4.4.5, " Steam Generators" which permits the sleeving of the steam generators tubes. Subsequent to that amendment request, the Nuclear Regulatory Commission (NRC) transmitted Reference 2 (See Attachment 5) requesting CECO to commit to a near term license amendment to address conditions associated with the approval of the Westinghouse and Babcock and Wilcox Nuclear Technologies steam generator sleeves. Four conditions were specified in the letter:

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1. Amend the Byron and Braidwood licenses to reflect a primary-to-secondary leakage rate limit of 150 gallons per day (gpd). t

2. Amend the Byron and Braidwood licenses to reflect an inservice inspection of a  ;

minimum of 20 percent of a random sample of the sleeves for axial and circumferential indication at the end of cycle.  !

3. Add a condition to the Byron and Braidwood licenses to conduct additional  !

corrosion testing to establish the design life for the kinetically or laser welded sleeved tubes in the presence of a crevice. ,

4. Perform post weld heat treatment at 1400 F minimum soak temperature with a 4 5 minute minimum soak time on freespan kinetically or laser welded joints until additional supporting data becomes available. ,

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Mr. Russell June 3,1994 In Reference 3 (See Attachment 6), CECO committed to these conditions and stated  ;

that the proposed license amendments to incorporated cor-litions 1,2, and 3 would be issued within 90 days from the issuance of the sleeving amendment. The approved amendment and Safety Evaluation Report (SER) was subsequently issued on March .

4,1994, and transmitted in Reference 4. l To ensure completion of these commitments, pursuant to 10 CFR 50.90, CECO ,

proposes to amend Facility Operating Licenses NPF-37, NPF-66, NPF-72, and NPF- i 77 and to amend the associated Appendix A, Technical Specifications. The proposed amendment request revises Technical Specification 4.4.5.2 and 3.4.6.2, " Steam '

Generators", and the associated bases.

The amendment request is subdivided as follows: ,

Attachment 1: Description and Safety Analysis of Proposed Changes Attachment 2: Proposed Revision to the Operating Licenses and Technical l Specifications for Byron and Braidwood Stations Attachment 3: Evaluation of Significant Hazards Considerations ,

Attachment 4: Environmental Assessment The proposed changes have been reviewed and approved by the On-site' and Off-site  !

Review Committees in accordance with CECO procedures. CECO has reviewed this  :

proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists.  ;

CECO is notifying the State of Illinois of our application for these amendments by transmitting a copy of this letter and the associated attachments to the designated ,

State Official. ,

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Mr. Russell June 3,1994 i To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my l personal knowledge, but on information furnished by other CECO employees, contractor employees, and/or consultants. Such information has been reviewed in  :

accordance with company practice, and I believe it to be reliable.  !

Please address any comments or questions regarding this matter to this office. {

Sincerely, i 7

x i Denise Sa comando Nuclear Licensing Administrator i

Attachments cc: G. F. Dick, Byron Project Manager - NRR R. R. Assa, Braidwood Project Manager - NRR l H. Peterson, SRI - Byron i S. G. Dupont, SRI - Braidwood i B. Clayton, Branch Chief - Region til  !

Office of Nuclear Facility Safety - IDNS 6 /993 M

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MY COMMISSION ExF*  ;

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REFERENCES l

1. J. A. Bauer letter to T. E. Murley dated August 13,1993, transmitting '

Application for Amendment to Facility Operating Licenses for Byron and 4 Braidwood Stations pertaining to Steam Generator Tube Sleeving Methodology

2. R. R. Assa letter to D. L. Farrar dated February 22,1994, requesting that  :

CECO commit to a near term license amendment to address conditions associated with the approval of Westinghouse and BWNT Steam Generator Sleeves

3. D. M. Saccomando letter to T. E. Murley dated February 24,1994, committing to the conditions specified in Reference 2 above r

4. R. R. Assa letter to D. L. Farrar dated March 4,1994, transmitting Amendment 58 (Byron) and Amendment 46 (Braidwood) and associated SER approving the use of Westinghouse and BWNT sleeves i

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ATTACHMENT 1 DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 AND APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES '

NPF-37, NPF-66, NPF-72, AND NPF-77 De.acription of the Proposed Chanaejs ,

Commonwealth Edison Company (CECO) proposes to amend Technical Specification (TS) 3.4.5 and 3.4.6.2, and facility operating licenses NPF-37, NPF-66, NPF-72 and NPF-77 to include the conditions contained in the Nuclear Regulatory Commission's (NRC) Safety Evaluation Report (SER) for TS Amendment 58 (Byron) and TS .

Amendment 46 (Braidwood), " Steam Generator Sleeving", issued March 4,1994, (TAC NOS. M87227, M87228, M87229, and M87230).

In a J. A. Bauer letter to T. E. Murley dated August 13,1993, CECO submitted an i amendment request for Byron and Braidwood Stations proposing to perform steam generator tube sleeving in accordance with the Westinghouse and B&W processes.

The NRC Technical Staff found the proposed amendment request acceptable subject to certain conditions. These four conditions were discussed in a teleconference on February 18,1994 and subsequently documented in a R. R. Assa letter to D. L. Farrar dated February 22,1994 (Attachment 5). Byron and Braidwood found these four contingencies acceptable and formally committed to address these conditions as documented in a D. M. Saccomando letter to T. E. Murley dated February 24,1994 (Attachment 6). Upon CECO's acceptance of the specified conditions, the SG sleeving processes addressed in the amendment were approved for Byron and Braidwood, r

The conditions specified in R. R. Assa's letter dated February 22,1994, are restated below:

1. Amend the license to reflect a primary-to-secondary leakage rate limit of 150 gpd.  ;

2. Amend the license to reflect an inservice inspection of a minimum of 20% of a .

random sample of the sleeves for axial and circumferentialindications at the end of each cycle. In the event that an imperfection of 40% or greater depth is detected, an additional 20% (minimum) of the unsampled sleeves should be l inspected, and if an imperfection of 40% or greater depth is detected in the second sample, all remaining sleeves should be inspected. The inservice b nlatbyron:slevfol:2

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Attachment 1 Page 3  !

5 inspection is required until the licensee demonstrates the corrosion resistance for laser-welded or kinetically welded joints in tubes that bound the material properties of the tubes installed in the Byron and Braidwood Steam Generator.  !

If conformance with the requirements of the plant TS for tube structuralintegrity is not confirmed, the tubes containing the sleeves in question shall be removed

• from service.  !

3. Add a condition to the license to conduct additional corrosion testing to establish the design life for the kinetically or laser welded sleeved tubes in the presence of a crevice. The testing should determine the effects that material microstructure, chemistry, and joint crevices will have on primary water stress corrosion cracking (PWSCC) initiation and growth, bound the material condition that exist in Byron and Braidwood steam generators, and include the  !

associated stress intensity values.

4. Perform post weld heat treatment at 1400 F minimum soak temperature with a- l 5-minute minimurr. soak time on freespan kinetically or laser welded joints until l additional supporting data becomes available. i Conditions 1 and 2 are being addressed by amending Technical Specifications 3.4.5, t end 3.4.6.2. Condition 3 will be addressed by adding a license condition to Operating l Licenses NPF-37, NPF-66, NPF-72, and NPF-77. Condition 4 will be addressed by j making appropriate sleeving installation procedure changes to address the PWHT  !

requirements and therefore is not addressed in this submittal. i The applicable Technical Specification and License pages, with the changes indicated, are included in Attachment 2 of this submittal. 1 i

1. Amend the License to Reflect a Primary-to Secondary Leakage Rate Limit of 150 gpd Description and Bases of the Current Reauirements '

For Byron Units 1 and 2, and Braidwood Unit 2, Technical Specification l 3/4.4.6.2 c " Operational Leakage", limits the total reactor-to-secondary leakage  ;

to 1 gallon per minute (gpm) through all steam generators not isolated from the Reactor Coolant System (RCS) and 500 gallons per day (gpd) through any one  ;'

steam generator (SG).

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f Attachment 1 i Page 4 Please note that for Braidwood Station Unit 2, TS 3.4.6.2. c " Operational Leakage", limits the total reactor-to-secondary leakage to 1 gallon per minute  !

through all steam generators not isolated from the Reactor Coolant System l (RCS) and 500 gpd through any one steam generator. This specification was  !

revised from 1 gallon per minute (gpm) and 500 gpd respectively by the  :

Braidwood Unit 1 Interim Plugging Criteria (IPC) Amendment (Amendment 50) i issued May 7,1994. The proposed change to operational leakage, noted below, is not applicable to Braidwood Unit 2 since this license change has already been approved. ,

The primary-to-secondary leakage is maintained by limiting the leakage to 500 gpd through any one steam generator and a total of 1 gpm through all unisolated steam generators. This limit ensures that the dosage contribution from SG tube leakage will be limited to a small fraction of the 10 CFR 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. Cracks having a primary-to-secondary leakage less than this limit  :

during operation wil! have an adequate margin of safety to withstand the loads l imposed during normal operation and by postulated accidents. The 1 gpm limit ,

is consistent with the assumptions used in the analysis of these accidents. The r 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line break or under loss of

• coolant accident (LOCA) conditions. Operating plants have shown that the i 500 gpd leakage limit per steam generator can be readily detected by the steam generator blowdown radiation monitors.  ;

Description and Bases of the Requested Revision Specification 3.4.6.2, Reactor Coolant System Ooerational Leakage NOTE: This proposed change is applicable to Byron Units 1 and 2, and Braidwood Unit 2.

The RCS leakage limit of 500 gpd for primary-to-secondary leakage through  !

any one steam generator has been reduced to 150 gpd. This reduction requires the total leakage limit of 1 gpm through all steam generators not .

isolated from the RCS to be limited to 600 gpd.

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Attachment 1 Page5 i

Bases for 3/4.4.5. Steam Generators j References to the reactor-to-secondary leakage of "500 gpd" per steam generator, in the second paragraph of this bases section, are being changed to .

"150 gpd". In addition, the radiation monitors, specified in the second i paragraph, that have proven effective in identifying SG tube leakage, have also ,

been changed.

1 These changes read as follows:

"The extent of cracking during plant operation will be limited by the limitation.

of steam generator tube leakage between the RCS and the Secondary Coolant System (reactor-to-secondary leakage = 150 gpd per steam  !

generator)." (NOTE: This change has already been approved for Braidwood  !

Unit 1 and therefore is not applicable to the Braidwood Unit 1 license). l

" Operating plants have demonstrated that reactor-to-secondary leakage of 150 gpd per steam generator can readily be detected by the radiation {

monitors of steam generator blowdown, main steam lines, or the steam jet 1 air ejectors." (NOTE: The addition of the radiation monitors is also  !

applicable to the Braidwood Unit 2 license). j i

Bases for 3/4.4.6.2. Ooerational Leakage The third paragraph of this bases section discusses steam generator tube ,

leakage. The total steam generator tube leakage limit of "1 gpm" for all steam-generators is changed to "600 gpd". The "500 gpd" leakage limit per steam .

generator is changed to "150 gpd". The sentence that formerly read: "The 1  ;

gpm limit is consistent with the assumptions used in the analysis of these accidents", has been enhanced for clarification purposes. ,

s Please note that in the Braidwood Unit 1 Interim Plugging Criteria Amendment -

(Amendment 50) issued May 7,1994, the sentence addressing the accident analysis assumptions had been changed to read, "The 600 gpd limit is -

consistent with the assumptions used in the analysis of these accidents." This i sentence will also be changed as noted below for consistency and clarification  !

purposes.

i The third paragraph of the basis will be changed to read as follows: l "The total steam generator tube leakage limit of 600 gpd for all steam i generators not isolated from the RCS ensures that the dosage contribution  :

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Attachment 1 Page 6 from the tube leakage will be limited to a small fraction of the 10 CFR Part 100 dose guideline values in the event of either of steam generator tube rupture or steam line break. A 1 gpm total steam generator tube leakage limit for all steam generators and a 500 gpd limit per steam generator were the assumptions used in the analysis of these accidents in Chapter 15 of the ,

UFSAR. The assumptions in Chapter 15 remain valid since the leakage limitations implemented for total and individual steam generator leakages are more conservative that those used for the analysis. The 150 gpd leakage-limit per steam generator ensures that steam generator tube integrity shall be ,

maintained in the event of a main steam line break or under LOCA  ;

conditions."

Historically, SG sleeve-to-tube joints were basically mechanical seals and were j not considered leak-tight. The NRC has not required that sleeve repairs be leak-tight but only leak-limiting. The staff evaluated leakage based on plant-  ::

specific TS requirements for primary-to-secondary leakage limits under normal and accident conditions. CECO has analyzed the leakage effects of the HEJ, the HEJ with laser-welded joint, the freespan laser-welded joint and kinetically  !

welded joint. The analyses show that even under extreme postulated _  ;

conditions, the three joints will maintain satisfactory integrity against leakage.

The continuous circumferential laser welded joints are inherently leak-tight.

I Degraded tubes that were restored to operation as a result of sleeving are susceptible to additional degradation in'the same SG environment. The sleeve  ;

is designed to extend past the welded joint and into the tubing. In the event  !

that a sleeved tube fails'near or at the weld, the sleeve extension will restrict ,

tube movement and leakage. Leakage monitoring devices are intended to alert j plant personnel to implement the appropriate procedures. However, based on i experience with various causes of leakage through tubes including experience related to tubes repaired by sleeving, the staff has concluded that the current  ;

I primary-to-secondary leakage limits in the Byron and Braidwood TS are not sufficient to detect early stages of sleeve degradation. To reach a satisfactory conclusion regarding the acceptability of the sleeving application, particularly in view of corrosion issues, the staff requested that the license be amended to reflect a primary-to-secondary leakage limit of 150 gallons per day per steam ,

generator.

Impact of the Proposed Chanoe The original amendment requested tubesheet sleeves and tube support plate sleeves be approved as an alternate tube repair method for Byron and i i

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Attachment 1 Page 7 Braidwood Units 1 and 2. The steam generator sleeves approved for installation use the Westinghouse process (laser welded joints) or the Babcock and Wilcox (B&W) process of kir'etically welded joints. The sleeve configuration was designed and analyzed in accordance with the criteria of Regulatory Guide (RG) 1.121 and the design requirements of Section ill of the American Society of Mechanical Engineers (ASME) Code. Fatigue and stress analyses of the sleeved tube assemblies for both processes produced acceptable results as documented in the Westinghouse and B&W topical reports submitted in the original sleeving package. Mechanical testing has shown that the structural strength of the sleeves under normal, faulted, and upset conditions is within acceptable limits. Leakage rate testing for the tube sleeves has demonstrated that primary-to-secondary leakage is not expected during all plant conditions. Any leakage through the sleeved region of the tube is fully bounded by the leak-before-break considerations and ultimately the existing steam generator tube rupture analysis included in the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR).

Reduction of the leakage rate requirement from 500 to 150 gpd per SG will continue to ensure SG tube integrity is maintained in the event of a Main Steam Line Break (MSLB) or under LOCA conditions. The reduction to 150 gpd also limits the allowable primary-to-secondary leakage from 1 gpm to 600 gpd for all SGs not isolated from the RCS. This previous leakage limit, used in the UFSAR accident analysis, ensured the dosage contribution from tube leakage would be limited to a small fraction of the 10 CFR 100 dose guideline values in the ewnt of either a steam generator tube rupture or MSLB.

l Reducing these limits does not result in a reduction of any safety margins.

2. Inservice inspection of Sleeves Description and 63ases of the Current Requirements The current surveillance requirements for inspection of the steam generator tubes ensure that the structuralintegrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain l surveillance of the conditions of the tubes in the event that there is evidence of I

mechanical damage or progressive degradation due tc design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

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Attachment 1 Page 8 Rescription and Bases of the Requgated Revision Soecification 4.4.5.2, Steam Generator Tube Samole Selection and IrSpection NOTE: This proposed change is applicable to Byron Units 1 and 2, and Braidwood Units 1 and 2.

Section of 4.4.5.2 d has been added to the specification and reads as follows:

• A random sample of at least 20 percent of the total number of sleeves shall be inspected for axial and circumferential indications at the end of each cycle. In the event that an imperfection of 40 percent or greater depth is detected, an additional 20 percent of the unsampled sleeves shall be inspected, and if an imperfection of 40 percent or greater depth is detected in the second sample, all remaining sleeves shall be inspected. These inservice inspections will include the entire sleeve and the tube at the heat treated area. The inservice inspection for the sleeves is required untii the corrosion resistance for the laser welded or kinetically welded joints in tubes that bound the material parameters of the tubes installed in the steam generators has been demonstrated acceptable. If conformance with the acceptance criteria of section 4.4.5.4 for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service."

The surveillance requirements for inspection of the SG tubes ensure that the structuralintegrity of this portion of the RCS is maintained. The program for inservice inspection of SG tubes is essentialin maintaining a surveillance on the condition of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of the steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The growth rate of previously identified tube degradation will be monitored and new imperfections identified, documented, and monitored for growth rates over the upcoming cycles.

Impact of the Proposed Chanog The sleeve sample size has been increased from three (3) to twenty (20) percent during each scheduled inservice inspection. Increasing the random sample size of the sleeves to be inspected willincrease the monitoring of k:nla: byron:slevfol:8

Attachment 1 Page 9 those tubes for any further degradation while remaining inservice if the sample identifies a sleeve with an imperfection of greater that 40 percent depth, an additional 20 percent of the sleeves shall be inspected. The sleeves that have identified imperfections of greater than 40 percent shall be evaluated and removed from service. The inservice inspections and additional corrosion testing for the sleeves and welded joints will continue until the corrosion resistance is demonstrated acceptable to the NRC. If conformance with the ,

acceptance criteria of section 4.4.5.4 for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.

3. License Condition to Address Corrosion Testing Description and Bases of the Current Requirements There are currently no license conditions to conduct additional corrosion testing for steam generator sleeves. Corrosion testing of the Westinghouse and B&W sleeves was however performed as described in the technical reports submitted in support of the original sleeving license amendment.

Qgacription and Bases of the Reauested Revision Facility Operating License No. NPF-37Jart 2.C (Pvron Unit 1)

A license condition (item 16) will be added to the Unit 1 license to specify that a corrosion testing program, to establish sleeve design life and corrosion resistance to confirm tube structuralintegrity, will be performed for Byron. If the tube structural integrity is not confirmed or the corrosion testing program is not accepted by the NRC within the next two fuel cycles, the tubes containing the sleeves in question shall be removed from service.

The NRC staff and CECO concur that additional corrosion testing is needed to establish the design life for the sleeved tubes in the presence of the crevice created at the joint between the sleeve and the tube. The test data should determine the effects that material microstructure, chemistry, ano joint crevices will have on primary water stress corrosion cracking initiation and growth. This testing program should be conducted for the sleeving processes installed for Byron Unit 1.

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Facility Ooeratina License No. NPF-66 cart 2.C (Bvron Unit 2) ,

A license condition (item 5) will be added to the Unit 2 license to specify that a-corrosion testing program, to establish sleeve design life and corrosion resistance to confirm tube structural integrity, will be performed for Byron. ' If  ;

the tube structural integrity is not confirmed or the corrosion testing program is not accepted by the NRC within the next two fuel cycles, the tubes containing j the sleeves in question shall be removed from service.

The NRC staff and CECO concur that additional corrosion testing is needed to f establish the design life for the sleeved tubes in the presence of the crevice created at the joint between the sleeve and the tube. The test data should l determine the effects that material microstructure, chemistry, and joint crevices '!

will have on primary water stress corrosion cracking initiation and growth. This testing program should be conducted for the sleeving processes installed for ,

Byron Unit 2.

Facility Ooeratino License No. NPF-72 oart 2.C (Braidwood Unit 1)  ;

A license condition (item 6) will be added to the Unit 1 license to specify that a l corrosion testing program, to establish sleeve design life and corrosion .i resistance to confirm tube structural integrity, will be performed for Braidwood. j If the tube structural integrity is not confirmed or the corrosion testing program  ;

1s not accepted by the NRC within the next two fuel cycles, the tubes  !

containing the sleeves in question shall be removed from service.

The NRC staff and CECO conce that additional corrosion testing is needed to establish the design life for the sleeved tubes in the presence of the crevice I created at the joint between the sleeve and the tube. The test data should . .i determine the effects that material microstructure, chemistry, and joint crevices i will have on primary water stress corrosion cracking initiation and growth. ' This j testing program should be conducted for the sleeving processes installed for l Braidwood Unit 1.

i Facility Ooeratina License No. N 7F-77 oart 2.C (Braidwood Unit 2)  ;

A license condition (item 5) will be added to the Unit 2 license to specify that a l corrosion testing pregram, to establish sleeve design life and corrosion resistance to confirm tube structuralintegrity, will be performed for Braidwood.

If the tube structuralintegrity is not confirmed or the corrosion testing program is not accepted by the NRC within the next two fuel cycles, the tubes containing the sleeves in question shall be removed from service.

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Attachment 1 Page 11 The NRC staff and CECO concur that additional corrosion testing is needed to establish the design life for the sleeved tubes in the presence of the crevice created at the joint between the sleeve and the tube. The test data should determine the effects that material microstructure, chemistry, and joint crevices will have on primary water stress corrosion cracking initiation and growth. This testing program should be conducted for the sleeving processes installed for Braidwood Unit 2. ,

impact of the Proposed Change The addition of a license condition to conduct additional corrosion testing to establish the design life for the kinetically or laser welded sleeved tubes will enhance the confidence that sleeves will maintain integrity for extended operations. The corrosion testing program will be conducted in a laboratory environment and will not in any way adversely affect plant operations or safety.

Schedule Requirements This license amendment is being submitted as requested in the Safety Evaluation Report for Amendment 58 (Byron) and Amendment 46 (Braidwood) dated March 4, 1994. Byron and Braidwood Stations formally committed to address the conditions in the SER in a D. M Saccomando letter to T. E. Murley dated February 24,1994.

Administrative controls are already in effect at Byron and Braidwood which address the proposed steam generator leak rate limits. Since no steam generator sleeves are currently installed at Byron or Braidwood, inservice inspection of the sleeves will not become an issue until completion of Byron Unit 1 Cycle 6 (spring 1996), assuming sleeves are installed in Byron Unit 1 during the fall 1994 outage. The corrosion testing program is currently under development. Based on this information, CECO respectfully requests that this amendment be approved prior to February,1996.

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ATTACHMENT 2 PROPOSED CHANGES TO FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72 AND NPF-77 Revised Pages: NPF-37 pg 5 NPF-66 pg 2 NPF-72 pg 3 NPF-77 pg 3 PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS FOR FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72 AND NPF-77 Revised Pages: 3/4 4-13*

3/4 4-14 3/4 4-21 B 3/4 4-3 B 3/4 4-4 Pages indicated with and asterisk (*) are not being revised and are only included for convenience.

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NPP-37 1

(Continued)

(15) Use of advisors who were licensed only at the RO level will be evaluated on a case-by-case basis. Advisors shall be trained on plant procedures, technical specifications and plant systems, and shall be examined on these topics at a level sufficient to assure familiarity with the plant. For each shift, the remainder of the

' shift crew shall be trained as to the role of the advisors.

These advisors shall be retained until the experience levels identified in the first sentence above have been achieved. The NRC shall be notified at least 30 days prior to the date that the m f-licensee proposes to release the advisors from further service.

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D.

The A, E facility and J torequires 10 CFRexemptions from certain requirements of Appendices Part 50. These I include (a) an exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J. the testing of containment air locks at times when containment integrity is not required (Section 6.2.6 of the SER), (b) an exemption from GDC-2 of Appendix A, the requirement that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes (section 3.10 of SSER #5), (c) an exemption from GDC-13 and GDC-11 of Appendix A, the requirement that instrumentation be provided to monitor variables and systems over their anticipated ranges, and the requirement that provisions be included to  ;

i minimize the probability of losing electric power (Section 9.5.4.1 of SSER #5),

(d) an exemption from GDC-19 of Appendix A, the requirement t that the control room have adequate radiation protection to permit access and occupancy under accident conditions (Section 6.5.1 of SSER

1. 6), and (e) an exemption from the requirement of Section IV.F of Appendix E that a full participation emergency planning exercise be conducted within one year before issuance of the first operating license for full power and prior to operation above 5% of rated power (Section 13.3 of SSER #6). These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of ,

these exemptions the facility will 91erata, to the extent authorized l herein, in conformity with the 6p;.lication, as amended, the provisions of the Act, and the roics enu m3u1 Atioa of the commission.

E.

The licenace shall fully implement and maintain in effect all provision 3 of the Commission approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10CPR 73.55 (51 FR 27817 and )

27822) and to the authority cf 10CFR 50.90 and 10CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: " Byron Nuclear Power Station Security Plan," with revisions submitted through January 14, 1988; " Byron Puclear Power Station Security Personnel Training and Qualification Plan," with revisions submitted through September.26, 1986; and " Byron l'ucicar Power Station Safeguards Contingency Plan," with revisions ouhaitted through July 30, 1985.

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BYRON INSERT C (page 5 of NPF-37) t (16) Steam Generator Sleeving Corrosion Testing I The licensee shall conduct additional corrosion testing to establish the design i life for the kinetically or laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the kinetically or laser welded joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the ,

Beginning of-Cycle 9. If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.

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NPF-66 (5) CECO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulaticens set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

4 (1) Maximum Power Leve5 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein. The items be identified completed in Attachment I to this license shall as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Enecifications and Envire , ental Protection 1

Elan The Technical Specifications contained in Appendix A (NUREG-1113), as revised through h n h nt No. 61 and revised by Attachment 2 to NPF-60, 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 C accordance with the Technical Specifications and the Environmental Protection Plan. (Revised 04-18-94:

Am. 61)

(3) Initial -t Proorgs Any changes to the Initial Startup Test Program described in Chapter 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(4)

Reculatorv Guide 1.97. Revision 2 comnliance l The licensee shall submit by March 1, 1987, a preliminary report describing how the requirements of Regulatory Guide 1.97, Revision 2 have been or will be met.

The licmsee shall submit by September 1, 1987, the final report and a schedule for implementation M

(assuming the NRC approves the DCRDR by March 1, 1987).

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l BYRON INSERT D (page 2 of NPF-66) i (5) Steam Generator Sleeving Corrosion Testing The licensee shall conduct additional corrosion testing to establish the design life for the kinetically or laser welded sleeved tubes in the presence of a  ;

crevice. The corrosion testing shall demonstrate the corrosion resistance for  ;

the kinetically or laser welded joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the  :

Beginning-of-Cycle 8. If conformance with the requirements of the plant [

Technical Specifications for tube structural integrity is not confirmed, the tubes  ;

containing the sleeves in question shall be removed from service. i I

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