ML17333B010
| ML17333B010 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/26/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17333B009 | List: |
| References | |
| NUDOCS 9708280041 | |
| Download: ML17333B010 (17) | |
Text
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SAFETY EVALUATIONBY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF TO USE ASME ANSI OM CODE-1995 FOR SNUBBER TESTING INDIANAMICHIGANPOWER COMPANY DONALD C. COOK NUCLEAR PLANT UNIT NOS.
1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
The U.S. Code of Federal Regulations, 10 CFR 50.55a, requires that inservice inspection (ISI) of certain ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PVJ Code applicable edition and
- addenda, except where specific written relief has been requested by the licensee and granted by the Commission pursuant to paragraph 10 CFR 50.55a(g)(6)(i), or alternatives approved pursuant to 10 CFR 50.55a(a)(3).
In requesting relief, the licensee must demonstrate that the requirement is impractical for their facility. In proposed alternatives, the licensee must demonstrate that (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, Additional requirements for the ISI of snubbers are contained in the Technical Specifications (TS).
In addition to reiterating the requirements of 10 CFR 50.55a, the TS provide specific requirements for visual inspections, visual inspection acceptance
- criteria, functional tests, hydraulic snubbers functional test acceptance criteria, and snubber service life monitoring.
Section 50.55a authorized the Commission to grant relief from ASME Code requirements upon making the necessary findings.
The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. The NRC staff's finding, with respect to granting or not granting the relief request as part of the licensee's ISI program, is contained in this Safety Evaluation (SE).
This SE evaluates a letter dated July 3, 1997, from the Indiana Michigan Power Company, requesting approval to use the ASME/ANSI OM Code-1995, Subsection ISTD, as an alternative to the ASME BRPV Code,Section XI, requirements.
The licensee's snubber testing program, third 10-year ISI interval, is based on the requirements of Section XI of the ASME B5PV Code, 1989 edition and addenda through 1988.
The 1989 edition of the ASME BSPV Code;- Section XI, defers snubber examination and testing requirements to ASME/ANSI OMa-1988, Part 4.
9'7082800@i 970826 PDR ADOCK 050003i5 P
t 2.0 RELIEF REQUEST By letter dated July 3, 1997, the licensee requested relief pursuant to 10 CFR 50.55a(3)(i) from performing snubber examination and performance testing in accordance with the ASME B5PV Code,Section XI, requirements as defined in ASME/ANSI OMa-1988, Part 4, for steam generator snubbers.
2.1 Licensee's Pro osed Alternative The NRC,staff developed Generic Letter (GL) 90-09, "Alternative Requirements For Snubber Visual Inspection Intervals and Corrective Actions," to provide an alternative to the existing excessively restrictive inspection schedule for snubber visual inspections.
The Donald C. Cook Nuclear Plant, Unit Nos.
1 and 2 Technical Specifications have been revised to incorporate the GL 90-09 inspection schedule recommendations.
The ASME/ANSI OM Code-1995, Subsection ISTD, includes the guidance of GL 90-09 and is consistent with the Donald C. Cook Nuclear Plant, Unit Nos, 1 and 2 revised TS. The flexibilityprovided in the inspection schedule requirements of the ASME/ANSI OM Code-1995, Subsections ISTD, will result in a reduction of the occupational radiation exposure by as much as 50%.
The ASME/ANSI OM Code-1995, Subsection ISTD, provides an acceptable level of quality and safety and will have no adverse impact on public health and safety.
3.0 EVALUATION The staff developed GL 90-09 to provide an alternative schedule for snubber visual inspections that maintains the same confidence level as the existing inspection intervals and allows for inspections and corrective actions during plant outages.
The GL 90-09 guidance on inspection intervals is based on the number of unacceptable snubbers of the last inspection in proportion to the size of the various snubber populations or categories.
The required current interval for visual inspection is based only on the number of unacceptable snubbers found during the last inspection without regard to the snubber population.
The staff determined that the visual inspection schedule of GL 90-09 is an acceptable alternative and encouraged licensees to amend the TS to include GL 90-09 snubber visual inspection criteria.
By letter dated May 1, 1992, as supplemented June 18, 1993, the licensee submitted a
request for an amendment to the Donald C. Cook Nuclear Plant, Unit Nos.
1 and 2 TS, to change the snubber visual inspection intervals and corrective actions in the TS surveillance requirements to the alternative criteria provided in GL 90-09.
The proposed amendments to the TS were approved and issued as Amendment Nos. 173 and 156 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2, respectively, by NRC letter dated July 9, 1993.
The staff has determined that the proposed use of ASME/ANSI OM Code-1995, Subsection ISTD, provides an acceptable level of,quality and safety since it is consistent with the guidance provided by the staff in GL 90-09 in that it maintains the same confidence level as the existing inspection schedule.
However, although the ASME/ANSI OM Code-1995
4-was approved by the Board of National Standards and Codes and issued by ASME on February 28, 1995, it has not been generically approved for use by the NRC. Ongoing 10 CFR 50.55a rulemaking involving the OM Code is expected to be endorsed for publication in the U.S. Code of Federal Regulations in 1998.
Accordingly, the staff's approval of this relief request is granted on an interim basis pending staff review and approval of the ASME/ANSI OM Code-1995 on a generic basis.
At such time, the interim nature of the staff's approval will cease.
4.0 CONCLUSION
The staff concludes that an interim approval of the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i), in that the proposed alternative will provide an acceptable level of quality and safety.
Principal Contributor:
J. Hickman F. Grubelich (By precedent)
Dated:
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August 25, 1997 Mr. E. E. Fitzpatrick, Vice President Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive
- Buchanan, Ml 49107
SUBJECT:
DONALDC. COOK NUCLEAR PLANT, UNIT NOS.
1 AND 2 - REQUEST FOR ADDITIONALINFORMATION RE: THERMO-LAG RELATED AMPACITY DERATING ISSUES (TAC NOS. M85538 AND M85539)
Dear Mr. Fitzpatrick:
By letter dated March 20, 1997, Indiana Michigan Power Company submitted a response to the NRC Request for Additional Information related to Generic Letter 92-08, "Thermo-Lag 330-1 Fire Barriers," for the Donald C. Cook Nuclear Power Plant, Unit Nos.
1 and 2. The NRC staff, in conjunction with its contractor, Sandia National Laboratories, has completed the review of your response and has identified a number of open issues and concerns requiring clarification.
Please respond to the enclosed Request for Additional Information within 45 days.
Please contact me at (301) 415-3017 if you have any questions on the above.
Sincerely, Original Signed By John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosure:
Request for Additional Information cc w/encl: See next page DISTRIBUTION:
Docket File PUBLIC PD3-3 R/F EAdensam (EGA1)
GMarcus OGC ACRS RJenkins LTran DOCUMENT NAME: G:)DCCOOKttCO85538.RAI To receive a copy of this document, Indicate In the box: M~ Copy without attachment/enctoaure "E ~ Copy with attachment/enctoatxe N ~ No copy OFFICE PM:PD33 E
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NAME JHickman DATE g/a~/9 Boyle
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2058~1 August 25, 1997 Mr. E. E. Fitzpatrick, Vice President Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive Buchanan, Ml 49107
SUBJECT:
DONALDC. COOK NUCLEAR PLANT, UNIT NOS.
1 AND 2 - REQUEST FOR ADDITIONALINFORMATION RE: THERMO-LAG RELATED AMPACITY DERATING ISSUES (TAC NOS. M85538 AND M85539)
Dear Mr. Fitzpatrick:
By letter dated March 20, 1997, Indiana Michigan Power Company submitted a response to the NRC Request for Additional Information related to Generic Letter 92-08, "Thermo-Lag 330-1 Fire Barriers," for the Donald C. Cook Nuclear Power Plant, Unit Nos. 1 and 2. The NRC staff, in conjunction with its contractor, Sandia National Laboratories, has completed the review of your response and has identified a number of open issues and concerns requiring clarification. Please respond to the enclosed Request for Additional Information within 45 days.
Please contact me at (301) 415-3017 ifyou have any questions on the above.
Sincerely, John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosure:
Request for Additional Information cc w/encl: See next page
E.
E. Fitzpatrick Indiana Michigan Power Company I
CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street
- Lansing, HI 48913 Township Supervisor Lake Township Hall P.O.
Box 818
- Bridgman, HI 49106 Al Blind, Site Vice President Donald C.
Cook Nuclear Plant 1 Cook Place
- Bridgman, MI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, HI 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, NW.
Washington, DC 20037 Mayor, City of Bridgman P.O.
Box 366
- Bridgman, HI 49106 Special Assistant to the Governor Room 1 State Capitol
- Lansing, MI 48909 Drinking Water and Radiological Protection Division Michigan Department of Environmental guality 3423 N. Martin Luther King Jr Blvd P.O.
Box 30630 CPH Hailroom
- Lansing, HI 48909-8130 Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Steve J.
Brewer Indiana Michigan Power Company Nuclear Generation Group 500 Circle Drive
- Buchanan, HI 49107
RE UEST FOR ADDITIONALINFORMATION DONALD C. COOK NUCLEAR POWER PLANT UNIT NOS. 1 AND 2 FIRE BARRIER AMPACITYDERATING ISSUES TAC NOS. M86638 AND M85539
1.0 BACKGROUND
By letter dated Nlarch 20, 1997, Indiana Michigan Power Company (the licensee) submitted its response to the second NRC Request for Additional Information (RAI) related to Generic Letter 92-08, "Thermo-Lag 330-1 Fire Barriers," for the Donald C. Cook Nuclear Power Plant
{DCNPP), Unit Nos.
1 and 2. This response included the following attachments: :
Response
to RAI Regarding Thermo-Lag Related Ampacity Derating Issues for Cook Nuclear Plant; :
Table Depicting Correlation Between the Predicted and Measured Ampacities; :
Model Computer Code; :
Test Report CL<92; :
Comparison Tables Providing Base Information Regarding Trays and Conduits Cable Full Load Amperes, and Comparison of Calculated Ampacities vs. ICEA Ampacities; and :
Ampacity vs. Depth of Fill Plot for¹12 AWG Cable in Tray.
The DCNPP ampacity derating methodology involves the application of two separate but related models.
In the first model (denoted as Part 1 Analysis), the licensee calculates the overall heat rejection capacity for a given cable tray (conduit) based on heat transfer correlations and calculations.
In this analysis, a series of calculations are performed for each uniquely sized cable in the tray and the limiting heat intensity is used in the Part 2 analysis for all cables in the tray.
In the second model (denoted as Part 2 Analysis), the licensee calculates, based on the Part 1 analysis, the allowable ampacity limitfor individual cables.
This calculation is basically a partitioning of the total heat load to individual cables.
The NRC staff, in conjunction with its contractor, Sandia National Laboratories {SNL), has completed the review of the licensee's submittal, and requires that the questions listed below
're addressed by the licensee.
2.0 QUESTIONS 2.1 The staff requested in its RAI dated December 2, 1996, (Item 2.2 - Part 1 Analysis, Appendix A of Attachment A) that Appendix C, as well as any other documentation that willsupport the validation of the experiments cited in the licensee submittal dated May 12, 1995, be submitted for staff review.
ENCLOSURE Although the licensee submittal dated March 20, 1997, provided additional clarification, SNL found that the licensee plot of ampacity versus (d/Rg" and its initial validations were based only on those tests that involved a nondiverse cable fill. The original concern raised by SNL pertained to the fact that the licensee has not shown that the thermal model, specifically, the diameter-based partitioning method, was appropriate for diverse cable loads actually installed in the plant. An analysis of the licensee test data by SNL indicates that the licensee assumed the diameter partitioning method may not be appropriate and may result in overestimation of the ampacity limits for smaller cables in diversely loaded trays (see Section 3.2 and Appendix C of the attached SNL report for further details.)
The licensee is requested to address the subject concerns and to provide an adequate validation of the heat load partitioning methodology.
2.2 The staff requested in its RAI dated December 2, 1996, (Item 2.3 - Part 2 Analysis, Appendix B of Attachment) that the licensee provide a direct comparison of predicted cable ampacity limits to those measured in experiments on the corresponding system in order to validate its calculations.
SNL finds that the response provided in the licensee submittal dated March 20, 1997, to be only partially responsive to the subject concerns.
Specifically, model versus experimental comparisons have only been provided for two specific tests.
None of the cables involved with the subject tests were loaded to its ampacity limit. SNL identified the following concerns with the licensee validations efforts:
(1) the licensee model predicts an ampacity value significantly higher than measured in the test; (2) the licensee studies did not consider the full range of test dates available to the licensee; and (3) SNL assessments have identified potential inconsistencies between the model results and experimental test data.
Overall, SNL finds that the licensee validations study, as provided by the subject licensee submittal, to be inadequate to justify reliance on the licensee approach to thermal modeling techniques (see Appendix C, Section 2.2, and Section 3.3.2 of the attached SNL report for further details.)
The licensee is requested to address the subject concerns and to provide a complete validation of ampacity model results to applicable experimental test data.
2.3 The staff requested in its RAI dated December 2, 1996, (Item 2.3-Part 2 Analysis, Appendix 8 of Attachment) that the licensee provide the supporting validation results which justified its treatment of cable trays using equivalent annular region assumptions.
As stated in Item 2.2 above, SNL has found that the licensee validation studies to be inadequate and that, in its own validation studies, SNL identified inconsistencies and potential nonconservatisms in the licensee thermal model.
(See Sections 3.3.6 and 4.2, and Appendices B and C of the attached SNL report for further details.)
The licensee is requested to address the subject concerns regarding validation of its thermal model.
2.4 The staff requested in its RAI dated December 2, 1996, (Item 2.4 - Representative Calculation Results) that the licensee provide a direct comparison between ampacity loads installed at DCNPP to the ICEA Standard P-54<40 ampacity limits and to justify any cases which exceed ICEA P-54-440 limits. SNL reviewed Attachment 5 calculations which were contained in the licensee submittal dated March 20, 1997, and identified four apparent errors.
As a result of a reanalysis of the most significant of the licensee cable applications, SNL identified four specific cables which appear to be nominally overloaded even in the absence of the fire barrier system and six additional cables which may have insufficient ampacity derating margin.
(See Section 4.5 and Appendix A of the attached SNL report for further details.)
The licensee is requested to provide additional justification for the acceptability of the ampacity loading on the 10 cables as identified in the attached SNL report.
t 2.5 SNL found that several licensee responses as provided in the licensee submittal dated March 20, 1997, did not provide sufficient information to resolve the ide'ntified concerns cited in the staff RAI dated December 2, 1996.
However, all of the deficient responses were directly related to thermal modeling concerns.
Those deficient responses which were not identified in the above items are discussed in Sections 3.3.3, 3.3.6, 3.3.7, 3.3.8, 3.3.9, 3.4.1, 3.4.2, 3.4.6, and 3.4.8 of the attached SNL report.
Although the staff agrees with its contractor SNL that the pursuit of an integrated resolution of the concerns for the licensee thermal model and its implementation is the primary review objective, the staff requests that the licensee address the above deficient responses in those instances where the applicable concern may materially impact the acceptability of the licensee ampacity model results and validation studies.
Overall, SNL has recommended that the licensee thermal model should not be credited as an appropriate basis for analysis without significant further review and validation.
Further, in the case where the licensee plans to continue to seek acceptability of its current thermal model, SNL recommends that the licensee provide the following changes or additional information:
The actual computer code implementation should be reformulated in order to comply with accepted engineering practice for a structured program.
This revision should include an organized logical code structure that allows for a direct and clear understanding of the program flow and flow control, and the explicit definition of internal variables, input variables, and flow control parameters.
The revised code should be amenable to independent implementation and verification.
The validation study should be supplemented to provide a direct comparison of total allowable heat rejection capacity as calculated by the thermal model and as measured in these tests, i.e., the 11 specific tests cited in the licensee CL-492 test report.
The subject tests include those tests performed using the "solid bottom/solid cover/no fire barrier" tray configuration and those tests performed using the "vented bottom/no cover/solid fire barrier" tray configuration.
The staff fully supports the above SNL recommendations.
The staff requests that the licensee either implement those recommendations or provide an alternate technical analysis to resolve the subject concerns regarding the licensee thermal model and its implementation at the Oonald C. Cook Nuclear Plant.
Attachment:
Ltr Rpt to USNRC, Rev. 0, dtd 6/19/97
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