ML20217G279
| ML20217G279 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/05/1997 |
| From: | Schopfer D SARGENT & LUNDY, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9708070155 | |
| Download: ML20217G279 (6) | |
Text
A h.A Sargerv%$Lundy9
- h Don K. Schopter
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August 5,1997 Project No. 9583100 Docket No. 50-423 Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 3 Independent Corrective Action Verification Program
- United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 In accordance with Subsection 4.3a of the Independent Corrective Action Verification Frogram
'(ICAVP) Audit Plan, Revision 3, this letter requests the Nuclear Regulatory Commission's (NRC) approval for sampling some aspects of the pipe support and pipe stress calculations. This request specifically addresses the pipe stress and component support calculations for the Service Water (SWP), Quench Spray (QSS) and Contair. ment Recirculatior (RSS) systems. However, it is expected that acceptance of the proposed sampling criteria for the QSS, RSS and SWP systems would also apply to the second at of systems selected for the Tier I review._ In addition, the approach and selection criteria described herein for pipe supports will also be applied to the conduit, cable tray and ductwork supports.
i The proposed sample sizes as well as the technical basis for the proposed sample size are described in the discussion below.
-A.
PIPE STRFSS CALCULATIONS A.1 Total Population The total population of pipe stress and related calculations for tiie three systems include:
SWP QSS/RSS Type-System System Total -
Large Bores Stress 62 36 98 Calculations Small Bore Stress 23-12 35
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Calculations 7
Tubing Stress Calculations 50 21 71 M0003 Vent and Drain Stress 91 36 127
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Calculation.4 Time History Load 3
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United States Nuclear Regulat::ry Commission August 5,1997 Document Control Desk Project No. 9583-100 Page 2 A.2 Large Bore Stress Calculations Ninety-eight (98)large bore piping stress calculations have been identified for review. We propose to conduct the review for these calculations using a two-level process. The first level review will be for 100% of the calculations and would verify that the design information, as stated in the calculations, is consistent with the system desigt. package, applicable design criteria, and other references provided in the calculations. The second level review will verify that this design information is correctly and accurately entered in the NUPIPE computer code and the computer output supports the conclusions of the calculations. We propose to review a sample of ten calculations (approximately 10%) for this second level review due to their repetitive nature. The rationale for selecting a 10%
sample is that the first level review would verify that the design information for 100% of the calculation is appropriate.
The 10% sample verifies the repetitive task of incorporation of the design infonnation as input to the NUPIPE computer code. The repetitive tasks include developing the piping model, preparing the NUPIPE input data, perfonning the computer analysis, and verifying that the computed stresses are within code allowable stresses. Every tenth calculation sorted by the calculation number will be selected for the second tier review. This selection process ensures an unbiased sample.
Due to the large population of the calculations, the selection of every tenth calculation also provides reasonable assurance that the overall large bore piping stress analysis process and its correct application will be thoroughly reviewed.
A.3 Small Bore Stress Calculations We propose a 100% review of small bore stress calculations.
AA Tubing Stress Calculations These types of calculations are primarily based on implementation of Procedure NETM-25 and a limited number of special case supplemental calculations which are referenced in each calculation as applicable. The basis for implementing these procedural requirements is to preparc a series of calculations which repetitively address the following eight individual design considerations:
- 1. Tabulate the applicable analytical data from the affected EK series isometric drawing and NETM-25,
- 2. Tabulate maximum allowable spans and minimum thermal offset per procedure.
- 3. Identify assumption or approaches which may differ from procedure.
- 4. Tabulate all applicable references.
- 5. Tabulate actual maximum spans and minimum offset froin isometric.
- 6. Calculate required thermal offset based on operating temperature.
- 7. Tabulate large bore thermal header movements from applicable large bore piping calculation at root valve location.
- 8. Verification that support loads meet BZ-600 ser:es support capacities (if applicable).
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= United States Nuclear Regulatory Commission August 5,1997 l
Document Control Desk Project No. 95$3-100 l
Page 3 l
l There are approxirantely 71 tubing stress calculations which are formatted in this manner, p
Based on the rationale presented in A.2 above, we propose to review eight (10%) of these types of calculations. The NETM procedure will also be reviewed for adequacy to code, design and licensing basis requirements.
A.5 Vent and Drala Stress Calculations These types-of camlations are cased on implementation of Procedure NETM-24, Millstone Nuclear Power Station Unit 3 Design and Installation of Small Bore Piping, and based on approaches defined in master Calculation (s) NP(F)-C.1-SWP/RSS/QSS. The methodology for implementing the above requirements is to prepare a series of calculations which repetitively address the following 15 individual design considerations for each piping configuration:
- 1. Local model configuration based on referenced design ISO.
- 2. Tabulation of piping data, properties, material allowables, etc., from referenced sources.
- 3. Tabulation of component, hardware weights from referenced sources.
- 4. C 'celation for configuration frequency check to verify applicability of method.
- 5. Calculation for critical stress location.
- 6. Calculation for Eq. 8 (pressure and dead weight stress evaluation),
- 7.. Tabulation of applicable local pipe, building and design acceleration values.
- 8. Calculation for seismic stress (normal / upset).
- 9. Calculation for Eq. 9 compliance (Steps 6 and 8 above),
- 10. Calculation for seismic stress (emergency / faulted).
I1. Calculation for Eq. 9 compliance (Steps 6 and 10 above),
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- 12. Tabulation of header thermal displacement and branch thic thermal expansion.
- 13. Thermal stress calculation.
- 14. Calculation for Eq.10 corapliance.
- 15. Flange load check.
There are approximately 127 vent and drain calculations. Based on the rationale presented in A.2, we propose to review 15 (10%) of these types of calculations. We will also review the NETM procedure for adequacy tu code, design and licensing basis requirements.
A6 Time History Load Calculations It is proposed to review 100% of these calculations.
s
United States Nuclear Regulatory Commission August 5,1997 Document Control Desk Project No. 9583-100 Page 4 5
B.
PIPE SUPPORT CALCULATIONS B.1 Total Population The total population for pipe support and related calculations for the three systems is as follows:
SWP QSS/RSS Type System System Generic Master Suppert N/A N/A 20 Calculations Non-Standard Pipe Support 567 463 N/A Calculations Non-Standard Tubing 65 40 N/A Support Calculations Standard Pipe Support N/A N/A 97
_ Calculations Standard Tubing Support N/A N/A 103 Calculations Liner Plate Calculations 0
116 N/A Embedment Plate 265 11 N/A Calcubtions Structural Steel 108 100 N/A Calculations B.2 Master Support Calculations The master support calculations contain design assumptions, methodology and generic criteria for the design of steel members, plates, welded connections, anchors and local stresses-in structural steel reembers.
We propose 100% detailed review of these calculations.
B.3 Non-Standard Pipe Support Calculations These calculations contain the structural evaluation of auxiliary steel members and standard support components. We propose a 10% smart sample of these calculations. A 10% sample is deemed sufficient since these calculations are repetitive in nature and since the generic design criteria and assumptions applicable to the individual pipe support calculations are contained in the master calculations, which, as stated above, will receive a 100% review.
The repetitive tasks inclade trutsposing design loads from stress calculations, developing computer models, and performing mathematical computations.
The 10% sample of non-standard supports will be se ected to ensure that adequate diversification of standard components types (i.e., strutr,, snubbers, straps, clamps, spring hangers, guides, etc.) is achieved and that a sufficient number of cases with high stress conditions is selected.
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United States Nuclear Regulatoty Commission August 5,1997 Document Control Desk Project No. 9583-100 Page5 Therefore, we propose the three step process as described below:
- 1) The total population of non-standard supports will b+ reviewed and grouped by the component type.
- 2) Select 10% of the total population of each component, The minimum size of the sample shall not be less than two (2) for any population greater than one support.
- 3) Magnitude of loads, thermal movements resulting in additional frictional loads and configuration types will be reviewed to ensure that potentially highly stressed conditions are selected.
B.4 Non-Standard Tubing Support Calculations
.These calculations contain the structural evaluation of auxiliary steel members and mechanical components. Using the same rationale as for Item B.3, we propose a 10%
sample these calculations.
B.5 Standard Pipe Support Calculations These calculations contain the structural evaluation of auxiliary steel members and mechanical components, Using the same rationale as for Item B.3, we propose a 10%
sample of these calculations.
B.6 Standard Tubing Support Calculations These; calculations contain the: structural evaluation-of auxiliary steel members and -
mechanical components.1Using the same rationale as foc Item B.3, we propose a 10%
sample these calculations.
B.7 -
Liner Plate Calculations These calculations contain evaluations of attachments to steel plates that are part of the
. containment liner. Using the same rationale as for item B.3, we propose a 10% sample of
'these calculations.
B.8-Embedment Plate Calculations These ' calculations contain evaluations of attachments to steel plates anchored to reinforced concrete structures. Using the same rationale as for item B.3, we propose a 10% sample of these calculations.
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United States Nuclear Regulat ry Commission August 5,1997 Document Control Desk Project No. 9583-100 i
Page 6 1
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Structural Steel Calculations These calculations contain evaluations of attachments to in-place structural steel beams.
Using the same rationale as for item B.3, we propose a 10% sample of these calculations.
In summary, we bdieve the proposed sampling will provide the level of review necessary to ensure the selected Tier 1 systems are capable of performing their dmctional requirements and that the design of the systems piping and pipe supports are consistent with the design and licensing basis as required by the confirmatory order. This conclusion is based on the following:
The proposed sampling only affects those calculations or calculation steps that are repetitive in nature.
The proposed sample size, where applicable, covers a sufficient number of the repetitive type calculations, such that a determination can be made as to the integrity of the repetitive type calculations, and the calculation process. If errors are identified which indicate potential process problems, expansion of the sample size will be discussed with :he NRC, The proposed sampling is consistent with the NRC's Oversight Inspection Plan which states that the inspection methodology for the ICAVP thall be similar to inspections described in inspection procedures 93801, Safety System Functional Inspections, and NC-2530, Integrated Design Inspection Program.
Both of these inspection procedures allow selection of appropriate inspection samples.
We are proceeding with the review based on the proposed sample sizes. _You may direct any questions to me at (312) 269-6078.
Yours very truly, h
s s
D. K. 'achopfer j
Vice, President and ICAVP Manager DKS:AAN:spr Copies:
E. Imbro (1/4) Deputy Director, ICAVP Oversight T. Concannon (1/2) Nuclear Energy Advisory Council J. Fougere (1/3) NU B. A. Erler (1/0)
File. (MNCAVNMepCaf9MR0805+ doc)
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