ML20248L805

Resubmits 980417 Suppl Info on Seismic Qualification of MSIV Leakage Path to Condenser,Per Telcon W/Nrc Staff.Ltr Supersedes Entirely

ML20248L805
Person / Time
Site:	Monticello
Issue date:	05/29/1998
From:	Hammer M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-96238, NUDOCS 9806110369
Download: ML20248L805 (44)

Text

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Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello. Minnesota 55362-9637 May 29,1998 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Demonstration of the Seismic Qualification of the MSIV Leakage Path at Monticello (TAC No. 96238)

Ref.1 Letter from M.F. Hammer, NSP, to NRC Document Control Desk, "NSP Response to Supplemental Request for Additional Information Conceming the Monticello Nuclear Generation Plant Power Rerate Program (TAC No. M96238)," March 26,1998 Ref. 2 Letter from M.F. Hammer, NSP, to NRC Document Control Desk, " Submittal of Information Regarding the Seismic Verification of the MSlV Leakage Path at Mont! cello (TAC No. M96238)," April 17,1998 By letter dated March 26,1998 (Ref.1), NSP informed the staff of its intent to take credit for fission product removal in the main steam lines and the condenser in certain Monticello accident scenarios under rerate operating conditions. By letter dated April 17,1998 (Ref. 2), NSP provided supplementalinformation on the seismic qualification of thc MSIV leakage path to the condenser.

A conference c.all was held between the staff and NSP regarding the scope and content of Ref.

2. NSP subsequently decided to resubmit the subject information. Attachment 2 contains NSP's amended submittal. This letter supersedes Ref. 2 entirely.

Please contact Joel Beres at (612) 295-1436 if additional information is required.

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Michael F. Hammer Plant Manager Monticello Nuclear Generating Plant 9806110369 900529 (Y

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Regional Administrator - 111, NRC NRR Pioject Manager, NRC

. Sr. R,esident inspector, NRC State of Minnesota, Attn: Kris Sanda J. Silberg. Esq.

Attachments Attachment i NRC Affidavit Seismic Verification of MSIV Leakage Path l

UNITED STATES NUCLEAR REGULATORY COMMISSION j

i NORTH $RN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 1

l Demonstration of the Seismic Qualification of the MSIV Leakage Path at Monticello (TAC No. 96238)

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Northern States Power Company, a Minnesota' corporation, by letter dated May 29,1998 I

provides information regarding the seismic qualification of the MSIV leakage path to the condenser for the Monticello Nuclear Generating Plant to a US Nuclear Regulato.y

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Commission (NRC). This letter contains no restricted or other defense information.

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NORTHERN STATES POWER COMPANY By M14/A s Niicha'el F. HHihmer Plant Manager Monticello Nuclear Generating Plant On this Y dayof hhr

\\%Rb before me a notary public in and for said County, personally appeaked Michael F. Hammer, Plant Manager, Monticello Nuclear Generating Plant, and being first duly swom acknowledged that he is authorized i

to execute this document on behalf of Northem States Power Company, and that to the best of his knowledge, information, and belief the statements made in it are true.

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' Samuel l. Shirey

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I Notary Public - Minnesota SAMUEL.t. SHIREY Sherbume County nomy rusue.nmanssota

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1 Seismic Verification of MSIV Leakage Path 1

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m__________

1.0 Introduction The Monticello Nuclear Generating Plant (MNGP) power rerate radiological analysis has taken credit for deposition and holdup of radioactive iodine in the steam lines downstream of the Main Steam isolation Valves (MSIVs) and in the main condenser.

The main condenser and a pathway from the MSIVs were evaluated to assure they would retain sufficient structural integrity following a safe shutdown earthquake (SSE) to transport the MSIV leakage. The MSIV leakege pathway includes leakage through the MSIVs via the main steam piping and main steam drains to the condenser.

The methodology suggested in NEDC-31858P (Reference 1) was used to seismically evaluate this pathway. This report will discuss the applicability of this methodology for Monticello and how this methodology was use'd for the seismic evaluation of the pathway. This report will summarize the seismic evaluation that was performed for the piping and equipment in the MSIV leakage path for Monticello. The evaluation demonstrates that a reliable pressure boundary can be maintained in the pathway for the MSIV leakage to reach the condenser during and after a seismic event.

The method of seismic evaluation relies in part on the use of earthquake experience data and similarity principles. Plant specific analyses of piping and equipment were used in combination with the experience method. The evaluation method and results are described in this report.

Guidance on the use of experience method for qualification of piping systems is described in reference 1 and also in the supporting documents cited within the reference.

Reference 1 provides an evaluation of the MSIV leakage issue for General Electric boiling water reactors, including Monticello. The Seismic Qualification Utilities Group (SQUG) Generic Implementation Procedure (GlP) described in reference 2 was used for seismic qualification of certain existing equipment in the MSIV leakage path.

2.0 Scope of Piping and Equipment The primary components in the MSIV leakage path which are relied on for pressure boundary integrity are the main condenser, the Main Steam (MS) lines from the MSIVs to the turbine stop valves and to the turbine bypass valves, and the drain lines to the condenser. Figure 2-1 shows a simplified diagram of the leakage pathway.

The MSIV leakage pathway that has been selected utilizes the drain lines from each of I

the four main steam lines. These drain lines are located downstream of the MSIVs and connect into a drain header that connects to the condenser. The leakage path utilizes three separate drain lines f;om the MS piping to the drain header. The three drain lines include the main steam drain lines, the main steam cross tie drain, and the turbine bypass line drain. Each of these finas can be isolated by Motor Operated Valves (MOVs). Each MOV has a bypass I ne with a restricting orifice. Since the MOVs are not powered by essential power and the y are normally closed valves, it is assumed that the leakage will be through the MOV bypass lines via the restricting orifices. This provides a passive pathway for the MSIV leakage to reach the condenser because no valve positioning or operator action is necessary to establish the pathway.

2

The branch lines which interconnect with the MSIV leakage path were included in the scope of the piping that was reviewed. The scope of the branch lines included the connection from the pathway to a location such as a closed valve that would assure that the MSIV leakage would be confined within the branch line, and leakage would be transferred to the condenser.

The turbine bypass valves are normally closed and fail closed. Because these types of valves are not well represented in the experience data, it was conservatively assumed that the valves would fail open as a result of the seismic event, and leakage would therefore go past the turbine bypass valves directly to the condenser. The piping from the turbine bypass valves to the condenser connects near the bottom of the condenser and was included in the scope of this evaluation.

The leakage path piping, equipment and supports are located in the following areas.

Reactor Building Turbine Building Recombiner Building

- HPCI Room

- Condenser Bay All areas

- RCIC Room

- Steam Jet Air Ejector (SJAE) Room

-Torus Area

- Mechanical \\'acuum Pump Room

- Main Steam

- Condenser Bay to SJAE Room Pipe Tunnel Chase

- Condensate Backwash Receiving Tank Room Table 2-2 presents a summary of the valves, in-line equipment, and attached equipment included in the scope. The piping was also segmented into 40 walkdown packages.

Table 21 identifies the piping packages.

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MAIN STEAM Steam Jet Air Ejectors ISOLATION VALVES o

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rm Figure 2-1 Monticello Main Steam IsolationVa Ive Leakage Pathways to the Main Condenser 4

4 Table 2-1: MSiV Piping Package List Package Piping System Description Location 2913-1 Main Steam Drain Condenser Bay / Steam Tunnel 2913-2 Main Steam Drains Condenser Bay l

i 2913-3 Pressure Equalizing Lines Condenser Bay l

2913-4 Piping from 10 P57-10-E to M 1617 Condenser Bay 2913-5 Pipe from Condenser Nozzle 8 to SJAE E-2B SJAE/ Pipe Tunnet/Cond. Bay 2913-6 Condenser Nozzle 8 to SJAE E-2B SJAE/ Pipe Tunnet/Cond. Bay 2913-7 RV33-6 -HB lines SJAE/ Pipe Tunnet/Cond. Bay 2913-8 RV34-6 -HB lines SJAE/ Pipe Tunnel /Cond. Bay 2913-9 MS to SJAE E-2A/E-2B SJAE/ Pipe Tunnet/Cond. Bay 2915-10 Ali injector Piping SJAE Room l

l 2913-11 T72 and T33 Tank Lines SJAE/MPV/ Hallway i

2913-12 From SJAE to Tank T72 and Off Gas Systern SJAE Room 2913-13 Off Gas Piping Recombiner Bldg / Buried 2913-14 Off Gas Piping Recombiner Bldg / Buried 2913-15 Drain Tank Feed and Discharge Lines SJAE/MPV Rooms 2913-16 Off Gas Steam Tap Line SJAE/ Pipe Tunnet/Cond. Bay 2913-17 Off Gas Small Bore Piping SJAE/MPV Rooms 2913-18 Off Gas Sample Line Con. Bay /SJAE/MPV Rooms 2913-19 Off Gas Sample System SJAE/MPV Rooms 2913-20 SHP System Steam Trap / Dryer SJAE Room 2913-21 Recombiner Trains Recombiner Bldg / Buried 2913-22 HPCI Pump Seal Lines Reactor Building 2013 23 C!cnd B!cwcr Diccharge Line Reacter Bui! ding 2913-24 MO-1739 Equa!! zing Line Condenser Bay 2913-25 MO-4000 Equalizing Line Condenser Bay 2913-26 Pressure Averaging System Condenser Bay / Turbine Deck 2913-27-1 Steam Seal System, Section 1 Condenser Bay 2913-27-2 Steam Seal System, Section 2 Condenser Bay 2913-27-3 Steam Seal System, Section 3 Condenser Bay 2913-27-4 Steam Seal RV Drain Lines Condenser Bay 2913-28 HPCl/RCIC Control Lines Reactor Building 2913-29 Off Gas Blower Discharge.

Buried /SJAE Room 2913-30 Hydrogen Water Chemistry System Recombiner Bldg 2913-31 Main Steam Stop Valve Drains Condenser Bay 2913-32 Bypass Valve Discharge Lines Condenser Bay 2913-33 Backwash Tank Drain Line Backwash Tank Room / Hallway 2913-34 Pump P-3 Feed / Discharge Pipe MVP Room 2913-35 T72 Tank Drain / Control Lines MPV Room 2913-36 SJAE Drain Lines SJAE Room 2913-37 Various l&C Lines SJAE Room / Condenser Bay 2913-38 V813 Tank Drain / Level Lines Condenser Bay 2913 Feedwater Heater Steam Trap Drain Lines Condenser Bay 2913-40 Misc Main Steam Drains and 1&C Lines Condenser Bay 5

Table 2-2: MS!V Leakage Path Equipment List Equipment ID(s)

Description 17-104 SAMPLE CHAMBER 17-116 OFF GAS SAMPLE RACK 17-136 OFF GAS SAMPLE BOX AO-1083A, AO-1083B 11 CDSR SUCT. ISOL.

AO-1084A, AO-1084B 12 CDSR SUCT ISOL SV-1, SV-2, SV-3, SV-4 TURBINE HIGH PRESSURE STOP VALVES CV-1242, CV-1243 SJAE STEAM SUPPLY CV-2046A, CV-2046B STEAM DRAIN TO MAIN CONDENSER CV-2082A, CV-2082B RCIC STEAM LINE DRAIN TO MAIN CONDENSER CV-4104, OV-4165 HWO O FLOWTO RECOMBINER CONTROLVALVE 2

E-1 A, E-1B HIGH PRESSURE, LOW PRESSURE CONDENSER E-204 HPCI GLAND SEAL CONDENSER E-2A, E-2B AIR EJECTORS E-4 STEAM PACKING EXHAUSTER K-200 GLAND SEAL BLOWER K-3A, K-3B STEAM PACKING EXHAUSTER BLOWERS LCV-7581 V-813 24" DELAY TANK VALVE MO-1048, MO-1049 STM PACKING EXHAUSTER BLOWER DISCH VALVES MO-2374 MAIN STEAM LINE DRAIN - OUTBOARD MO-2564 STEAM LINE DRAIN DOWNSTREAM MSIVs MO-2565 STEAM LINE DRAIN ORIFICE BYPASS MO-1045 STEAM SEAL REG FEED VALVE MO-4000 MA!N HEADER PRESSURE EQUAllZER DRAIN MOIST-SEP MOISTURE SEPARATOR PCV-7489A, PCV-7489B A RECMB TRAIN OG INLET VALVES PCV-7496A PCV-74968 OFFGAS BYPASS RETURN TO CONDENSER j

PCV-7497A, PCV-7497t3 OG STEAM SUPPLY VALVES j

PCV-7498A, PCV-74988 OG TRAIN STEAM SUPPLY VALVES

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RV-1007, RV-1011 SAFETY / RELIEF VALVE RV-1212, RV-1213 SAFETY / RELIEF VALVE RV-1244, RV-1245 SJAE STEAM SUPPLY RELIEF VALVES T-33 CONDENSATE BACKWASH RECOVERY TANK T-72 SEPARATOR TANK V-813 DRAIN COLLECTOR TANK V-F-11 HIGH EFFICIENCY FILTER I

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3.0 Application of Experience Data The staff and licensees have recently addressed seismic qualification of equipment in operating nuclear power plants as part of the resolution of Unresolved Safety issue A-46.

l Subsequent evaluations demonstrated that many non-seismically designed structures, f

systems, equipment and components have substantial inherent seismic ruggedness.

The Seismic Qualification Utility Group (SQUG) was formed in 1981 after an agreement with the NRC to develop alternative methods to resolve seismic safety issues for critical systems and components in operating nuclear stations. The primary method of equipment evaluation developed by SQUG and the staff uses empirical data from past earthquakes and from shake table tests (seismic experience data).

The seismic experience data approach includes tne foliowing oojectives Documentation of the most common causes of seismic damage or l

operational difficulties in facilities that contain structures, systems,

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equipment and components similar to those in nuclear stations.

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Credible definition of the threshold of seismic motion for various types of documented earthquake damage and shake table tests.

Identification of structures, systems, equipment and components that l

typically are not damaged in earthquakes much larger than design basis earthquakes for nuclear stations and other facilities and in shake table tests. inese cata provide insights to actual s ismic design margin.

Development of seismic integrity criteria that can credibly predict the performance of structures, systems, equipment and components in future earthquakes.

3.1 Experience Based Piping Capacity Experience from past strong motion earthquakes at conventional power plant and industrial facilities indicates that piping systems designed to industrial standards are rugged and can resist earthquakes of at least 0.5 g peak ground acceleration (PGA) [1].

This experience data includes piping systems which were not specifically designed for seismic loads. For all strong motion earthquakes affecting power stations in the United States since 1952, the amount of piping system failures observed was a very small 4

percentage (much less than 0.01 percent) of the total piping at risk. This leads to the conclusion that failure of piping in earthquakes is caused primarily by local conditions of weakness in the piping systems rather than global conditions of piping design or construction.

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Local failures in piping systems can stem from the following.

Relatively low piping flexibility in regions of relatively large displacements where piping is attached to building structures, massive equipment, or other piping.

Low piping ductility associated with the use of cast iron, PVC or other low-ductility materials.

Threaded pipe joints or other regions of reduced cross section with sharp corners susceptible to fatigue, ratchet cracking, or rupture when subjected to cyclic seisraic loads.

Regions of degraded pipe caused by corrosion or erosion.

Weak joints associated with friction type connections, or weak joints or repairs which result from poor welding.

Failure of piping associated with loss of non-ductile pipe supports.

For this effort, walkdown evaluations compared the subject piping systems to piping systems which have actually experienced strong motion earthquakes (experience data) to verify the seisrnic adequacy of the main steam piping leakage path. This process i

differs from the practice used historically in the nuclear power industry where the seismic adequacy of piping systems has been determined by analysis explicitly using computer modeling techniques. The results of the screening evaluation process work have been benchmarked against computer analysis results, which also demonstrate that the screening methodology can reliably be used to demonstrate the seismic adequacy of piping systems. This method utilizes a capacity vs. demand spectrum comparison, augmented by extensive walkdowns, worst-case calculations, and documentation to insure acceptable piping spans, piping support configurations, design attributes, and the cbsence of known seismic vulnerabilities.

The capacity spectra that were used in the establishment of the piping seismic capacity were based on the experience surveys and evaluations conducted in Reference [1].

Damage surveys at the facilities investigated indicated a very low piping failure (<0.01%)

and concluded that this failure rate was a result of isolated local weakness in piping systems which could be best screened by an in-plant walkdown. Reference [1] provides a seismic database from 123 sites occurring over 25 different earthquakes. The peak ground acceleration (PGA) estimates far exceed Monticello's design basis PGA of 0.12g for all but one site (Cachi Dam, Valle de Estrella Costa Rica earthquake) for which the PGA was 0.12g.

Figure 3-1 shows selected ground acceleration response spectra plotted against the MNGP SSE ground spectrum from three documented earthquakes occurring in Califomia. These include the 1971 San Femando (Valley Steam Plant - USGS Estimate), the 1979 Imperial Valley (El Centro Steam Plant), and the 1989 Loma Prieta (Moss Landing). The Valley Steam Plant record was obtained from Reference 8 and the remaining records are from Ret'erence 1. All of these earthquakes produced ground motions wellin excess of the MNGP SSE ground spectrum.

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Horizontal Ground Response spectra at 5% Damping 16 1.4 -

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Figure 3-1: Selected Spectra from References [1], [8] vs MNGP SSE Ground Spectrum Figure 4.1 of Appendix D of Reference [1] presents ground spectra at several of the survey sites and also shows the MNGP Design Response Spectrum.

Appendix D of Reference [1] describes the review and survey of piping experience data in relationship to main steam piping and condensers 3.2 Experience Based Condenser Capacity An evaluation of the seismic ruggedness of condensers and condenser anchorage for GE BWR plants is reported in Reference [1]. The configurations of the GE BWR condensers were compared to condensers in the earthquake experience data.

Condensers in the earthquake experience data exhibited substantial seismic rugDedness even when they were not designed to resist earthquakes. Comparisons of condenser designs in GE BWR plants with those in the earthquake experience data revealed the GE plant designs are similar to those that exhibited good earthquake performance. The study concluded that a failure and significant breach of pressure boundary in the event of a design basis earthquake is highly unlikely and contrary to a large body of historical experience data. The conclusions of that study were verified by detailed comparison of L

the Monticello condenser configuration to the earthquake experience data The comparison included a detailed evaluation of the Monticello condenser anchorage capacity.

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3.3 Experience Based Capacity of Related Equipment Other equipment in the scope of the leakage path review includes valves, instruments, and tanks which are referred to as Related Equipment in this report. The SQUG GIP methodology, documented in reference 2, is well suited to address the seismic adequacy of the equipment listed above. The GIP provides a formal procedure for evaluating these l

classes of equipment against the earthquake experience data. The GIP has been l

reviewed by the NRC as documented in Ref. 3. The implementation of the GIP procedure at Monticello is documented in Reference (4].

Figure 3-2 shows the GlP Reference Spectrum, the GIP Bounding Spectrum, and the MNGP SSE ground spectrum. Figure 3-2 shows that the MNGP SSE spectrum is well bounded by the GIP Spectrum.

HorizontalResponso Spectra at S% Damping 1<

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l 4.0 Seismic Evaluation Methodology 4.1 Piping and Supports l

The evaluation of piping included the following.

Walkdowns of the piping systems and associated supports which included identification of items judged to have inadequate seismic capacity, worst case pipe l

supports, and items requiring limited analytical reviews.

j A comparison of piping system demand versus experience-based capacity.

Limited analytical reviews and pipe support evaluations for piping systems l,

identified during the walkdowns.

Generation of Piping System Seismic Screening Work Sheet (PSSSWS), a formal L

method of documenting the walkdown, the limited analytical reviews, the worst case support evaluations, and the final seismic capacity evaluation.

The sections below provide details on the piping and support evaluations.

i 4.1.1 Comparison to the Exper!ence Data Piping Considerations The leakage path piping was compared to the piping in the experience data to insure the piping systems fall within the database contained in Reference [1] and within the ANSI B31.1 Power Piping Code. Key parameters in the comparison include the following.

(a)

Piping is fabricated and designed to B31.1, B31.3 or ASME BPVC Section Ill.

(b)

Piping sizes and materials fabrication fall within experience data.

(c)

Piping support vertical and lateral span ratios fall within the data base assumed by verifying the following span criteria below are met. These span criteria were p

based on a review of the data in referencs [1].

For Welded Steel Pipe:

- Vertical Spans are less than (1.5) times the suggested B31.1 Deadweight Spans.

- Horizontal Spans are less than six times the suggested B31.1 Deadweight Spans.

For Threaded Steel Pipe:

- Vertical Spans are less than (1.5) times the suggested B31.1 Deadweight Spans.

- Horizontal Spans are less than four times the suggested B31.1 Deadweight Spans.

(d) Piping operating pressures and temperatures fall within the experience data.

(e). Piping does not exhibit known failure modes or areas of potential weakness.

(f)

Pipng' support system is adequate, consistent with the piping systems in the experience data, and would be expected to exhibit a ductile failure mode.

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A comparison to the experience data was performed for the Monticello leakage path piping and is documented in Section 5 herein. For that comparison, materials, sizes, spans, and temperature ranges were compared to piping in the experience data to verify that the Monticello piping is adequately represented in the experience data.

Equipment Considerations.

In many inW nces, piping systems terminate at mechanical equipment such as pumps and tanks. There are three items of concem at these equipment piping interface locations.

(a)

Anchoraae of the equipment (b)

Nozzle loads applied to the equipment by the piping (c)

Equipment displacements applied to the piping system.

The walkdown procedure requires that the Seismic Review Team (SRT) address these concems. The SRT members were qualified in accordance with applicable industry criteria.

4.1.2 Limited At alytical Review of Piping and Supports This section defines the capacity criteria that was used in the limited analytical reviews of piping sy=tama and in the eve!uation of worst case supports. The capacity criteria is a stress-based criteria, and the demand criteria is in terms of an applicable input seismic excitation level. For specific analytical reviews such as Rod Hanger Fatigue reviews, a different Demand / Capacity criteria is used and is defined in the applicable analytical review package. For piping systems for which limited analytical reviews or analyses were conducted the capacity criteria below was used:

P+.75*i*[(M /Z)] s 1.0 S (4.1)

A P+.75*i'[(M /Z)+(Mei /Z)] s 2.4 S (4.2)

A i*[ Mc/Z + Me..m /Z ] s 2 SA (4.3)

P

= Pressure Loadings M

= Applied Moments Due to Deadweight Loadings A

Mai

= Applied Moments due to SSE seismic inertial Loadings Mes m = Range of Applied SSE Moments due to Seismic Anchor Motion (SAM)

Loadings Mc

= Range of Applied Moments due to Thermal Expansion and Thermal Anchor Motions Z

= Piping Section Modulus S

= Allowable Primary Stress limit per the B31.1 Code S

= Allowable Expansion Stress range per B31.1 Code A

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= Stress intensification factor as defined in the B31.1 Code 12

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l Equation 4.1 is the standard deadweight allowable stress equation per the B31.1 Power Piping Code. In equation 4.2, S is the basic allowable material stress per the B31.1 Power piping Code which is the lesser of 5/8 Sy (2/3 Sy in later code editions) or Su/4..

The majority of the piping under review is A-106B Carbon steel pipe which has S=15000 psi, Sy=35000 psi and Su=60000 psi. Therefore Equation 4.2 limits the Pressure +

Deadweight + Seismic Inertial Stresses to less than 1.03 Sy which insures elastic behavior. Equation 4.3 addresses self-limiting, secondary stress, where S for Carbon A

steel pipe is approximately 1.5 S which is approximately 22,500 psi, and therefore 2.0 SA

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is approximately 1.2 Sy.

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The piping support acceptance criteria used in the worst case support evaluation is as follows:

(a) Structural Steel DWT+TH s 1.0 AISC Allowable (4.4)

DWT+TH+SSE (Inertia and SAM) s 1.7 AISC Allowable (4.5) l (b) Component Supports

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DWr+TH s 1.0 ANSl/ MSS SP-58 Allowable (4.6)

DWT+TH+SSE (inertia and SAM) s 1.7 ANSI / MSS SP-58 Allowable (4.7)

This willinsure that the maximum stresses in the support members are at or slightly less l

than the material yield stress. In many of the MNGP calculations a factor of 1.6 was used in lieu of 1.7. This adds additional conservatism to the calculations and support l

evaluations. The 1.7 is based on the Part 11 allowables of the AISC Steel Construction Manual.

4.2 Condenser l

The seismic adequacy of the Monticello condenser was verified by reference to the l

BWROG report on MSIV leakage [1]. In Appendix D of reference 1, the seismic demand

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at earthquake experience sites with condensers was compared to seismic demand at GE BWR sites including Monticello. Condensers of similar configuration to Monticello experienced strong motion in excess of the Monticello design basis earthquake without failure. Reference 1 concluded that a condenser failure from a design basis earthquake at any GE BWR site was highly unlikely. In addition, the adequacy of the Monticello I

specific condenser configuration was verified by a comparison of the Monticello l

condenser to the earthquake experience data and by an evaluation of the Monticello j

condenser anchorage capacity.

4.3 Related Equipment Capacity The seismic adequacy of related equipment was verified using the GlP methodology as detailed in reference 2. Seismic capacity, caveat compliance, anchorage, and seismic spatial interaction concems were addressed. The GlP Bounding Spectrum that was obtained from earthquake experience data was used to establish seismic capacity of all related equipment.

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The majority of the related equipment are valves located at the lower elevations. Valve operability is not a concem for Monticello because all of the valves in Table 2-2 are not required to reposition to establish the leakage path or fail safe with respect to the leakage path. Since there is no reliance on standby power, none of the motor-operated valves were credited for operation.

4.4 Related Building Capacity The equipment and piping are confined to three buildings: the Reactor building, the Turbine building and the Recombiner building. The Reactor building is a Class 1 structure and has been designed to withstand the earthquake loads associated with the Montice!!o SSE. The Recombiner building was designed and built for seismic Class I conditions; however, the desion criteria for this building was later downgraded to Class 11 in accordance with Regulatory Guide 1.143. See Section 12.2.2.9 of the MNGP USAR

[6]. Portions of the Turbine building are also Class I (e.g., switchgear room) and have been designed to withstand the effects of the SSE where the applied accelerations are those from equivalent elevations of the Reactor building. See Section 12.2.1.9 of the j

MNGP USAR [6].

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1 The Reactor building equivalent elevation accelerations were used because an explicit dynamic model of the Turbine building was not developed. The Class I portions of the Turbine building are within the reinforced concrete structure of the building.

Consequently,',,4e reinforced concrete portion of the structure may be considered to be designed to Class I requirements even though the USAR only designates specific rooms 4

and areas as Class 1. All of the piping is located within the concrete portion of the Turbine building. The equipment is located in the concrete portion of the Turbine building with the exception of a few instruments which are located at the operating floor of the Turbine building (elevation 951'). The Turbine building above elevation 951'is a steel superstructure and is classified as Class ll; however, the superstructure was also seismically evaluated for the Reactor building SSE equivalent elevation accelerations.

j See Section 12.2.1.4 of the USAR [6].

4.5 Seismic Demand Allitems in the leakage pathway were evaluated for the SSE demand. The SSE ground response spectrum is identified in the MNGP Updated Safety Analysis Report. The MNGP SSE ground response spectrum is shown in Figure 3-1. The corresponding SSE horizontal peak ground acceleration (PGA) is 0.12g. The vertical demand was taken as 2/3 of the horizontal demand. The sections below describe SSE input for equipment in the leakage path.

4.5.1 Piping Seismic Demand Comparison of Demand To Experience-Based Capacity Spectrum The majority of the piping is located in the Turbine Building, Recombiner Building or buried. A small amount of the piping is located in the Reactor Building including the Steam Tunnel. The demand spectrum for piping in the Turbine Building, Recombiner Building, and buried piping is the 5% damped MNGP SSE design basis ground 14

Response Spectrum (Figure 3-1). Based on the comparison of experience based spectra contained in References 1 and 8, the capacity spectra all envelop the MNGP SSE ground spectrum with significant margin. The demand spectrum for piping in the Reactor Building was the 5% damped amplified floor response at the applicable elevation.

j Limited Analytical Reviews of Piping For limited analytical reviews of piping in the Turbine Building and the Recombiner building (all of which is less than 40' above grade) when dynamic analysis is applied, the horizontal piping demand is based on the 5% damped MNGP ground response spectrum l

shown in Figure 3-1 multiplied by a factor of 1.5. This method for estimating median-centered amplified floor spectra was used because amplified floor response spectra fcr these buildings at Monticello does not exist. The vertical demand is 2/3 of the horizontal demand. The resulting spectra were considered to be acceptable for the following reasons.

(a) The ground response spectrum is the licensing basis spectrum for the plant.

(b) -The piping which is located at elevations less than 40' above grade is in a l

concrete shear wall building, and the largest majority of this piping is below grade near the building foundation. Consequently, no significant building amplification of the design basis ground response spectrum would be anticipated.

l (c)

The Monticello floor spectra are classified as " Conservative Design" spectra by the staff [4).

For limited analytical reviews of piping systems when static analysis techniques are applied, the demand static load coefficient was 1.5 times the peak of the ground i

response spectrum in the horizontal direction and 1.5 times two-thirds of the peak of the l

ground response spectrum in the vertical direction.

l For piping in the reactor building the horizontal demand was based on the applicable 5%

damped amplified floor response spectrum and the vertical demand was 2/3 of the l-horizontal demand.

Limited Analytical Review of Buried Piping System For the evaluation of buried piping systems, the seismic demand is the design basis SSE l

ground response spectrum.

i Worst Case Support Reviews Seismic loads for use in worst case support reviews are determined as follows.

I (a) The span length of piping which would be expected to be restrained by the support in question was determined. This span length included an additional equivalent length of piping for included valves, or other in-line components.

15

(b) The total weight per unit length of piping considering pipe material weight, fluid weight, insulation weight, and any other weights in the piping system was 1

determined.

(c)

For determination of horizontal loads the value determined in (b) was multiplied by the peak of the applicable horizontal' response spectrum. For verticalloads 2/3 of the horizontal value was used. The applicable horizontal spectrum for all piping except that in the reactor building was 1.5 times the 5% damped ground response spectrum. For the reactor building, the applicable amplified floor response spectra was used.

4.5.2 Condenser Demand Spectra The Monticello condenser is located below grade at the lowest level of the Turbine Building (Elevation 911). The applied seismic demand was the SSE ground spectrum shown in Figure 3-1.

4.5.3 Related Equipment Demand Spectra Applied seismic demand for related equipment is based on the SSE ground spectrum shown in Figure 3-1 and the corresponding Floor Response Spectra (FRS). Consistent with the Monticello USAR [6), the Reactor Building FRS at an equivalent elevation is used to define the FRS for equipment in the turbine and recombiner buildings. These FRS were also used for USI A-46 resolution and were judged to be " conservative design" spectra when used with the GIP [4]. In addition and consistent with the GIP methodology,1.5 times the ground spectrum was optionally used as " realistic, median centered" demand for some equipment items meeting the GlP 40-foot-above-grade elevation limitation and the 8 Hertz lower bound frequency limitation. This was only done for equipment at or below grade. As with the piping, the largest majority of the equipment is located at the lowest elevations in the buildings.

i i

16

i

.4 5.0 Summary of Seismic Evaluation Results 5.1 Piping and Supports 5.1.1 Results Summary The piping material data, size, and schedules were obtained from piping and instrument diagrams (P&lDs) and line specifications. The line specifications also provide the design pressure and temperature data. Exceptions to the above were the GE supplied Steam Seal System and Moisture Separator Systems. Material and pipe size data for this

- system was taken from GE documents. The main steam lines between the MSIVs and the main turbine have been previously evaluated to meet the requirements of Class I loading which includes SSE loads.

j

. The walkdowns evaluated the seismic capacity of the subject piping system. As part of the walkdown, pipe supports, equipment supports and other modifications to reduce the seismic vulnerability of piping systems being screened were specified. These i

modifications were then considered in the evaluation of the acceptability of the piping systems. If necessary a detailed evaluation and verification calculation was conducted for the as-built modifications.

Worst case supports were identified, and detailed evaluations were conducted for these supports.- Rod hangers susceptible to fatigue failure, "hard spot" short rod hangers, and

)

U-bolts subjected to significant lateral loads were identified. Detailed evaluations were conducted to evaluate both the fatigue capacity of the rod hangers and the lateralload capacity of the U-bolts. See section 5.1.4 for a summary of these qualifications.

i The downstream side of the steam seal system was determined to be the worst case piping system based on the size of the system and its support configuration. For this system a detailed analysis using the criteria of ASME BPVC, Appendix N was conducted. In addition, limited analytical reviews were conducted for portionc of other piping systems which could be considered outside the screening criteria, which involved 1

complex spatial interactions, or for which a highly accurate prediction of piping support loads was required. One worst-case analytical review was conducted for all buried j

piping systems. See section 5.1.3 for a summary of these analyses.

5.1.2 Correlation with the Piping Experience Data After completion of the piping system walkdowns, evaluations were conducted to insure that the Monticello piping systems fall within the range of the piping systems

.which constitute the experience data.

Piping Sizes Table 5-1 presents a summary of the various piping, sizes, schedules and D/t ratios for each of the walkdown packages. Table 5-2 presents a general summary of the

- same data for the piping systems which constitute the experience data. More

- detailed summaries of the piping and the associated experience data are contained in Reference [1)). Table 5-3 presents a comparison of the D/t ranges of the

- 17 1

_.____________.________.____m_

__b

O I-Monticello piping to the experience data piping. The Monticello piping systems in the leakage path are enveloped by the experience data with the following j

exceptions.

.1.

The experience data does not specifically identify the existence of 3-1/2" and 5" diameter piping.

2. The Monticello 1" piping has lower bound D/t of 4 versus 5 in the experience data.

l

3. The Monticello 24" piping has lower bound D/t ratio 20 versus 23 in the experience data.

4. The 18" Monticello piping has an upper bound D/t ratio 48 versus 43 in the experience data.

For items (2) and (3), these lower D/t ratios are due to the use of thicker wall piping which would be stronger and have higher capacity than the experience data piping and therefore are not a concem. For (4), the exceedance is only 12 percent which is less than typical piping system fabrication tolerances. Therefore, this piping is adequately represented in the experience data. The 31/2" diameter piping and the 5" diameter, although not explicitly in the database, are enveloped by larger and smaller sizes. In addition, the 5" and 31/2" piping is in the. steam seal system that was analyzed in detail. Therefore, this pipirg is adequately enveloped by the experience data and the supporting analysis.

i Materials i

I 1

Table 5.4(a) provides a summary of the allowable stress capacity of the l-predominant piping materials of the experience data piping. Table 5.4(b) provides a similar summary for the Monticello piping. These tables demonstrate that the Monticello piping in leakage path is adequately represented in thn experience data piping.

I Support Spans

- Table 5.5 provides a summary of minimum and maximum ratios of the actual vertical support spans to the suggested ANSI B31.1 deadweight spans and the actuallateral support spans to the suggested ANSI B31.1 spans. Table 5.6 provides the suggested B31.1 deadweight support spans. Figures 5-1 through 5-4 compare the Monticello piping maximum span ratios, Vertical Support Ratio (VSR) and Lateral to Vertical Support Span Ratio (LVSSR) to the experience piping span ratio data.

These figures demonstrate that the Monticello piping support spans are well represented and adequately enveloped by the piping experience data.

5.1.3 Summary of the in-depth Piping Analyses This section provides a summary of the simplified and detailed piping analysis which were conducted for selected systems in the MSIV leakage path. Detailed dynamic computer based piping analyses were conducted for several piping systems. The criteria used in these piping evaluations and qualifications are given in section 4.1.2.

I Table 5.9 provides a summary of these analyses and the associated bases.

18

4 In addition to detailed dynamic piping analyses described in Table 5.9, localized equivalent static analyses were used to (1) evaluate SAMs, (2) evaluate spatial interaction concerns (3) evaluate localized areas of seismic vulnerability and (4) to determine loads used in the detailed support evaluations. Table 5.10 provides a summary of the equivalent static analyses conducted.

The steam seal discharge system piping was selected as the worse case piping system and required a detailed analysis. This was based on several factors including amourt of piping, pipe size variety, flexibility, and large in line equipment. judgment with the following considerations.

A detailed enveloping dynamic analyses was conducted for buried portions of the piping systems contained in six of the walkdown packages. These analyses included Soil Structure Interaction effects for the Turbine, Reactor, and Recombiner Buildings and evaluated both displacement effects and wave passage effects.

5.1.4 Summary of Detailed Support Qualifications Detailed Support Qualifications were based on identifying or establishing worse case supports during the walkdowns. The basis for the determination of these worst case supports included the following concerns.

(1) Short, fixed, or hard spot rod hangers that were judged to be susceptible to fatigue failure during a design basis SSE event.

(2) U-bolts susceptible to significant lateralloads. In many cases a system may contain multiple U-Bolts that could experience significant lateral loads. In such cases one or two enveloping evaluations for such a system were conducted.

(3) Supports that were judged to be the most susceptible to failure during a design basis seismic event based on field review.

(4) Supports on piping systems for which detailed seismic analyses were conducted.

Table 5.11 provides a summary of the number of supports subjected to detailed analytical reviews and the basis of these reviews. These supports represent approximately 15% of the support population in the MSIV leak path. In addition, these supports are most susceptible to failure during a design basis seismic event. By demonstrating the acceptability of these supports, it is reasonable to assume that the supports for the MSIV leak path piping has adequate seismic capacity.

5.1.5 Results i

The results and outlier resolution for piping and supports is listed in Table 5-8.

j i

l l

19

5.2 Condenser T'able 5-7 lists design data for the Monticello condenser and for the two experience data

. sites listed in Reference [1], Appendix D, Table 4-3 (Moss Landing 6 & 7, and Ormond

- Beach 1 & 2).. The Monticello condenser design data is similar to or bounded by data for j

the two experience data sites.- The Monticello SSE ground spectrum, which is the l

demand spectrum for the condenser, is enveloped by the Moss Landing and Ormond l

Beach spectra. The Monticello condenser design data is also well represented by the y

data presented in Reference [1], Appendix D, Table 4-3. The comparison verifies that j

the results of the Reference [1] evaluation for structural integrity are applicable to the Monticello condenser.

The Mont!Oe!!O 00ndenser ancherege 00n0!ct Of cight guided supporte viith enc cupport located at each corner of the two condenser shells. At each support, the condenser base bears again.=t a steel plate shear lug that is welded to an embedded solo plate.

The shear lugs rigidly resist lateral loads but are erranged to allow thermal growth.

Three 1.75 inch diameter cast-in-place anchor bolts are also located at each support (24 total). These bolts resist vacuum uplift loads. Companion bolt holes in the condenser base are 2.75 inches in diameter to allow for thermal growth. Figures 5-5 and 5-6 show guided support layout and details.

By Appendix D of Ref.1, GE evaluated lower and upper bound anchorage capacities of experience data and GE BWR condensers. For this evaluation, two capacity levels specific to the Monticello condenser were determined by detailed calculation for rigid and i

ductile beh0V! r. Cepecitico vicre derived from equet!0ne for 00pcciti 0 Of ench0rege elements defined in codes such as A!SC Manual of Steel Construction, ACl-349.

Capacities were defined in terms of allowed lateral acceleration. The calculations conservatively assume that cast-in-place bolts will not resist load in combination with the shear lugs. This is conservative because the condenser has oversized bolt holes and l~

the potential non-ductile failure of shear lugs.

.l For Monticello, a rigid-behavior anchorage capacity was obtained by crediting only the

]

shear lug load path at a support. Based on a detailed 1 valuation, the rigid-behavior capacity of the condenser anchorage was determineo to be 0.15g. The capacity is controlled by the direction transverse to the turbine axis. The shear lug load path capacity parallel to the turbine axis is 0.16g. A ductile-behavior anchorage capacity was obtained by crediting only the cast-in-place anchor bolts. The shear lug load path was assumed to fail in a brittle manner prior to bolt engagement and is given no credit in the ductile-behavior calculation. Based on detailed evaluation, the ductile-behavior capacity of the condenser anchorage transverse to the turbine axis was determined to be 0.24g.

j Parallel direction capacity is similar to transverse direction capacity. The rigid-behavior capacity of 0.15g exceeds the SSE PGA of 0.12g. The condenser shells are squat steel plated box struMures with substantial intemal stiffening, and the condenser is considered to be effectively rigid. Therefore rigid-behavior capacity exceeds SSE demand of 0.12g.

The Monticello lower and upper bound shear areas for the transverse direction are 0.000078 and 0.00021 square inches per pound respectively. The values for the parallel direction are 0.00010 and 0.00023 square inches per pound respectively. These values 20 l

I

are above corresponding vclues for the experience data sites shown in Figures 4-10 and 4-11 of Appendix D of Ref.1 The comparison of condenser data and the anchorage capacity evaluations demonstrates that the conclusions presented in Reference [1], Appendix D can be applied to the Monticello condenser. That is, a failure and significant breach of the condenser pressure boundary in the event of a design basis earthquake is highly unlikely and contrary to the experience data.

The condenser was also subject to a walkdown inspection which was summarized in a l

Screening Evaluation Work Sheets (SEWS). Some surface cracking of embedment j

i grout was observed at support locations. The pondenser was declared an outlier l

pending repair of the grout. This grout was repaired during the recent refueling outage at i

Monticello.

5.3 Related Equipment The condenser and the majority of related equipment were walked down. A Screening Evaluation Work Sheet (SEWS) was completed for each item. Each SEWS contains a capacity versus demand comparison, a checklist of bounding spectrum caveats, an anchorage review checklist, a spatialinteraction checklist, notes, and attached pictures (if available). The SEWS identify the determination of whether the item is acceptable or is an outlier and are signed by the SRT. The list of related equipment is provided in Section 2. Table 5-8 contains a list of equipment outliers and the associated resolution.

The majority of the related equipment are valves. All valves were found to meet GIP screening criteria. Valve operability is not a concern because all of the valves in Table 2-2 are passive in the case of motor-operated valves, or fail safe as in the case of air-and solenoid-operated valves.

1 l

1 l

1 l

21 i '

9 Table 5-1: Summary of Piping Properties for the Monticello Leakage Path Piping Walkdown Pipe Size Pipe Pipe Pipe ODit Material Package NPS (in)

Schedule OD (in)

Wall (in)

ASTM /ASME Designation 2913-1 6

80 6.625 0.432 15 A106B 3

80 3.5 0.3 12 A1068 1

160 1.315 0.25 5

A106B 2913-2 10 80 10.75 0.5Sa 18 A106B 2

160 2.375 0.344 7

A106B 1-1/2 160 1.9 0.281 7

A106B 2913-3 18 80 18 0.938 19 A672, Gr. 70 10 80 10.75 0.593 18 A672, Gr. 70

[- 13-4 4

80 4.5 0.337 13 A106B 2913-5 16 STD 16 0.375 43 A53B/A106B I

12 STD 1r. 75 0.375 34 A53B/A106B 10 STD

• 0.75 0.365 29 A53B/A106B 2913-6 16 STD 16 0.375 83 A53B/A106B 12 STD 12.75 0.375 34 A53B/A106B 10 STD 10.75 0.365 29 A53B/A106B 2913-7 6

40 6.625 0.28 24 A53B/A106B 3/4 80 1.05 0.154 7

A53B/A106B 2913-8 6

40 6.625 0.28 24 A53B/A106B 3/4 80 1.05 0.154 7

Af3B/A1063 2913-9 3

160 3.5 0.438 8

A10SB 3

STD 3.5 0.216 16 A53B/A106B 2

160 2.375 0.344 7

A106B 1

BOS 1.315 0.179 7

304SS 1

80 1.315 0.179 7

A53B/A106B 2913-10 6

80 6.626 0.432 15 CS'1) 2913-11 18 STD 18 0.375 48 A53B/A1068 12 STD 12.75 0.375 34 A53B/A106B 10 40 10.75 0.365 29 A53B//.106B 8

40 8.625 0.322 27 A53B/A106B 2913-12 6

160 6.625 0.718 9

A106B 6

120 6.625 0.562 12 A106B 6

80 6.625 0.432 15 A106B 4

80 4.5 0.337 13 A106B 2913-13 24 80 24 1.22 20 SA1068 6

120 6.625 0.562 12 SA106B 4

120 4.5 0.438 10 SA106B 2913-14 6

120 6.625 0.562 12 A106B 4

120 4.5 0.438 10 A106B 2913-15 3

160 3.5 0.438 8

A106B 2

XXH 2.375 0.436 5

A106B 1

XXH 1.315 0.358 4

A106B 2913-16 4

120 4.5 0.438 10 A106B 3

160 3.6 0438 8

A106B 22

I Table 5-1: Summary of PipMg Properties for the Monticello Leakage Path Piping Walkdown Pipe Size Pipe Pipe Pipe OD/t MateriaI Package NPS (in)

Schedule OD (in)

Wall (in)

ASTM /ASME Designation 2913-17 1

160 1.315 0.25 5

A106B 2913-18 1

160 1.315 0.218 6

A106B 1

160 1.315 0.218 6

A312-304L 1/2" Tubing N/A 0.625 0.049 13 A312-304 1/2" Tubing N/A 0.625 0.049 13 A376-316 2913-19 1

80 1.315 0.179 7

A106B 1

80 1.315 0.179 7

A312-304 1/2" Tubing N/A 0.625 0.049 13 A312-304 1/2" Tubing N/A 0.625 0.049 13 A376-316 2913-20

1. i XXH 1.315 0.358 4

SA106B 1/2 XXH 0.84 0.294 3

SA106B 29:3-21 4

120 4.5 0.438 10 SA106B 2913 1 80 1.315 0.179 7-A106B 3/4 160 1.05 0.219 5

A106B 2913-23 3

40 3.5 0.216 16 A53B/A106B 3

40 3.5 0.216 16 A312-304L 2913-24 1

160 1.315 0.25 5

A106B 2913-25 1-1/2 160 1.9 0.281 7

A106B 1

160 1.315 0.25 5

A106B 2913-26 2

40 2.375 0.154 15 SS(2) 1-1/2 40 1.9 0.145 13 SS(2) 1 40 1.315 0.133 10 SS(2) 3/4 40 0.75 0.113 7

SS(2) 1/2" Tubing N/A 0.625 0.035 18 SS(2) 2913-27-1,-2,-3 16 40 16 0.5 32 CS(1) 12 40 12.75 0.406 31 CS(1) 10 80 10.75 0.593 18 CS(1) 10 40 10.75 0.365 29 CS(1) 8 40 G.625 0.322 27 CS(1) 6 80 6.625 0.432 15 CS(1) 6 40 6.625 0.28 24 CS(1) 5 80 5.563 0.375 15 CS(1) 5 40 5.563 0.258 22 CS(1) j 4

40 4.5 0.237 19 CS(1) 3 */2 80 4

0.3 13 CS(1) 3 40 3.5 0.216 16 CS(1) 2 40 2.375 0.154 15 CS(1) 1-1/2 40 1.9 0.145 13 CS(1) 1 40 1.315 0.133 10 CS(1) l l

2913-27-4 1-1/2 40 1.9

.145 13 CS(1) 3/4 40 1.050

.113 9

CS(1) i 2913-28 1-1/2 160 1.9 0.281 7

A106B 1

160 1.315 0.25 5

A106B i

2913-29 14 STD 14 0.375 37 A53B/A106B 10 40 10.75 0.365 29 A53B/A106B 23 w _- __-_ _

a s

Table 5-1: Sur, mary of Piping Properties for the Montice!!o Leakage Path Piping Walkdown Pipe Size Pipe Pipe Pipe OD/t Material Package NPS (in)

Schedule OD (in)

Wall (in)

ASTM /ASME Designation 3

40 3.5

.216 16 A53B/A106B 1-1/2 80 1.9

.2 10 A53B/A106B 2913-30 3/4 XXH 1.050

.308 3.5 A106B 3/4 XXS 1.050

.308 3.5

' B42-Copper 1/2 XXS

.840

.294 3.0 B42-Copper 1/4 XXS

.540

.119 4.5 B42-Copper 2913-31 1

40 1.315

.133 10 CS(1) 2913-32 8

100 8.625

.593 14.5 A106B 2913-33 6

40 6.62o

.2e 24 A106B 2

80 2.375

.218 11 A106B 2913-34 1-1/4 40 1.660

.140 12 A106B 5

60 5.563

.258 21.5 A106B 5

40 5.563

.258 21.5 Cast iron (3) 2913-35 2

80 2.375

.218 11 A106B 1

80 1.315

.179 7.5 A106B 1/2 80

.840

.147 6

A106B 2913-36 2

80 2.375

.218 11 A106B 2913-37

.fi~

80 1.050

.154 7.0 A106B 1/2 - Tubing

.065" Wan

.5

.065 7.7 SS(2) 5/8 - Tubing

.065" Wall

.625

.065 9.5 SS(2) 5/8 - Tubing

.065" Wall

_6?5

.065 95 A213-3041.

2913-38 3/4 80 1.050

.154 7.0 A106B /A312-304L 3/8 - Tubing

.065

.375

.065 5.8 SS(2) 2913-39 6

.375 Wall 6.625

.375 17.5 A106B/A312-304L 3

40 3.5

.216 16 A106B/A312-304L 3/4 80 1.050

.154 7 A106B/A312-304L 2913-40 3/4 160 1.050

.218 5

A100B 1/2 - Tubing

.065" Wall

.5

.065 7.7 SS(2)

(1) CS = Carbon Steel Pipe; (2) SS = Stainless Steel Pipe (3) Cast iron was Fittings Only and Limited Analytical Review was Conducted to Demonstrate Acceptability 24 L_____._______.___.____.___..

Table 5-2: Seismic Experience Piping Data [1]

Pine Size Pipe Fipe Pipe OD/t Plant NPS (in)

Schedule OD (in)

Wall (in)

Valley Steam Plant 24 20 24.00 0.375 64 Units 1 and 2 20 20 20.00 0.375 53 18 30 18.00 0.437 41 16 30 16.00 0.376 43 14 30 14.00 0.375 37 12 40 12.75 0.406 31 12 30 12.75 0.33 39 10 160 10.75 1.125 10 8

160' 8.6250 0 906 10 6

40 6.6250 0.26 24 4

160 4.5000 0.531 8

4 40 4.5000 0.237 19 3

160 3.5000 0.437 8

3 80 3.5000 0.3 12 3

40 3.5000 0.216 16 2

160 2.3750 0.343 7

2 40 2.3750 0.154 15 11/2 160 1.9000 0.281 7

1 1/2 40 1.9000 0.145 13 1

40 1.3150 0.133 10 3/4 160 1.0500 0.218 5

3/4 40 1.0500 0.113 9

Moss Landing 16 N/A 16.00 1.394 11 Units 1,2, & 3 12 N/A 12.75 1.148 11 Moss Landing 24 40 24.00 0.687 35 Units 4 & 5 24 N/A 24.00 1.066 23 N/A 18.30 2.287 8

J 16 40 16.00 0.5 32 16 N/A 16.00 0.9C2 18 j

N/A 13.20 1.668 8

Moss Landing 30 N/A 30.00 0.632 47 Units 6 & 7 26 N/A 26.00 1.128 23 18 N/A 18.00 3.444 5

12 N/A 12.75 2.444 5

l 12 N/A 12.75 0.601 21 Ormond Beach 30 N/A 30.00 1.298 23 Units 1 & 2 30 N/A 30.00 0.719 42 i

21 N/A 21.00 3.793 6

25 i

i

Table 5-2 Seismic Experience Piping Data [1]

Pipe Size Pipe Pipe Pipe OD/t Plant NPS (in)

Schedule OD (in)

Wall (in)

Humboldt 12 80 12.75 0.687 19 Unit 3 10 80 10.75 0.593 18 6

80 6.625 0.432 15 9

El Centro Steam Plant 20 STD 20.00 0.375 53 18 160 18.00 1.7810 10 18 XS 18.00 0.5000 36 18 STD' 18.00 0.3750 48 14 40 14.00 0.4370 32 14 STD 14.00 0.3750 37 12 160 12.75 1.3120 10 12 STD 12.75 0.3750 34 10 40 10.75 0.3650 29 8

160 8.625 0.9060 10 8

120 8.625 0.7180 12 8

40 8.625 0.3220 27 6

120 6.625 0.3620 12 6

40 6.625 0.2800 24 4

80 4.500 0.3370 13 4

40 4.500 0.2070 19 3

160 3.50 0.4370 8

3 80 3.50 0.3000 12 3

40 3.50 0.2160 16 2

160 2.375 0:3430 7

2 80 2.375 0.2180 11 l

2 40 2.375 0.1540 15 1 1/2 160 1.90 0.2810 7

1 1/2 80 1.90 0.2000 10 11/2 40 1.90 0.1450 13 1

80 1.315 0.1790 7

1 40 1.315 0.1330 10 3/4 80 1.050 0.1540 7

3/4 40 1.05('

O.1130 9

26 l

L___-__-__-______

f Table 5-3: Dit Range Comparison Nominal Pipe Size Monticello Experience Data (NPS)(ID)

Piping Dit Ranges Piping Dit Ranges 3.5-9 5-9 1

4-10 5-20 1-1/4 12 1%

7-13 7-13 2

5-15 5-15 3

8-16 8-16 l

3 1/2 13 4

10-19 8-19 5

15-22 6

9-24 9-24 8

27 10-31 10 18 10-29 12 31-34 10-34 14 37 32-37 16 32-43 11-43 18 19-48 5-41 24 20 23-35 Table 5-4(a): Predominant Materials of the Experience Data

~

Mater! !

ANS! B31.1 Allowable Strsss, psi ASTM Designation A53B 15000 A106 B 15000 A335 14000 A120 (1)

A139 12000 (1) Stress allowables not provided by B31.1. E31.9 provides an allowable stress value of 10000.

Table 5-4(b): Predominant Materials of Monticello Piping Material ANSI B31.1 Allowable Stress, psi ASTM Designation A53B

'15000

~

A106B 15000 312-304 15900 376-316 17000 312-304L 13700 B42 - Copper 6000 )

0 (1) This is the lowest va:Je for B42 Copper given in the 931.1 Code.

27 L____________.___________________..__________________._________________________________________

i

l Table 5-5: MNGP Span Ratios in Comparison to ANSI B31.1 Suggested Deadweight Spacing Pl:e Type Maximum Minimum Maximum Minimum Walkdown i

Package SB = Small Vertical Vertical Lateral Lateral Bore (<2.5")

Support Support Support Support LB= Large Actual Actual Actual Actual Bore (>2.5")

Spacing Spacing Spacing Ratio Spacing

[ Based on Ratio to Ratio to to B31.1 Ratio to Predominant B31.1 B31.1 Suggesied B31.1 Pipe Size]

Suggested Suggested -

Support Suggested

}

Support Support Spacing Support l

Spacing (2)

Spacing (LVSSR-Max)

Spacing i

(2)

(LVSSR -

l Min) l 2913-1 LB 1.5 1

4.2 1

2913-2 SB 1.5

.5 3

.5 2913-3 LB 1

1 3

1 29134 LB 2.2 (1) 1.5 7

1 2913-5 LB 1.5 1

3 2

)

2913-6 LB 1.5

<1 2

2 2913-7 LB 1

.5 5

N/A 2913-8 LB 1

.5 5

N/A 2913-9 LB 1

.75 6.2 5.5 '

SB 1

.75 2

1 2913-10 LB 1

N/A 1.5 1

2913-11 LB 1

1.25 5.25 2

2913-12 LB 1.5

<1 2.75 1

2913-13 LB (3)

(3)

(3)

(3) 2913-14 LB (3)

(3)

(3)

(3) 2913-15 SB 1

1.5

<1 2913-16 LB 2

2.5 1

2913-17 LB 1.5

<1 6

<1 2913 18 SB 1.5 1.3 5.5 1.3 2913 19 LB (3)

(3)

(3)

(3) 2913-20 SB 1.5

<1 2

1 2913-21 LB 1.5 1

1.5 1

2913-22 SB 1

.5 1.5 1

2913-23 LB 1.5 1

5 5

2913-24 SB 1

1 1.5 1

2913-25 SB 1

1 2

1 2913-26 SB 1.5

<1 1.5

<1 2913-27-1,-

LB,SB (4)

(4)

(4)

(4) 2,-3 2913-27-4 SB 1.5 1

5 2

2913-28 SB 1

1 3

1 2913-29 LB 2

<1 2.7 2.7 l

2913-30 SB 1

<1 2

1 2913-31 SB 1.5 1

5.0 2

2913-32 LB 1

1 3

2 2913-33 LB,SB i

1 2

1 2913-34 LB,SB 1

1 2

1 2.8 l

1 l

Table 5-5: MNGP Span Ratios in Comparison to ANSI B31.1 Suggested Deadweight Spacing I

Walkdown Pipe Type Maximum Minimum Maximum Minimum Package SB = Small Vertical Vartical Lateral Lateral Bore (<2.5")

Support Support Support Support LB1 Large Actual Actual Actual Actual Bore (>2.5")

Spacing Spacing Spacing Ratio Spacing

[ Based on Ratio to Ratio to to B31.1 Ratio to Predominant B31.1 B31.1 Suggested B31.1 i

Pipe Size]

Suggested Suggested Support Suggested Support Support Spacing Support Spacing (2)

Spacing (LVSSR-Max)

Spacing (2)

(LVSSR -

Min) 2913-35 SB 1.5 1

4 2

2913-36 SB 1

1 3

1 2913-37 SB 2

1 4

1 2913-38 SB (3)

(3)

(3)

(3) 2913-39 LB,SB (3)

(3)

(3)

(3) 2913-40 SB 1

1 2

1 (1) These spans exclude considerate 0'1 of spring hangers.

(2) Spans include consideration of modified or added supports.

(3) These lines had obvious seismic design & short spans; accepted by inspection without detailed span evaluation.

(4) This was a worse case system and was qualified by detailed ana!ysis.

I I

I I

29

^

Table 5-6: Nominal Suggested Vertical Deadweight Spans per ANSI B31.1 Suggested B31.1 Deadweight Spans (ft) l Monticello Outside Pipe Water Service Steam. Gas or l

Nominal Pipe Diameter (in)

Air Size ** (in)

Service 3/4 1.050 6*

8*

l 1

1.315 7

9 1 1/2 1.900 9*

11*

2 2.375 10 13 3

3.500 12 15 3 1/2 4.000 11*

12*

4 4.500 14 17 f

5 5.563 16 19*

6 6.625 17 21

)

8 8.625 19 24 i

10 10.750 21*

26*

12 12.750 23 30 14 14.000 25*

33*

16 16.000 27 35 18 18.000 29*

37*

24 24.000 32 42

• Interpolated values - not given directly in ANSI B31.1.

" There are smail amounts of 1/2" piping and I/C tubing (1/8",1/4",1/2", 5/8" and 3/4") not presented in this table.

l l

i 30 1

Y -

i Table 5-7: Monticello Condenser Design Data Versus Experience Data [1]

Parameter I.

"o Moss Landing Ormond Beach 6&7 1&2 Manufacturer Worthington ingersoll Rand Southwestern Flow Type Single Pass Single Pass Single Pass Shell Dimensions HP: 40' x 30' x 65' x 36' x 47' 52' x 27' x 20' (L x W x H) 35' LP: 36' x 30' x 35' Tube Area per Shell HP: 210,000 ft" 435,000 ft' 210,000 ft' 2

LP: 189,000 ft Shell Material ASTM A285C ASTM A285C ASTM A285C Shell Thickness

% inch

% inch

% inch Operating Weight HP: 1,900,000 3,115,00b lbs.

1,767,000 lbs.

Ibs.

LP: 1,800,000 lbs.

Tube Material Type 304 S.S.

Al-brass 90-10 Cu-Ni Tube Size 1 inch 1 inch 1 inch Tube Length 36 to 40 feet 65 feet 53 feet 1 ube Wall I hickness 16 to 22 Bwg 16 Swg 20Hwg Number of Tubes 20,056 per shell 25,590 15,220 per shell Tube Sheet Material Munz Metal Munz Metal Munz Metal Tube Sheet Thickness 1% inch 1% inch 1% inch No. of Tube Support 13 per shell 15 14 j

Plates Tube Support Plate ASTM A285C not identified ASTM A285C Material Tube Support Plate 3/4 inch 3/4 inch 5/8 inch Thick.

Tube Support Plate 33 inches 48 inches 36 to 36.5 Spacing inches Waterbox Material ASTM A285C 2% Ni cast iron ASTM A285C ASTM A-48 CL 30 Waterbox Plate 3/4 inch N/A 5/8 to 1 inch i

Thickness Expansion Joint Rubber belt Rubber belt St. sts:-i Hot Well Capacity 43,000 gallons 20,000 gallons 34,338 gallons Hot Well Hold Time 2 min N/A N/A 31

l Table 5-8; Summary of Concerns and Resolution identifier Concerns Resolution

~

Package 2913-4 Spatialinteraction Loose equipment moved or restrained Package 2913-5 Loose hanger Hanger repaired Package 2913-4 (a) Broken U-Bolt (b) Missing U-(a) Replaced (b) Installed (c)

Bolts (c) Spatialinteraction Potential target conduits determined to be not required for normal or accident conditions Package 2013-11 (a) Lack of Lateral Restraint (b)

(a) Support modified (b) Repaired l

Loose rod hanger (c) Short rod (c) System qualified assuming this rod hanger (d)Poorly supported l&C line hanger failed (d) Reroute /resupport line Package 2913-12 (a) Loose U-Bolt (b) Loose rod (a) Repaired (b) Repaired (c) U-Bolt hanger added (c) Additionallateralsupport required q

l Package 2913-16 (a) Lack of lateral restraint (b)

(a) Support modified (b) Block wall Spatialinteraction braced Package 2913-19 (a) Sample Chamber Lacks Vertical (a) Support added (b) Bands and Support (b) Tubing could Fall From covers added to trays (c) Restraint Trays (c) Tubing needs added lateral / vertical restraint (2 places)

(d) Lead blocks restrained (d) Spatialinteraction for SV-2 and 17-104 Packago 2913-20 (a) Missirig U-Bolt (b) Spatial (a) U-Bolt installed (b) Block wall interaction braced Package 2913-22 Spatialinteraction Crane rail demonstrated to be seismically adequate Package 2913-24 (a) Lateral support required (b)

(a) Support added (b) Piping Short rod hanger qualified assuming hanger would fail.

Package 2913-26 (a) Lack of seismic support (b)

(a) Lino resupported for earthquake Loose rod hanger (c) Loose U-Bolt (b) Repaired (c) Repaired Package 2913-Lack of Lateral Support Two new supports added 27-1.

Package 2913-Lack of lateral support & spatial Seven new pipe supports added 27-2 interaction concems Package 2913-Lack of lateral support Three supports added 27-4 Package 2913-28 Missing support Support reinstalled E-2A, E-28, E-4 Anche ge Bracing was added to reduce anchor loads

~

T-33 Anchorage Bracing and anchors were added

~

V-813 Anchorage Plates added l

l 32 u----_--------_------------------_-----

Table 5-8 Summary of Concerns and Resolution identifie'r Outlier issue Proposed Resolution 17-116,17-104 Interaction Shield blocks restrained 2913-OSVS-1 Corrosion / Erosion Piping Systems are in the Erosion / corrosion Monitoring Program.

2913-OSVS-2 Possible Corrosion Piping Systems are in Erosion / Corrosion Monitoring Program.

2913-OSVS-3 Spatial Interaction Added Support to 14" Piping.

E-1 A, E-1B Cracked grout Repaired with high strength epoxy grout.

I 2913-27-4 Piping Overspans Added three supports I

2913-40 Inadequately supported Re-support thr tubing system i

33

l e

Table 5.9 - Summary of Detailed Analysis Conducted Walkdown Description Basis for Detailed Analysis Package No.

2913-27-1 Steam Seal System - Discharge Worse Case System Portion. All large bore piping (>2 in diameter) including all possible leak paths to the condenser.

Displacements at all small bore (2" and under) connections to the.!arge bore lines were determined and used in evaluation of SAM effects on the Small Bore Systems 2913-27-1 Steam Seal System - 2" Branch Line Did not meet screening criteria 2913-27-1 Steam Seal System - 2" Branch Lines Did not meet screening criteria Multiple Four Large Bore (30" & 36" diameter)

Spatial Interaction Concems with Moisture Separator Systems (from several piping systems in the leak path Moisture Separators to the Intermediate Stop and Control Valves) 2913-27-2 Steam Seal System - 2" Branch lines Did not meet screening criteria - two hard spot rod hangers did not pass rod fatigue review. Analysis armmed these hangers failed.

2913-27-2 Steam seal system - 2" Branch Line Spatial Interaction Concerns and two 12" Steam Bypass lines 2917-24 Steam Equalizing Line Determined Support Loads 2917-12 SJAE to Tank T72 Determine Anchor Loads 2917-36 SJAE Drain Lines Evaluate the effects of corrosion on a portion the piping system 2917-30 Oxygen Injection Piping Although th; Se was well supported, the ASTM Be n'terialis not represented in tne experience database of references [1]

2917-37 Steam Seal System 1/C tubing Did not meet screening criteria 34

Table 5.10 - Summary of Equivalent Static Analyses Conducted Based for the Equivalent Static Number of Equivalent Static Analyses Analyses conducted for this Reason Evaluate SAMs 2

Evaluated Spatial Interactions 8

Evaluate Local Vulnerabilities 2

Determined Support loads for Evaluation 5

1 Table 5.11 - Summary of the Detailed Support Qualifications Basis of the Qualification Number of Supports Evaluated Red Fatigue Concems 25 Lateral U-Bolt Concerns 23 Worse Case Support Reviews 44 Supports on systems subjected to detailed Analysis 30 Modified or Added Pipe Supports 31 Total 153 35

l i

Large Bore VerJeal Support Span Ratio of Monticello Subject Piping Compared to the Experience Data i

k 1

l l

E2 i

S.

0BPerience Data j

i o

g g Monticello Data 1

z l

m e

o o

j Y

V' s

VSR l

Figure 5-1 Small Bore Vertical Support Span Ratio of the Subject j

Monticello Piping Compared to the Experience Data E2

[

! OSPerience Data i

o MMonticello Data u

2 E

l e

e o

o

=

l 1

0>

1 I

t a

a VSR Figure 5-2 36

e Large Bore Piping Comparision of the LVSSR of the Subject Monticello Piping to the Experience Data 0 xperience Data E

$}

g Nbnticello Data 2

~

E

=

• U im im

%~

LO LD O

LD O

O O

V' W

d j

y d

d h'

h O

c 6

^

o n

4 d

LVSSR Figure 5-3 Small Bore Piping Comparision of the LVVSR of the Subject MonticelloPiping to the Experience Data lii>

O xperience Data E

M 8

E g Nbnticelb Data E

i z

lE im LO LD O

LD O

O O

v' 4

d 4

'A c

4 O

6 O

e n

4 6

LVSSR Figure 54 37

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v o,m

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nf l

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w t sy Figure 5-5: MNGS Condenser support layout from Worthington DR-127368 Rev. B 1

i 1

1 i

1 38

_l

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Figure 5-6: MNGS Condenser support details from Worthington DR 127368 Rev. B, Support B is similar to Support A 39'

s.

l 6.0 References

[1-] NEDC-31858P, Revision 2, General Electric, "BWROG Report for increasing MSIV t

L Leakage Ratio Limits and Elimination of Leakage Control Systems," September 1993, l

(principally Appendix D thereof).

[2] EPRl/SQUG, " Generic implementation Procedure (GlP) for Saismic Verification of Nuclear Plant Equipment," Revision 2, February 1992.

[3] " Supplemental Safety Evaluation Report No. 2 (SSER #2) on GIP-2, "USNRC, Washington, DC, May 22,1992.

[4] "Monticello Nuclear Generating Plant Verification of Seismic Adequacy of Mechanical and Electrical Equipment, Unresolved Safety issue A-46 (SQUG)," Northern States Power Company, November 1995.

[5] EPRI Report NP-5617 Volume 1 and 2," Recommended Piping Seismic Adequacy Criteria Based on Performance during and after Earthquakes," January 1988.

[6] Monticello, Updated Safety Analysis Report (USAR).

[7] SSRAP, "Use of Experience and Test Data to Show the Ruggedness of Equipment in Nuclear Power Plants, Rev. 4.0, February 1991.

[8] Safety Evaluation - Duane Arnold Energy Center - Amendment No. 207 to Facility Operating License No. DPR-49, February 22,1995.

l l.

40 t

'O l

TRANSMITTAL MANIFEST NORTHERN STATES POWER COMPANY NUCLEAR LICENSING DEPARTMENT MONTICELLO NUCLEAR GENERATING PLANT Demonstration of the Seismic Qualification of the MSIV Leakage Path at Monticello (TAC No. 96238)

Manifest Date: May 29,1998 Correr,wndence Date: May 29,1998 Manticello Intemal Site Distribution Special Instructions Kaleen Hilsenhoff.......USAR File.................Yes No_x_

j Stsve Ludders..........NRC Commitment.....Yes No_x_

Lits Imholte................Monti OC Sec...........Yes No_x_ - 12, No dist to OC members below if YES SAC Secretary...........Monti SAC................Yes No_x_ - 5 j'

Monticello internal Site Distribution:

l Monti Document Control File J C Grubb, NGSS. OC M F Hammer, Plant MGR, SAC, OC L L Nc!cn, GSSA, OC C A Schibonski, GSE, OC MGR MTC, OC E M Reilly, GSM, OC Monticello Oper Exp Coord i

J E Windschill, GSRS, OC A E Ward, NCD B D Dcy, GSO, OC Monti Site Lic File 1

Dennis Zercher NRC Resident inspector's Office Steve Hammer NSP Intemal Distribution G T Goering, Chairman, SAC W A Shamla, Dir Gen Qual Serv, SAC l

Communications Dept Yes No_x_

l Extemal NSP Distribution iDoc Control Desk, NRC Kris Sanda, State of Minn Regional Admin-lli,NRC J E Silberg T J Kim, NRR-PM, NRC

..vo

- =,

i M-1

____