ML17332A928
| ML17332A928 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/13/1995 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17332A927 | List: |
| References | |
| NUDOCS 9509190393 | |
| Download: ML17332A928 (13) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO, 200TO FACILITY OPERATING LICENSE NO.
DPR-58 I DIANA MICHIGAN POWER COMPANY ONA D C.
COOK NUCLEAR AN UNIT NO.
1 DOCKET NO. 50-315
1.0 INTRODUCTION
By letter dated February 3,
- 1995, as supplemented April 25, 1995, the Indiana Michigan Power Company (the licensee) requested an amendment to the Technical Specifications (TS) appended to Facility Operating License No.
DPR-58 for the Donald C.
Cook Nuclear Plant, Unit No. 1.
The proposed amendment would
- revise, in part, TSs 4.4.5.2, 4.4.5.3, 4.4.5.4, 4.4.5.5, and 3.4.6.2 for D.C.
Cook Unit 1, for Cycle 15 operation to permit the use of voltage-based steam generator tube repair criteria.
The voltage-based steam generator tube repair criteria allows axially oriented outside diameter stress corrosion cracking (ODSCC) confined within the thickness of the tube support plates to remain in service based on the magnitude of the eddy current voltage response.
The NRC staff has reviewed the submittals noted above and additional information obtained in a phone call with the licensee on July 28, 1995.
The following is the staff's 'evaluation of the licensee's proposed TS amendment.
2.0 BACKGROUND
The staff has previously approved similar requests from the licensee to apply voltage-based tube repair criteria at D.C.
Cook Unit 1.
Implementation of voltage-based tube repair criteria for fuel Cycle 13 was approved as documented in a letter to the licensee dated July 29,
- 1992, "Donald C.
Cook Nuclear Plant, Unit 1
Amendment No.
166 to Facility Operating License No.
Similarly, implementation of voltage-based tube repair criteria for fuel Cycle 14 was approved as documented in an amendment to the license dated March 15,
- 1994, "Donald C.
Cook Nuclear Plant, Unit No.
1 Issuance of Amendment RE:
Incorporation of 2.0 Volt Steam Generator Tube Support Plate Interim Plugging Criteria For Cycle 14."
In the previous voltage-based tube repair amendments listed above, the staff concluded that the tube repair limits and leakage limits would ensure adequate structural and leakage integrity for indications accepted for continued service at D.C.
Cook Unit 1, consistent with applicable regulatory requirements.
This evaluation addresses comparable tube repair criteria for operating Cycle 15; however, in this amendment, the licensee has proposed to increase the voltage limits from 2.0/3.6 volts to 2.0/5.6 volts.
9509i9'0393 9509i3 PDR ADQCK 050003i5 P
The staff has been developing generic criteria for voltage-based limits for ODSCC confined within the thickness of the tube support plates.
The staff has published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, "Voltage-Based Interim Plugging Criteria for Steam Generator Tubes,"
and in a draft generic letter (GL) titled "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes."
The latter document was published for public comment in the federal
~e ister on August 12, 1994 (59 FR 41520).
On August 3, 1995, the staff issued GL 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," that took into consideration public comments on the draft GL cited above, domestic operating experience under the voltage-based repair criteria, and additional data which have been made available from European nuclear power plants.
The licensee submitted its amendment request prior to the date that the NRC issued GL 95-05.
Consequently, the amendment request was based on the guidance in the draft GL.
The licensee's current proposal is applicable to Cycle 15 operation and is similar to the licensee's prior amendment proposals which were approved as documented in the references mentioned previously.
Furthermore, the licensee's submittal is consistent with GL 95-05 except as noted below.
- 3. 0 PROPOSED INTERIM TUBE REPAIR CRITERIA Donald C.
Cook Nuclear Plant, Unit 1, Technical Specifications 4.4.6.2, 4.4.6.4, 4.4.6.5, and 3.4.7.2 and Bases 3/4.4.6 and 3/4.4.7 would be revised by this amendment request to specify the voltage-based tube repair criteria for ODSCC confined to within the thickness of the tube support plates.
Modifications have been made to the previously approved (Cycles 13 and 14)
TSs pertaining to the implementation of the voltage-based tube repair criteria so as to make the currently proposed TSs similar to those permitted by GL 95-05.
The changes in the TSs for Cycle 15 implementation of the voltage-based tube repair criteria include,'in part:
a.
Specifying that tube support plate indications left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during the following refueling outages.
b.
Specifying that the implementation of the steam generator tube support plate plugging criteria requires a
100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known ODSCC indications.
The determination of the cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a
20 percent random sampling of tubes inspected over their full length.
c.
Changing the Cycle 14 repair limits for tube support plate intersections with indications of ODSCC from 2.0 and 3.6 volts to the following for Cycle 15:
1.
Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.
2.
Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in c.3 below.
3.
Indications of potential degradation attributed to ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to 5.6 volts may remain in service if a rotating pancake coil inspection does not detect degradation.
Indications of ODSCC degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired.
d.
Adding the following reporting requirements:
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
- 1. If the estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing "Standard Review Plan-NUREG-0800" assumptions) during the previous operating cycle.
- 2. If circumferential crack-like indications are detected at the tube support plate intersections.
- 3. If the indications are identified that extend beyond the confines of the tube support plate.
- 4. If the calculated conditional burst probability exceeds 1 x 10'er reactor-year, as calculated per WCAP-14277, "SLB [Steam Line Break]
Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," notify the NRC and provide an assessment of the safety significance of the occurrence.
e.
Continue to specify a limit on primary-to-secondary leakage of 150 gallons per day through any one steam generator for Cycle 15.
In addition to the above TS changes, the licensee has also made the following commitments for implementing the voltage-based tube repair criteria:
1.
The requested actions of GL 95-05 will be followed with the exception of the use of a probe wear standard and the use of bobbin coil probes with the tolerance specified in Section 3.c.2 of the GL.
These exceptions are discussed in Sections
- 4. 1 and 4.2 of this evaluation.
In addition, the licensee has proposed not to include the mid-cycle equation for determining the voltage limits in the event of.
a forced outage not attributable to ODSCC at the tube support plates.
2.
Calculation of, the conditional probability of burst and total leak rate during a main steamline break (NSLB) will follow the methodology described in WCAP-14277.
As discussed in WCAP-14277, these methods are intended to be in accordance with the draft GL on voltage-based tube repair criteria.
J 4
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4 The methods as specified in the draft generic letter are unchanged in GL 95-05.
3.
The NRC will be notified prior to restart if any indications of primary water stress corrosion cracking (PWSCC) are detected at the tube support plate elevations.
Furthermore, the data analysts will be briefed on the possibility that PWSCC can occur at tube support plate elevations.
4.
No distribution cutoff will be applied to the voltage measurement variability distribution.
5.
All intersections where copper signals interfere with the detection of flaws will be inspected with a motorized rotating pancake coil probe.
6.
All intersections with large mixed residuals will be inspected with a rotating pancake coil probe.
7.
All bobbin flaw indications with voltages greater than 1.5 volts will be inspected with a rotating pancake coil probe.
4.0 EVALUATION 4.1 Ins ection Issues The licensee's inspection program is consistent with the guidance of GL 95-05 with the exception of the probe wear re-inspection requirements and the use of bobbin coil probes with the appropriate tolerances specified in Section 3.c.2 of the GL.
For the probe wear re-inspection requirements, the licensee proposes to use the same'ractices used during the last D.C.
Cook Unit 1 steam generator inspections as discussed in Attachment 5 to a letter from the licensee to the NRC dated December )5, 1993.
The requirements state that if any of the probe wear standard signal amplitudes prior to probe replacement exceed the il5 percent limit, by a value of "XX", then any indications measured since the last acceptable probe wear measurement that are within "XX" of the plugging limit will be reinspected with the new probe.
Alternatively, the voltage criterion may be lowered to compensate for the excess variation.
Section 3.c.2 of GL 95-05 specifies that the voltage response for the 40-percent to 100-percent through-wall holes of new bobbin coils calibrated on the 20-percent through-wall holes should not differ from the nominal voltage by more than 210 percent.
The licensee indicated that bobbin coil probes with such tolerances would not be available until after the licensee inspects the D.C.
Cook Unit 1 steam generators in the fall 1995 outage.
With respect to the use of the proposed alternate procedures for re-inspecting tubes that fail to meet the probe wear criterion, the staff has concluded that alternate methods may be used provided an assessment is performed demonstrating that (1) they provide equivalent detection and sizing capability on a statistically significant basis when compared to the guidance in GL 95-05 and (2) they are consistent with current methods for determining the end-of-cycle (EOC) voltage distributions which are used in the tube integrity analyses.
These assessments, along with the statistical criteria for
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demonstrating that the techniques are equivalent, should be provided to the NRC for review and approval.
With respect to this cycle specific application,
- however, the NRC staff has concluded that the methods to meet the probe wear criterion are acceptable.
The licensee indicated that bobbin coil probes with the voltage response tolerances specified in GL 95-05 will not be available until approximately 6 months after the NRC issues the GL.
The scheduled date for the inspection of the D.C.
Cook Unit 1 steam generators is less than 2 months following the release of GL 95-05.
The availability of the appropriate bobbin coil probes will be limited at the time of the inspection.
Due to the difficulty in obtaining bobbin coil probes with the response characteristics specified in Section 3.c.2 of GL 95-05, the licensee's decision not to inspect with such probes in the Cycle 15 refueling outage is acceptable to the NRC staff.
As a result of the potential for the possible development of PWSCC flaws at dented tube support plate intersections, the licensee will brief its eddy current analysts of the potential for PWSCC to occur at these locations.
Furthermore, the licensee has agreed to notify the NRC prior to plant restart if any PWSCC indications are detected at the tube support plate elevations.
The staff notes that if PWSCC is detected at tube support plate elevations, an evaluation to ensure voltage-based repair criteria are only applied to ODSCC indications may need to be performed.
The staff concludes that the inspection guidelines submitted by the licensee are acceptable since the proposed repair criteria are limited to one cycle.
In addition, the staff finds that the calibration, recording, and analysis requirements are consistent with the methodology used in the development of the databases and supporting evaluations.
4.2 Structural Inte rit 4.2.1 Deterministic Structural Inte rit Assessment The licensee's tube repair limits are based on a correlation between the burst pressure and the bobbin coil voltage of pulled tube and model boiler data.
This correlation is similar to that used in approving the voltage limits in the licensee's previous submittals and those used in GL 95-05.
The staff finds the licensee's proposed voltage limits acceptable given the current burst pressure/bobbin voltage database, the licensee's growth rates, and the non-destructive examination uncertainty estimates.
To confirm the nature of the degradation occurring at the tube support plate elevations, tubes are periodically removed from the steam generators for destructive analysis.
Tube pulls confirm that the nature of the degradation being observed at the tube support plate elevations is predominantly axially oriented ODSCC and also provide data for assessing the reliability of the inspection methods and for supplementing existing databases (e.g.,
burst
- pressure, probability of leakage, and leak rate).
GL 95-05 contains guidance that states utilities should periodically remove at least two tube support plate intersections with large voltage indications for destructive examination.
For subsequent tube pulls after the initial application of voltage-based repair criteria, GL 95-05 states that licensees should retrieve
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as many intersections as practical (minimum of two intersections) following accumulation of 34 effective full power months of operation or at a maximum interval of three refueling outages, whichever is shorter, following the previous tube pull.
In 1992, the licensee removed nine tube support plate intersections for metallogr aphic examination, burst testing, and leak rate testing from the D.C.
Cook Unit 1 steam generators.
As of the Cycle 15 refueling outage, D.C.
Cook Unit 1 will have accumulated approximately 29.5 effective full power months since the 1992 refueling outage.
In addition, the Cycle 15 refueling outage will only be the second outage since the previous tube pulls.
Therefore, the licensee has elected to not remove any tube support plate intersections in accordance with the guidance in GL 95-05.
Hetallurgical examination performed on the tubes removed during the 1992 refueling outage confirmed that the dominant degradation mechanism for the indications at the support plate elevations is axially oriented ODSCC.
The maximum voltage of the intersections removed was 2.02 volts.
The staff believes that no additional pulled tube data is required to support implementation of the 2.0-volt voltage limit for the next operating cycle (Cycle 15) provided no unusual inspection findings are identified during the inspection.
4.2.2 Probabilistic Structural Inte rit Assessment A probabilistic analysis for the potential for steam generator tube ruptures, given an HSLB, has been performed for the previous applications of this tube repair criteria.
The draft GL contains additional guidance on this analysis.
The licensee intends to perform this calculation per the guidance in the draft GL which will most likely re'suit in a higher conditional probability of burst than would have been obtained using the previous methodology since it includes parametric uncertainty.
The results of the probabilistic analysis will be compared to a threshold value of lxl0'er reactor-year according to the guidance in the draft GL.
This threshold value will provide assurance that the probability of burst is acceptable considering the assumptions of the calculation and the results of the staff's generic risk assessment for steam generators contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity."
Failure to meet the threshold value indicates that ODSCC confined to within the thickness of the tube support plate could contribute a
significant fraction to the overall conditional probability of tube rupture from all forms of degradation that was assumed and evaluated as acceptable in NUREG-0844.
The guidelines in GL 95-05 for calculating the potential for tube ruptures given an HSLB are equivalent to those of the draft GL.
The licensee referenced WCAP-14277, "SLB Leak Rate and Tube Burst Probability Analysis Hethods, for ODSCC at TSP Intersections,"
dated January
- 1995, as a
document containing the details of the methodology for calculating the conditional probability of burst given an HSLB.
The NRC staff previously approved the use of the methodology in WCAP-14277 as documented in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.
106 to Facility Operating License NPF-8, Southern Nuclear Operating Company, Inc., Joseph H. Farley Nuclear Plant, Unit 2, Docket No.
STN 50-364 dated April 7, 1995.
The staff concludes that the licensee's proposed methodology is consistent with the guidance in GL 95-05 and is acceptable for use in this outage-specific application.
4.3 Leaka e Inte rit 4.3. 1 ormal 0 erational Leaka e
Consistent with prior amendments approving the use of the voltage-based repair criteria at D.C. Cook Unit 1, the licensee will continue to limit the amount of operating leakage through any one steam generator to 150 gallons per day.
This requirement will be in effect for operation during Cycle 15.
432 ~id L
k The licensee indicated that it will calculate the leakage and MSLB tube burst probability following the guidance of the draft GL.
The guidelines of the draft GL are consistent with GL 95-05 with regard to leakage and conditional burst probability calculations.
In order to complete these calculations, the licensee will follow the methodology outlined in WCAP-14277.
The model for calculating the steam generator tube leakage from the faulted steam generator during a postulated MSLB consists of two major components:
(1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage (POL) model);
and (2) a model predicting leak rate as a function of voltage, given that leakage occurs (i.e., the conditional leak rate model).
The calculational methodology being proposed by the licensee for D.C.
Cook Unit 1 for determining t)e amount of primary-to-secondary leakage under
, postulated accident conditions has previously been reviewed and approved by the staff as documented in the Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.
106 to Facility Operating License NPF-8, Southern Nuclear Operating Company, Inc., Joseph M. Farley Nuclear Plant, Unit 2, Docket No.
STN 50-364 dated April 7, 1995.
The staff finds this methodology acceptable for an assessment of the D.C.
Cook Unit 1 steam generators for Cycle 15.
The staff notes that all applicable data should be included in the probability of leakage and conditional leak rate databases when performing this calculation.
The staff notes that some minor variations in the details of the modeling may be necessary for the case where the p-value test is invalid at the 5 percent level.
The staff, however, finds the licensee's proposal to perform the calculation using a methodology co'nsistent with the guidance of GL 95-05 acceptable.
The licensee completed calculations for the allowable steam generator leak rate in the faulted steam generator for its application to amend the D.C. Cook Unit 1
TS to apply a voltage-based plugging criteria during the refueling outage for Cycle 14.
The leakage value is intended to be consistent with maintaining the radiological consequences of a release outside containment to within a small fraction of the guideline values in 10 CFR Part 100.
As a result, if the primary-to-secondary leakage during a postulated MSLB is less than this allowable limit, the steam generator tubing will maintain adequate leakage integrity under these conditions.
The staff previously reviewed and approved the licensee's calculations for determining the maximum allowable
primary-to-secondary leakage in the license amendment approving the use of the voltage-based criteria for D.C.
Cook Unit 1 for fuel Cycle 14.
The staff concludes that the licensee's calculation of the allowable steam generator leak rate is acceptable since the previous results of these calculations remain valid for operation in Cycle 15.
S.O
~SU MAR The licensee submitted an application for a one cycle amendment to the Donald C.
Cook Nuclear Power Plant Unit 1 TS which would permit the use of voltage-based steam generator tube repair criteria.
The amendment application was submitted prior to the date when the NRC issued a final position regarding the implementation of voltage-based repair criteria in GL 95-05.
Consequently, the licensee's submittal follows the guidelines provided in the draft version of the GL.
The staff reviewed the proposed one cycle amendment to the D.C.
Cook Unit 1 TS and concluded that the methods proposed by the licensee are also consistent with the guidance in GL 95-05 except as noted above.
The staff concludes that adequate structural and leakage integrity can be ensured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied during Cycle 15 at the Donald C.
Cook Nuclear Power Plant, Unit 1.
The staff's approval of the proposed voltage-based repair criteria is based, in part, on the licensee being able to demonstrate that the projected EOC conditional probability of burst and the primary-to-secondary leakage during a postulated NSLB will be acceptable following the Cycle 15 refueling outage.
6.0 STATE CONSULTATION
In accordance with the Ci)mmission's regulations, the Hichigan State official was notified of the proposed issuance of the amendment.
The State official had no comments.
- 7. 0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
The staff has determined that the amendment involves no significant incr ease in the amounts, and no significant change in the types, 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 amendment involves no significant hazards consideration and there has been no public comment on such finding (60 FR 37093).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set 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 amendment.
8.0 CONCLUSION
The staff 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
P.
Rush September 13, 1995