ML20085N212
| ML20085N212 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/22/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085N203 | List: |
| References | |
| NUDOCS 9506300093 | |
| Download: ML20085N212 (7) | |
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UNITED STATES p
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20066 0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION' l
RELATED TO AMENDMENT NO. 72 TO FACILITY OPERATING LICENSE NO. NPF-37, AMENDMENT NO. 72 TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. NPF-72, AND AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS. 1 AND 2 I
BRAIDWOOD STATION. UNIT NOS. I AND 2 p_0CKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 50-457 0
1.0 INTRODUCTION
By letter dated May 20, 1994, Commonwealth Edison Company (Comed, the licensee) submitted a request to amend the Technical Specifications (TSs) as they apply to Byron Station, Unit 1, and Braidwood Station, Unit 1.
The i
application aas revised by letter' dated February 2,1995. The amendment j
requests proposed an alternative repair criteria for defects found in the steam generator tube expansion region within the tubesheet. Under the requests, steam generator tubes with degradation in excess of the current plugging limits could remain in service without repair provided the indications exist below a specified distance (F* (F-star)), from the secondary face of the tubesheet or the top of last hardroll, whichever is further into i
the tubesheet. To support the licensee's request, Babcock & Wilcox Nuclear Technologies (BWNT) completed a qualification test program to demonstrate that the proposed F* distance satisfies the necessary structural and leakage integrity requirements of Appendix A to 10 CFR Part 50 and the plant TSs. The results of the test program (W-04 F* Qualification Report) were included with the May 20, 1994, request in both the proprietary version (BAW 10196P) and non proprietary version (BAW 10196).
Surveillance requirements within plant TSs require a periodic inspection of J
steam generator tubes for the detection of potential degradation (i.e.,
u cracks, dents, corrosion, etc.) which could diminish the structural margins
)
and leakage integrity of the tubes. Detection of tube degradation in excess of the TS limits requires a repair or removal of the tube from service. The licensee has proposed a revised repair criteria that would allow steam generator tube defects to remain in place without repair provided the defects reside a specified distance below the secondary face of the tubesheet.
This 9506300093 950622 PDR ADOCK 05000454 P
PDR l
l distance is called F*.
Degradation identified in a steam generator tube below the F* length would be allowed to remain in service without repair. This is based on the results of testing which determined the minimum interference fit engagement length necessary to retain steam generator tubes within the tubesheet.
A staff request for additional information (RAI) was sent to the licensee on November 3, 1994.
The licensee responded by letter dated December 2, 1994. A second RAI was sent on February 22, 1995, and the licensee responded by letter dated March 14, 1995.
The forgoing responses by the licensee provided clarifying information within the scope of the license amendment applications and did not affect the proposed no significant hazards consideration determination. The staff has reviewed the information supplied by the licensee and completed an evaluation of the licensee's request to amend the Byron, Unit 1, and Braidwood, Unit 1, TSs to include the F* criteria.
Byron, Unit 1, and Braidwood, Unit 1, were both constructed with Westinghouse 04 steam generators.
Because the steam generators are of the same design and 1
the accident loads imposed on the tubes are similar, this safety evaluation addresses the proposed F* amendment for both units.
2.0 BACKGROUND
Steam generator tubes comprise a significant portion of the reactor coolant pressure boundary.
Maintenance of this barrier is provided by the integrity of the steam generator tube wall and the tube-to-tubesheet connection. The connection between the tube and tubesheet is an interference fit made by roll expanding the tube into a bore through the tubesheet. The tubes were originally installed by first expanding them into the tubesheet followed by seal welding at the primary face.
Step rolls were then performed to fully expand the tube within the tubesheet.
The inelastically deformed steam generator tube is held in place by the elastic springback of the tubesheet.
Undegraded, the tube-to-tubesheet joint provides sufficient strength to maintain adequate structural and pressure boundary (leakage) integrity.
General Design Criteria (GDC) 14, " Reactor Coolant Pressure Boundary," and GDC 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," of Appendix A to 10 CFR Part 50 state the requirements applicable to maintaining adequate structural and leakage integrity for steam generator tubes.
Regulatory Guide (RG) 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," describes an acceptable method for establishing the limiting safe conditions of tube degradation of steam generator tubing.
In order to demonstrate adequate structural margin for degraded steam generator tubes, the bases for the proposed steam generator tube repair criterion must address the limiting conditions during normal operation, anticipated operational occurrences, and postulated accident conditions. The margin of failure under normal operating conditions as recommended in RG 1.121 should not be less than three at any tube location.
Subsection NB-3225 of Section 111 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) specifies the margins of safety under postulated accident conditions.
- Structural loads imposed on the steara generator tube-to-tubesheet connections primarily result from the differential pressure between the primary and secondary sides of the tubes.
The peak postulated loading occurs during a steam line break due to a lowering of the secondary side pressure. However, normal operating loads, cyclic joint loading from major plant transients (i.e., startup/ shutdown), and potential thermal expansion loads can also be significant.
The analysis (BAW-10196P) supporting the licensee's proposed amendment to the Byron, Unit 1, and Braidwood, Unit 1, TSs addressed the limiting conditions necessary to maintain adequate integrity of the tube-to-tubesheet interference fit.
Specifically, the tube must not experience excessive displacement relative to the tubesheet.
The elastic preload between the tube and tubesheet not only prevents pullout of the tube from the tubesheet, but also provides a leaktight barrier minimizing the potential for primary to secondary coolant leakage. With sufficient length of hardroll, the tube-to-tubesheet connection will not allow any leakage under normal and faulted conditions.
Steam generator tube through-wall degradation within the roll expanded joint decreases the path length necessary for primary to secondary leakage. The licensee's proposed amendment would permit tubes with such degradation to remain in service provided there exists a sufficient length of undegraded hardroll. Therefore, an acceptable F* distance must be such that leakage integrity is not jeopardized during all analyzed conditions.
Leakage through steam generator tubes is limited by plant TSs.
The acceptance criteria for the qualification test program established a maximum primary-to-secondary leak rate of I gpm through any one steam generator.
The maximum tolerable leakage rate through a single F* tube was determined assuming all F* tubes in the affected generator were leaking at the maximum rate.
3.0 EVALVATION 3.1 Testina to Determine F*
Babcock & Wilcox Nuclear Technologies completed a test program to determine the F* distance.
Two failure criteria were considered for testing -- the ultimate pullout load of the tube from the tubesheet and primary-to-secondary leakage requirements. The following describes the methodology used in the test program.
3.1.1 Fabrication of Test Specimens Mockup blocks were fabricated to simulate the actual installed conditions for the tubes within the Byron, Unit 1, and Braidwood, Unit 1, steam generators.
Lengths of steam generator tubing were roll expanded into holes drilled through the mockup blocks.
Several peripheral tubes were roll expanded in the test block to simulate additional constraint by surrounding tubes. The interior tubes were used for testing.
In order to simulate tube wall degradation, tubes were severed at a certain distance below the upper face of the mockup block.
Such a configuration is representative of a 360 degree through-wall crack present in the tube.
- Several mockup blocks were fabricated for testing. After the tubes were expanded into the blocks and the hardroll length was verified by nondestructive evaluation methods, some of the test blocks were thermally soaked to simulate the effects of actual steam generator service temperature.
Heating the test block could theoretically lead to thermal stress relaxation in the roll expansion joint.
To account for potential factors which might affect the calculated F* length, several variables were changed within the test matrix.
For example, tubing with both high and low yield strengths were tested.
In addition, tubesheet bore surface roughness, as well as tubesheet bore diameter were varied in the test matrix. The results from the tests revealed the effects from these variables.
In the qualification test program the effects of boric acid corrosion were considered.
If primary coolant penetrated through the steam generator tube wall and came in contact with the carbon steel tubesheet the potential exists for initiating stress corrosion cracking in the tubesheet.
Based on previous studies, the likelihood of developing significant corrosion of the tubesheet bore due to boric acid corrosion is low.
3.1.2 Testino for F* Determination To determine the necessary roll expansion joint engagement length, BWNT completed a series of mechanical tests on the simulated steam generator tubes.
The qualification testing used a combination of both internal pressure and axial loading to simulate the applied loads on the steam generator tubes to determine the F* distance.
Under service conditions, the differential pressure acting over the cross section of the tube induces an axial force tending to force the tube out of the tubesheet. This axial load is counterbalanced by the frictional force between the tube and tubesheet from the roll expanded interference fit. The extent of radial interference between the tube and tubesheet increases at operating conditions due to differential thermal expansion forces. This increases the strength of the joint.
The primary-to-secondary differential pressure acts to both increase and decrease the tube-to-tubesheet interference fit.
The two competing factors involve the outward expansion of the tube in the radial direction and the tubesheet bowing effect. The primary-to-secondary differential pressure slightly expands the tube in the radial direction strengthening the joint.
However, the higher primary side pressure tends to bow the tubesheet in the upward direction. The tubesheet bowing effect results in the dilation of the tubesheet bores decreasing the tube-to-tubesheet joint strength.
The net radial force from these phenomena affects the frictional force between the tube and tubesheet resisting pullout. The reduction in tube-to-tubesheet loading from bowing was accounted for analytically in the testing.
Three different mechanical tests were conducted to determine F*; a locked tube test, pressure cycling, and an ultimate load test. All tests were conducted at ambient temperatures. The locked tube test simulated the loading applied
- to a steam generator tube during cooldown of the plant assuming the tube was locked at a tube support plate location. The unequal coefficients of thermal expansion of the tube wrapper and the tube would lead to an applied tensile load on the tube.
For the pressure cycling test, several tubes were subjected to pressure cycling between low and normal operating pressures. Motion of the tube was monitored during the cyclic loading.
Finally, tubes were subjected to an ultimate load test.
Tubes were internally pressurized and subjected to I
an increasing axial tensile load until failure.
Failure was defined as a relative movement of a specified distance between the tube and tubesheet.
As part of the test program to provide the basis for the proposed F* 1ength, steam generator tubes were subject to leak rate testing.
Tubes were internally pressurized to simulate differential pressures during normal operating and faulted conditions. The acceptance criteria for these tests specified an allowable leakage limit.
Tube displacements were also monitored during the tests.
3.1.3 F* Test Results l
Based on the results of the leakage rate and mechanical testing, the licensee determined a nominal engagement length necessary to ensure adequate margins of safety. Accounting for limited sample size, statistical scatter in the data, and NDE inspection error, this value was increased appropriately. The licensee has proposed that steam generator tubes with degradation in the roll expanded portion of the tube can remain in service if all degradation lies below the F* distance. The F* distance is equal to the 1.7 inches and is measured down from the secondary face of the tubesheet or the top of last hardroll, whichever is further into the tubesheet.
1 3.2 Evaluation of Proposed Technical Soecification Chanaes The licensee proposed a revision to the applicable Byron, Unit 1, and Braidwood, Unit 1, TSs to implement the F* criterion.
The following summarizes the proposed changes:
l i
1.
The TSs define the F* distance, that has been determined to be 1.7 inches, as the distance into the tubesheet from the secondary face of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet.
2.
The revised TSs include a definition of an F* tube, which is a steam generator tube with degradation below the F* distance and has no r
indications of degradation within the F* distance.
l 3.
In addition to the minimum sample size for steam generator tube inspection, all F* tubes in the tubesheet region will be inspected I
during all future outages.
In support of the proposed amendment to the Byron, Unit 1, TSs and the Braidwood, Unit 1, TSs, tests were completed to determine an acceptable F*
t
. i distance. The testing utilized specimens which reflect the actual tube-to-tubesheet joint configuration within the plant steam generators.
Unknown variables, which could potentially affect the calculated F* distance were taken into consideration in developing the test matrix. Applied loads for structural assessment and leakage rate testing were specified in accordance with staff recommendations in RG 1.121 and the ASME Code. The licensee's proposed changes to the Byron, Unit 1, and Braidwood, Unit _1, TSs are consistent with the conclusions from the test program to determine F*.
To ensure continued integrity of F* tubes, the licensee has incorporated a l
requirement into the plant TSs to reinspect F* tubes during all future l
outages.
The continued inspection of F* tubes during each examination will minimize the potential for the existence of degradation within the F* distance of all F* tubes.
The staff has reviewed the TS change related to the implementation of the F*
criterion proposed by the licensee in their submittals dated May 20, 1994, December 2, 1994, February 2, 1995, and March 14, 1995.
Based on information provided in these submittals, the staff finds the licensee's proposed changes acceptable.
4.0
SUMMARY
The licensee submitted a proposed amcndment to the applicable TSs for Byron Station, Unit 1, and Braidwood Station, Unit 1.
The proposed changes would permit. steam generator tubes to remain in service with degradation in excess of the current plugging limit provided the degradation exists below the F*
distance. The proposed F* distance is 1.7 inches measured down into the tubesheet from the secondary face of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet.
The licensee evaluated a worst-case flaw present.at the F* distance and concluded that the proposed changes are consistent with the NRC guidelines for developing steam generator tube plugging criteria.
The staff has reviewed the licensee's submittals and concludes that the proposed TS changes applicable to Byron, Unit 1, and Braidwood, Unit 1, on steam generator tube surveillance requirements are acceptable.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments.
The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with regard to the installation or use of a facility component located within the restricted area and change surveillance requirements.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types,
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.t
- i 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (59 FR 34659 and 60 FR 16184). Accordingly, r
the amendments meet the eligibility criteria for categorical exclusion set i
forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
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 Commission's regulations, and (3) the issuance.of the amendments will not be. inimical to the common defense and security or to the health and safety of the public.
j Principal Contributor:
P. Rush Date: June 22. 1995 I
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