B12568, Application for Amend to License DPR-65,adding Requirement That Annual Rept Submitted to NRC Must Include Documentation of All Failures & Challenges to Pressurizer PORV or Safety Valve.Application Submitted Per 870615 Commitment.Fee Paid
| ML20235S048 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/14/1987 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20235S050 | List: |
| References | |
| TASK-2.K.3.03, TASK-TM B12568, TAC-45357, TAC-65852, TAC-66116, NUDOCS 8707210531 | |
| Download: ML20235S048 (7) | |
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.mm ..v niara ce-HARTFORD. CONNECTICUT 06141-0270 L t j *[^5i'"]$ '(""jg (203) 665-5000 July 14,1987 Docket No. 50-336 B12568 Re: 10CFR$0.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk
. Washington, D.C. 20555 Gentlemen:
. Millstone Nuclear Power Station, Unit No. 2 Proposed Changes to Technical Specifications Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its Operating License, DPR-65, by incorporating the attached proposed changes into the Tephnical Specifications for Millstone Unit No. 2. In a letter dated June 15,1987,ll) NNECO committed to providing these proposed changes to the Technical Specifications by July 17, 1987. This submittal fully addresses that commitment.
Description of Changes Proposed changes to Page 3/4 4-7a, Surveillance Requirement 4.4.5.1.4, specifies that the plugging criteria for steam generator (SG) tube sleeves is a reduction in nominal wall thickness of 40 As part of this change, a note is removed from the bottom of the page that states that "the plugging limit for sleeves would be determined prior to the next refueling outage." This proposed change reflects the fact that a plugging limit for gleeves was defined by NNECO as documented in a letter dated May (2{,1984(2i and approved by the NRC Staff in a letter dated January 28,1985.31 Included with this package are two proposed changes to Section 6 of the Technical Specifications. The proposed changes to Page 6-19, Section 6.9.1.5 (1) E. 3. Mroczka letter to U.S. Nuclear Regulatory Commission, " Timeliness of Submittals," dated June 15,1987.
(2) W. G. Counsil letter to 3. R. Miller, " Resolution of Open items, Amendment 89 to DPR-65," dated May 25,1984. (
(3) ~3. R. Miller letter to W. G. Counsil, " Steam Generator Primary-to- l I Secondary Leak Rate Determination and Justification of 40% Plugging l Criteria for Degraded Sleeves on Millstone 2," dated January 28,1985.
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U'.S.' Nuclear Regulatory Commission B12568/Page 2 July 14,1987-l-
(Annual Report),' adds the requirement that the Annual Report submitted to' the NRC Staff is to include documentation of all failures of and challenges to the' pressurizer PORVs or safety valves. " Failures" are defined as the inability to lift within the tolerances allowed by' the design basis for Millstone Unit No. 2. The proposed change to Page 6-23 adds a new Section, Section 6.17 Secondary Water Chemistry, that ' specifies a program shall be maintained for monitoring secondary water chemistry.to inhibit steam generator tube degradation. The changes. to Section 6 of the Technical Specifications are be{ng proposed as requested by the NRC Staff in a letter dated October 29, 1986.(4/
Safety Assessment
- 1. SG Tube Plugging Limit Sleeving is a process used to repair steam generator tubes which exhibit wall degradation in excess of 40 percent, it consists of inserting a 40-inch inconel 625/690 tube into the degraded tubes to bridge the degraded section,~and expanding the top and bottom of the sleeve to form an upper and lower joint. Sleeving was first used at Millstone Unit No. 2 during the 1983 refueling octage.
Prior to the 1983 Millstone Unit .No. 2 sleeving program, an extensive qualification program was performed by the vendor to demonstrate the sleeve's corrosion resistance, leak tightness and structural capabilities.
The results of the qualification program demonstrated that a 40 percent sleeve plugging limit'provides sufficient margin against tube failure. This will result in the. requirement to remove from service all sleeves with defects greater than 40 percent throughwall, by plugging both ends of the affected sleeved tube.
Several tests and calculations were performed to determine the minimum sleeve wall thickness required to sustain normal and accident condition
, loads. These evaluations, which were performed in accordance with the guidelines of Regulatory Guide 1.121, assumed the surrounding tube to be completely severed; that is, no credit was taken for the residual strength of the tube. As required- by Regulatory Guide 1.121, several loading conditions (i.e., flow induced vibration, thermal, tube /tubesheet interaction, and pressure loads) as they relate to normal operating and test conditions were evaluated. Due to the location of the sleeves (sleeves were installed between the bottom of the tubesheet and the first egg crate) only those stresses resulting from internal and external pressure were found to be significant. The remaining stresses were found to be insignificant and, when added to the axial pressure stresses, the total stress was found to be less than the hoop pressure stresses. Several sleeve degradation scenarios (i.e., uniform sleeve wall thinning, part through infinitely long circumferential and axial flaws, and through-wall finite length flaws) were postulated and the resultant stresses and stress (4)- D. H. Jaffe letter to 3. F. Opeka, " Implementation Dates for Licensing Actions," dated October 29,1986.
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i intensities calculated. The results of these calculations showed that a factor of safety of 3 against failure, as recommended by Regulatory Guide 1.121, exists for the postulated scenarios.
Two types of accidents were evaluated: Main Steam Line Break (MSLB) {
coupled with Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident i (LOCA) plus SSE. Due to the sleeve locations, no significant bending l stresses exist at the sleeve locations due to these conditions. The only significant sleeve stresses ( in addition to the internal pressure stress) during an accident are the stresses resulting from the pressure differential across the horizontal portion of the tube bundle during a main steam line break. These stresses are accounted for by conservatively increasing the effective sleeve internal pressure from 2250 psi to 3150 psi. The same sleeve degradation mechanisms, as postulated under normal operation, were assumed and the stresses and stress intensities evaluated. These results showed that the structural criteria of both Regulatory Guide 1.121 and the ASME Section III were satisfied for emergency conditions.
The margin to burst during a postulated MSLB condition is a function of the mean radius to thickness ratio, based on a maximum permissible leak rate of 0.5 gpm due to an assumed normal operating pressure differential of 1365 psi. Based on this leak rate, it was determined that the maximum permissible throughwall crack length for Millstone Unit No. 2 is less than 0.495 inches while the critical throughwall axial crack length for burst for the Millstone Unit No. 2 sleeve based on a AP of 2285 psi is 0.71 inches.
(Axial flaws provide lower leak before break margin than circumferential flaws.) Since the critical throughwall crack length for burst is greater than the permissible throughwall crack length for the allowable primary-to-secondary steam generator leak rate, the leak-before-break condition is assured.
The failure of a sleeve could result in a single or multiple steam generator tube rupture. Even though the single tube rupture is an accident which has been previously analyzed, a multiple tube rupture could result in a serious health risk to the public, and either a single or multiple tube rupture could challenge the structuralintegrity of the reactor coolant system. Based on the postulated consequences of this failure, the following steps have been taken at Millstone Unit No. 2 to ensure that the potential for this failure is minimized:
- a. The sleeving material and sleeving installation process have been extensively tested and analyzed and the installation performed in accordance with qualified sleeving procedures.
- b. Periodic tube inspection is performed to ensure that tube / sleeve degradation is maintained within acceptable allowables.
- c. Leak-before-break conditions (as described above) increase the probability of a leaking flaw to be detected allowing the plant to be safely shutdown prior to a tube rupture.
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- d. The allowable plugging limit proposed Iri this change was obtained through several conservative assumptions, such as assuming uniform tube wall degradation, ignoring the residual degraded tube strength, etc. In addition to the above conservatism, a factor of safety of three on normal operating d P ensures that substantial margin exists between the maximum required tube strength and the ultimate tube structural capabilities.
The accepted industry value for eddy current test (ECT) sizing uncertainty is 110 percent throughwall for the inspection of inconel tubing. The bimetallic sleeve has no unique impact on the inspectability of the sleeve / tube assembly. While the existence of transitions and the sleeve end in the sleeve / tube assembly makes it necessary to use cross wound probes, which eliminate the influence of axisymmetric geometry variations, this does not add to the uncertainty of sizing ECT indications caused by flaws. Thus, the use of an ECT sizing uncertainty value of 110 percent throughwallis reasonable for the sizing of sleeve flaws.
Since sleeve inspections are performed only during refueling outages, the plugging limit must include a margin for flaw growth which may occur between inspection periods. Any flaw determined to be acceptable during an inspection must not degrade before the next inspection tc a point such that less than 40 percent of the sleeve wall remains intact. The bimetallic Inconel 625/690 sleeves exhibit resistance to stress corrosion cracking (SCC) under primary and secondary chemistry conditions which is equivalent or superior to the inconel 600 tube material in-service at Millstone Unit No. 2. Sleeve degradation is unlikely under typical Millstone Unit No. 2 chemistry conditions. However, to ensure that defects will not grow beyond the required sleeve wall thickness, the current conservative allowann of 10 percent throughwall defect growth per operating cycle for tubes was conservatively applied to sleeves.
Subtracting the ECT sizing uncertainty (10 percent) and the defect growth rate allowance (10 percent) from the allowable degradation (60 percent),
yields a value of 40 percent. Thus, sleeve degradation identified by ECT to be less than 40 percent through wall is acceptable for service. Any degradation 40 percent throughwail or greater will require plugging.
NNECO has reviewed this proposed change pursuant to 10CFR50.59 and has determined that it does not constitute an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety (i.e., safety related) previously evaluated in the safety analysis report is not increased, since a 40 percent sleeve plugging limit provides an equivalent margin against failure as a 40 percent tube plugging limit. The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created since failure of a sleeve, at worst, would be essentially equivalent to a tube rupture which is an accident previously considered. The margin of safety, as defined in the basis for any technical specification is not reduced since the plugging limit proposed in this evaluation was determined in accordance with the criteria of the
- ASME B and PV code,10CFR50 and Regulatory Guide 1.121.
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U.S. Nuclear Regulatory Commission B12568/Page5 July 14,1987 1
- 2. Channes to Administrative Controls l Both of the proposed changes to Section 6 of the Technical Specifications are administrative in nature. Proposed Section 6.9.1.5.d requires that the Annual Report include a listing of all failures (inability to lif t within the tolerances allowed by the design basis) and challenges to the pressurizer PORVs or safety valves. This change has no impact on plant operation or -
safety related equipment. Proposed Section 6.17, Secondary Water Chemistry, adds a requirement for maintaining a program to monitor secondary water chemistry to inhibit steam generator tube degradation.
The elements of the program described in the proposed Section are already being conducted in accordance with previously approved station procedures at Millstone Unit No. 2. The conduct of such a program enhances the safe operation of 'the unit and the addition of this section to the Technical
' Specifications formalizes the requirement to conduct the program.
NNECO has reviewed these proposed changes pursuant to 10CFR50.59 and has determined that they do not constitute an unreviewed safety question.
The probability of occurrence or the consequences of a previously analyzed accident have not been increased and the possibility for a new type of accident has not been created. The margin of safety defined in the Technical Specifications basis has not been reduced.
Significant Hazards Consideration NNECO has reviewed the attached proposed changes in accordance with 10CFR50.92 and has concluded that they do not involve a significant hazards consideration in that these changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. A 40 percent sleeve plugging limit provides the equivalent margin against failure as a 40 percent tube plugging limit. The proposed changes to Section 6 are administrative in nature and do not impact any of the design basis events.
- 2. Create the possibility of a new or different kind of accident from any previously analyzed. The failure of a SG tube sleeve would be equivalent to a tube rupture, which is an accident previoush evaluated. The administrative changes proposed do not introduce any new accidents but instead provide increased controls over plant equipment.
- 3. Involve a significant reduction in a margin of safety. The margin of safety associated with SG tube sleeves is equivalent to that of a SG tube. The proposed changes to Section 6 do not impact the margins of safety since they are administrative in nature. The only potential impact of these added requirements might be to increase the margin of safety.
The Commission has provided guidance concerning the application of the standards in 10CFR50.92 by providing certain examples (51FR7750, March 6, 1986). Example (vi) most closely resembles this change, i.e., "a change which
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. Nuclear Regulatory Co.mmission f B12568/Page 6 July 14,1987 either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan." The proposed plugging limit of 40 percent for SG tube sleeves provides 3 level of safety equivalent to the 40 percent plugging limit for SG tubes. The proposed changes to Section 6 are administrative in nature and represent additional controls and requirements.
The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the l attached proposed changes and has concurred with the above determinations.
l Although the plugging limit for SG tube sleeves at Millstone Unit No. 2 has been established and approved as documented in docketed correspondence, NNECO recognized that updating the Technical Specifications to reflect this ilmit would be prudent. Therefore, NNECO hereby requests that, if approved, this amendment be issued by the NRC Staff by December 1,1987 to be certain this specification is in place prior to the start of the refueling outage.
In' accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment.
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment request is the application fee of $150.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY A/
E. J.V' oczka' //
Seniof Vice President Attachment cc: Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, CT 06116 W. T. Russell, Region I Administrator D. H. Jaffe, NRC Project Manager, Millstone Unit No. 2 T. Rebelowski, Resident Inspector, Millstone Unit Nos. I and 2 l
U.S. Nuclear Regulatory Commission Bi2%8/Page 7 July 14,1987 STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD )
Then personally appeared before me E. 3. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensee herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
gNotary chhnin] Publif Y 1x/h My Commission ug.in Much 31,1988
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