ML20249C516

Provides Revised Info & Supersedes Re Demonstration of Seismic Qualification of MSIV Leakage Path. Rev to Verification Info & Seismic Calculations Included

ML20249C516
Person / Time
Site:	Monticello
Issue date:	06/15/1998
From:	Hammer M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20249C517	List: ML20249C516 ML20249C521 ML20249C522 ML20249C524
References
TAC-M96238, NUDOCS 9806300223
Download: ML20249C516 (47)

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'. )! .N  : . . Y Y b ?Y !I Y 0 Y-Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello, Minnesota 55362-9637 I

June 15,1998 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Seismic Verification 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 Concerning 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, " Demonstration of the Seismic Qualification of the MSIV Leakage Path at Monticello (TAC No. M96238)," May 29,1998 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 letters dated April 17 and May 29,1998 (Ref.

2), NSP provided supplementalinformation on the seismic verification of the MSIV leakage path-to the condenser. A meeting was held on June 4,1998 between the staff and NSP regarding Ref. 2. NSP stated that it would revise the submittal and provide calculations for the staff to review. Accordingly, this letter provides the revised information and supersedes Ref. 2 entirely.

Attachment 2 contains the revision to the verification information. Attachment 3 contains a package of seismic calculations.

This submittal supplements NSP's power rerate license amendment dated December 4,1997.

The changes to this submittal do not affect the demonstration of no significant hazards included in the power rerate license amendment. 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 I

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, d v t. ' ib 6n5/96 eb f pec J UCENSE)JOELGTUFFWSEIS4 DOC 9906300223 990615 PDR ADOCK 05000263 p FUR

c: Regional Administrator-Ill, NRC NRR Project Manager, NRC Sr. Resident inspector, NRC  ;

State of Minnesota, Attn: Kris Sanda J. Silberg, Esq.

Attachments Attachment 1 NRC Affidavit Attachment 2 Seismic Verification of MSIV Leakage Path Attachment 3 Seismic Calculation Package I

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UNITED STATES NUCLEAR REGULATORY COMMISSION I

NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Seismic Verification of the MSIV Leakage Path at Monticello (TAC No. 96238)

Northern States Power Company, a Minnesota corporation, by letter dated June 15, 1998 provides information regarding the seismic verification of the MSIV leakage path to the condenser for the Monticello Nuclear Generating Plant to a US Nuclear Regulatory Commission (NRC). This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY l

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[ COLLEEN A, HANNON By L as miyrusus.am mE00fA ,

Michael F. Hammer uposam ap. hse.  :

! Plant Manager Monticello Nuclear Generating Plant On this 15 day of 32 M before me a notary pubiic in and for said County, personally appeared Michael F. Hammer, Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, and that to the best of his knowledge, information, and belief the statements made in it are true.

M Colleen A. Hannon Notary Public - Minnesota Sherburne County

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Attachment 2 Seismic Verification of MSIV Leakage Path l

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.

l 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 leakage pathway includes leakage through the l MSIVs via the main steam piping and main steam drains to the condenser. l The methodology suggested in NEDC-31858P (Reference 1) was used to seismically evabate this pathway. This report will discuss the applicability of this methodology for l Monticello and how this methodology was used for the seismic evaluation of the l pathway. This report will summarize the seismic evaluation that was performed for the l piping and equipment in the MSIV leakage path for Monticello. The evaluation I 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.

l 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 j (SQUG) Generic implementation Procedure (GIP) described in reference 2 was used for seismic qualification of certain existing equipment in the MSIV leakage path.

2.0 Scope of Pipiag 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 linen 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 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 from 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 lines can be isolated by Motor Operated Valves (MOVs). Each MOV has a bypass line with a restricting orifice. Since the MOVs are not powered by essential power and they are normally closed valves, it is assumed that the l leakage will be through the MOV bypass lines via the restricting orifices. This provides a l passive pathway for the MSIV leakage to reach the condenser because no valve i I positioning or operator action is necessary to establish the pathway. Therefore, periodic testing to demonstrate valve operability is not required.

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l 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.

t Reactor Building Turbine Building Recombiner L

Building

- HPCI Room - Condenser Bay All areas l- - RCIC Room - Steam Jet Air Ejector (SJAE) Room

! -Torus Area - Mechanical Vacuum Pump Room

! - Main Steam - Condenser Bay to SJAE Room Pipe Tunnel Chase l

- Condensate Backwash Receiving Tank Room l j 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 2-1 identifies the piping packages.

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MAIN STEAM l StenniJet Air Ejectors i ISOLATION VALVES 4 t {

Recombiner +- TURBINE STOP & 1 COfGROL VALVES mm

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, 2 rm STEAM LINE DRAINS Figure 2-1 l l Monticello Main Steam Isolation Va Ive Leakage Pathways to the Main Condenser l l

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Table 2-1: MSIV Piping Package List Package Piping System Description Location i 2913-1 Main Steam Drain Condenser Bay / Steam Tunnel 2913-2 Main Steam Drains Condenser Bay 2913-3 Pressure Equalizing Lines Condenser Bay 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. Esay ,

2913-7 RV33-6 -HB lines SJAE/ Pipe Tunnel /Cond. Bay j 2913-8 RV34-6 -HB lines SJAE/ Pipe Tunnel /Cond. Bay 2913-9 MS to SJAE E-2A/E-2B SJAE/ Pipe Tunnel /Cond. Bay 2913-10 Air injector Piping SJAE Room 2913-11 T72 and T33 Tank Lines SJAE/MPV/ Hallway 2913-12 From SJAE to Tank T72 and Off Gas System 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 Tunnel /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 2913-23 Gland Blower Discharge Line Reactor Building 2913-24 MO-1739 Equalizing 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 l 2913-27-2 Steam Seal System, Section 2 Condenser Bay l 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 I 2913-36 SJAE Drain Lines SJAE Room 2913-37 Various I&C Lines SJAE Room / Condenser Bay 2913-38 V813 Tank Drain / Level Lines Condenser Bay 2913-39 Feedwater Heater Steam Trap Drain Lines Condenser Bay 2913-40 Misc Main Steam Drains and I&C Lines Condenser Bay l 5 l

Table 2-2: MSIV 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-10848 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-4164, CV-4165 HWC O2 FLOWTO RECOMBINER CONTROL VALVE E-1 A, E-1 B 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 MAIN HEADER PRESSURE EQUALIZER DRAIN '

MOIST-SEP MOISTURE SEPARATOR PCV-7489A, PCV-7489B A RECMB TRAIN OG INLET VALVES PCV-7496A PCV-7496B OFFGAS BYPASS RETURN TO CONDENSER PCV-7497A, PCV-74978 OG STEAM SUPPLY VALVES l PCV-7498A, PCV-74988 OG TRAIN STEAM SUPPLY VALVES 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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l i 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. i l Subsequent evaluations demonstrated that many non-seismically designed structures, I l

systems, equipment and components have substantial inherent seismic ruggedness.

The Seismic Qualification Utility Group (SQUG) was formed in 1981 after an agreement L with the NRC to develop alternative methods to resolve seismic safety issues for critical l

l 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 incluP3 the following objectives l Documentation of the most common causes of seismic damage or l operational difficulties in facilities that contain structures, systems, equipment and components similar to those in nuclear stations.

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

Identification of structures, systems, equipment and components that typically are not damaged in earthquakes much larger than design basis earthquakes for nuclear stations and other facilities and in shake table tests. These data provide insights to actual seismic 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 l industrial facilities indicates that piping systems designed to industrial standards are  !

rugged. 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 percentage (much less than 0.01 percent) of the total piping at risk. This leads to j 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.

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

Threaded pipe joints or other regions of redur,od cross section with sharp corners susceptible to fatigue, ratchet cracking, or rupture when subjected to cyclic seismic 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 seismic adequacy of the main steam piping leakage path. This process 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 l 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 assure acceptable piping spans, piping support configurations, design attributes, and the  ;

absence of known seismic vulnerabilities.

The capacity spectra that were used in the establishment of the piping seismic capacity were based on the expeilence 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 l- systems which could be best screened by an in-plant walkdown.

In this report, MNGP piping is directly compared to a sample of earthquake experience data. The earthquake experience data that is directly being used for comparison to the MNGP piping is obtained from these five site-event pairs, allin California p

1 El Centro Steam Plant subject to the Imperial Valley 1979 earthquake (0.42g PGA)

2. Valley Steam Plant subject to the San Fernando 1971 earthquake (0.30g PGA)

3. Moss Landing Power Plant subject to the Loma Prieta 1989 earthquake (0.34g PGA)

4. Ormond Beach Power Plant subject to the Point Mugu 1973 earthquake (0.20g PGA)

L 5. Humboldt Bay Power Plant subject to the Ferndale 1975 earthquake (0.30g PGA) 8 l

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l Summary information concerning these earthquake can be found in Reference [1]. The l above earthquake PGA values are per those documents.

Figure 3-1 shows selected ground acceleration response spectra plotted against the ,

MNGP SSE ground spectrum from three documented earthquakes occurring in {

California. These include the 1971 San Fernando (Valley Steam Plant - USGS Estimate), the 1979 Imperial Valley (El Centro Steam Plant), and the 1989 Lcma Prieta (Moss Landing). The Valley Steam Plant record was obtained from Reference 8 and the j remaining records are from Reference 1. All of these earthquakes produced g ound j motions wellin excess of the MNGP SSE ground spectrum. l Horizontal Ground Response spectra at 5% Damping 1.6 l

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f I . Valley Steam USGS I . _ . Moss Landing j l .;. MNGS SSE 0.6 lI r.- s, i

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Figure 3-1: Selected Spectra from References [1], [8] vs. MNGP SSE Ground Spectrum (Note: Response Spectra for Ormond Beach and Humboldt Bay not publicly available.) l l

Figure 4.1 of Appendix D of Reference [1] presents ground spectra at several of the survey sites and also shows the MNGP Desig.1 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 Capaclty An evaluation of the seismic ruggedness of condensers and condenser anchorage for GE BWR riants is reported in Reference [1]. The configurations of the GE BWR

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condensers were compared to condensers in the earthquake experience data.

, Condensers in the earthquake experience data exhibited substantial seismic ruggedness l 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 the Monticello condenser configuration to the earthquake experience data. In particular, l

detailed comparisons to the Moss Landing and Ormond Beach condensers were l performed. The comparison included a detailed evaluation of the Monticello condenser anchorage capacity.

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 classes of equipment against the earthquake experience data. The GIP has been 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 GIP Reference Spectrum, the GIP Bounding Spectrum, and the MNGP SSE Ground Spectrum. Figure 3-2 shows that the MNGP SSE spectrum is well l bounded by the GIP Spectrum.

HorizontalResponse Spectra atS% Damping 14 a ,..........,

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4.0 Seismic Evaluation Methodology 4.1 Piping and Supports 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 supports, and items requiring limited analytical reviews.

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

Limited analytical reviews ard pipe aupport evaluations for piping systems identified during the walkdowns.

Generation of Piping Systern Seismic Screening Work Sheet (PSSSWS), a formal 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.

4.1.1 Comparison to the Experience Data A comparison of plant piping to the experience data was done to verify that system walkdowns, limited analytical reviews, worst case support evaluations and final seismic capacity evaluations were within the parameters shown by the experience data for the earthquakes listed in Section 3.1 above.

Piping Considerations The leakage path piping was compared to the piping in the experience data to assure 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 based on a review of the data in reference [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 831.1 Deadweight Spans.

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

A comparison to the experience data was performed for the Monticello leakage path piping and is documented in Section 5 heroin. 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.

E'quipment Considerations.

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

(a) Anchorage 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 concerns. The SRT members were qualified in accordance with applicable industry criteria.

4.1.2 Limited Analytical Review of Piping and Supports This section defines the capacity criteria that was used in the limited analytical reviews and analysis of piping systems and in the evaluation of worse case supports. The capacity criteria was 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 which was defined in the applicable analytical review package.

Piping The majority of piping systems under review were originally designed to the 1967 B31.1, Power Piping Code. The original design only considered loadings due to pressure, dead load, design mechanical loads, and thermal loads. The original design capacity criteria for the piping can be summarized as follows.

(a) The sum of the longitudinal stresses due to pressure, weight, and other sustained loads shall not exceed the allowable stress in the hot condition, Sn.

[B31.1-1967 Section 103.3.2(d)] .

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2 2 (b) The longitudinal pressure stress, S e,t shall be calculated as Pd /(Do -d 2),

where:

P = Design Pressure (psig)

Do = Outside Diameter (in) di = Inside Diameter (in) l l

l l [B31.1-1967 Section 103.3.2(d))

(c) The thermal expansion stress, Se, ( calculated per Section 119.6.4), shall be  ;

less than SA, where SA s Calculated per section 102.3.2(c). Further when the j sum of the longitudinal stresses, as calculated in (a) above, is less than Sn, the difference between Sn and this sum may be added to the SA Value for

determining tne allowable value of Se (B31.1-1967 Section 102.3.2(c)].

1 (d) The sum of the longitudinal stresses produced by intemal pressure, live and dead loads and those produced by occasional loads such as the temporary supporting of extra weight may exceed the allowable stress values given in the allowable stress tables the amounts and duration of time given in Paragraph 102.2.4 [B31.1-1967, Section 102.3.3(a)]. This permits a 20% increase in the Sn value.

Since the consideration of a Design Basis SSE event was not in the original design basis for the piping systems under review, the capacity criteria given below was established for use in piping system limited analytical reviews and detailed analyses.

P+ .75i[(MA/Z)] < 1.0 S (4.1a) i'[Mc/Z) < SA+ {S-(P+.75iiMA/Z])} (4.1b)

P+ .75i[(MA/Z)+(MBI /Z)] < 2.4 S (4.2) i*[ MC/Z + MBsam /Z] < 2 SA (4.3)

P = Pressure Loadings MA = Applied Moments Due to Deadweight Loadings MBl = Applied Moments due to SSE seismic Inertial Loadings MBsam= Range of Applied SSE Moments due to Seismic Anchor Motion 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 SA = Allowable Expansion Stress range per B31.1 Code i = Stress intensification factor as defined in the B31.1 Code 13

l Equations 4.1a and 4.1b are the standard deadweight and thermal stress evaluation equations per the B31.1 Power Piping Code. It is consistent with the original design basis for applied deadweight and thermal loadings.

The basis for the establishment of equation 4.2 is as follows:

(a) It is slightly more conservative than the capacity criteria used by the ASME Boiler and Pressure Vessel Code, Section Ill, Division 1, for Class 3 piping systems subjected to Level D or faulted loading conditions (1992 and earlier Editions). The SSE event is classified as a Level D or Faulted condition for tb Monticello Plant. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1, Class 3, capacity criteria is the criteria specified for Quality Class C piping in Regulatory Guideline 1.26. The systems under review were judged to fall with in the classification criteria of Quality Class C piping systems, as put forth in Section C.2.d of Regulatory Guideline 1.26.

(b) S is the basic allowable material stress per the B31.1 Power piping Code, which is the lesser of 5/8 Sy(2/3Sy in later code editions) or Su/4. The majority l of the piping under review is A-106B Carbon steel pipe, which has S=15000 psi, j Sy=35000 psi and S u=60000 psi. Therefore Equation 4.2 limits the Pressure +

Deadweight + Seismic inertial Stresses to less than 1.03Sy which assures elastic behavior, i.e., no significant yield or inelastic behavior would be permitted to

. occur during or after a Design Basis SSE event.

The basis for the establishment of equation 4.3 is as follows:

(a) The ASME Boiler and Pressure Vessel, Section Ill, Division 1, Class 3 capacity criteria requires that the sum of the longitudinal bending stresses due to thermal expansion, thermal anchor motions, and OBE seismic anchor motions be i

limited to SA(as defined above). The ASME Boiler and Pressure Vessel, Section Ill, Division 1, Class 3 capacity criteria for Level C (Emergency) and Level D (Faulted) Conditions explicitly excludes consideration of stresses resulting from  ;

l- SSE seismic anchor motions. This basis for this exclusion is, for Level C and D j l loadings, the Code only requires the consideration of primary stresses and does '

not require explicit consideration of secondary stresses. The code classifies I stress from seismic anchor motions as secondary stresses (displacement limited

(. stresses). For this program it was decided to explicitly consider SSE seismic

anchor motions in conjunction with thermal anchor motions and thermal expansion stresses. This was done (1) to assure that the secondary stresses which could occur during a design basis SSE event would be bounded, and (2) from the experience data, piping failures due large seismic anchor motions is a known failure mode. The capacity criteria selected for this review was 2 SA which is twice the capacity criteria used for the Level B load case of OBE seismic anchor motions, thermal expansion stresses and thermal anchor motions. This l use of a Level D limit which is twice a Level B limit is consistent with the philosophy of the ASME Boiler and Pressure Code, Section 111, Division 1.

I 14

f carbon steel pipe is approximately 1.5 S which is approximately (b) S A or 22,500 psi and therefore 2.0 SA s i approximately 1.2 S . These stresses are y

secondary in nature and therefore, if they are limited to less than 2.0 S y , this will assure that elastic shakedown will occur, no significant membrane stress rupture will occur, and accumulated cyclic damage will be elastic. The 1.2 Sy limit used here is significantly less than the upper bound 2.0 Sy elastic shake down limit for secondary stresses.

Piping Supports The piping support acceptance criteria used in the worst case support evaluation are 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)

(b) Component Supports DWT+TH s 1.0 MSS-SP-58 Allowable (4.6)

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

This capacity criteria will assure that the maximum stresses in the support members are at or slightly less than the material yield stress. In many MNGP calculations a factor of 1.6 was used in lieu of 1.7. This simply adds additional conservatism to the calculations and support evaluations. The 1.7 is based on the Part ll allowables of the AISC Steel Construction Manual.

Concrete Anchorage Based on investigations conducted by NSP during the USl A-46 resolution program and field reviews during this program, it was concluded that the concrete anchor bolts used for pipe supports were "Philips - Redheads" of the self drilling type. The anchor bolt capacities of Appendix C of the SQUG-GIP were used as they provide well established capacities for anchorage consistent with the use of experience data. In addition, since the SQUG-GIP was used to evaluate the equipment, the use of the SQUG-GIP anchorage capacities provided a consistent factor of safety between the piping support anchorage and the equipment anchorage. For pipe supports, all concrete anchorage tensile and shear load interactions were evaluated on a linear basis.

4.2 Condenser The seismic adequacy of the Monticello condenser was verified using experience data contained in reference 1 with specific comparisons to the Moss Landing and Ormond Beach condensers. Per reference 1, these condensers are of similar configuration to Monticello and experienced strong motion in excess of the Monticello design basis earthquake without failure. In addition, the adequacy of the Monticello-specific 15

L l

condenser configuration was verified by an evaluation of the Monticello condenser anchorage capacity.

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

The majority of the related equipment are valves located at the lower elevations. Valve operability is not a concern 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

Monticello SSE. The Recombiner building was designed and built for seismic Class I conditions; however, the design criteria for this building was later downgraded to Class ll 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 MNGP USAR [6].

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

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 instrument lines which are located at the operating i floor of the Turbine building (elevation 951'). The Turbine building above elevation 951' is a steel superstructure and is classified as Class 11; however, the superstructure was also seismically evaluated for the Reactor building SSE equivalent elevation  :

accelerations. See Section 12.2.1.4 of the USAR (6].

4.5 Seismic Demand i 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 0.129 . The vertical demand was taken as j i

1 16

l 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 j 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 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 l SSE ground spectrum with significant margin. The demand spectrum for piping in the i Reactor Building was the 5% damped amplified floor response at the applicable elevation.

Limited Analytical Reviews of Piping i

I 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 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 for 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 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.

(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 maximum spectral acceleration of the ground response spectrum in the horizontal direction and 1.5 times two-thirds of the maximum spectral acceleration of the ground response spectrum in the vertical direction.

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 horizontal demand.

17

l l Limited Analytical Review of Buried Piping System l

For the evaluation of buried piping systems, the seismic demand is the design basis SSE ground response spectrum.

! Worst Case Support Reviews .

Seismic loads for use in worst case support reviews are determined as follows.

l (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.

l (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 I determined.

(c) For determination of horizontal loads the value determined in (b) was multiplied by the maximum spectral acceleration 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 i building, the applicable amplified floor response spectra was used.

l 4.5.2 Condenser Demand Spectra l

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.

li 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 j with the Monticello USAR [6], the Reactor Bui! ding 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 USl A-46 resolution and were judged to be " conservative design" spectra when used with the GIP [4]. In addition and consistent with the GlP methodology,1.5 times the ground spectrum was optionally used as " realistic, median centered" demand for some equipment items meeting the GIP 40-foot-above-grade l 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.

l

'he Turbine Building below the 951' elevation is a stiff reinforced concrete shear wall structure typical of nuclear plant structures. All related equipment in the Turbine Building is located at or below the 951' elevation. The Recombiner Building is a stiff reinforced concrete shear wall building also typical of nuclear plant structures.

i l

18 l R______--___ _ _ _ . _ . _ _ . -

l 5.0 Summary of Seismic Evalu.. tion Results 5.1 Piping and Supports l

5.1.1 Results Sumrnary --

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.

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 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 lateral load capacity of the U-bolts. See section 5.1.4 for a summary of these qualifications.

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. The selection of this system was based on the following considerations.

(a) It is a large, complex piping system containing multiple pipe sizes (12",10",

8", 6", 5", 3", 2",1-1/2",1").

(b) It is one of the most flexible systems reviewed and therefore has the potential for significant seismic response.

(c) It contains a significant number of small bore branch lines, and the displacement of the mainline could result in significant seismic anchor motion stresses in the branch lines. Accurate run pipe displacements were required to review and evaluate the seismic anchor motion effects on the branch lines.

(d) The potential for several spatialinteractions between this system and other systems, structures or components exists. Accurate displacement data was necessary to review these potential spatial interaction issues (e) It is located at an upper elevation, and it is expected to experience a greater level ofinput seismic excitation.

19

l For this system a detailed spectral modal analysis using the criteria of ASME. BPVC, Appendix N was conducted. In addition, limited analytical reviews were conducted for

portions of other piping systems which could be considered outside the screening criteria, which involved complex spatial interactions, or for which a highly accurate i prediction of piping support loads was required. One worst-case analytical review was conducted for all buried 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 assure that the Monticello piping systems fall within the range of the piping systems  ;

which constitute the experience data.

L 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 i 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 Monticello piping to the experience data piping. The Monticello piping systems in the leakage path are enveloped by the experience data with the following

! 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.

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 concern. 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 piping is adequately enveloped by the experience data and the supporting analysis.

20

Materials Table 5-4(a) provides a summary of the allowable stress capacity of the 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 represer ted in the experience data piping.

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 actual lateral support spans to the suggested ANSI B31.1 spans. Table 5-6 provides the suggested 831.1 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 response spectra modal analyses were conducted for several piping systems. The criteria used in these piping evaluations and qualifications are given in section 4.1.2. Table 5-9 provides a summary of these analyses and the associated bases.

In addition to detailed dynamic piping ar:alyses 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 discussed in Section 5.1.1. Table 5-12 provides a summary of maximum stresses determined from this detailed analysis.

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 Detaileo SU.nport 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.

21

l I

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

i Table 5-11 provides a summary of the number of supports subjected to detailed j analytical reviews and the basis of these reviews. These supports represent l approximately 15% of the support population in the MSIV leak path. In addition, these j 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 j supports for the MSIV leak path piping has adequate seismic capacity.

l 1

5.1.5 Results l

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

5.2 Condensar Table 5-7 tists design data for the Monticello condenser and for the two experience data sites lis'.ed in Reference [1], Appendix D, Table 4-3 (Moss Landing 6 & 7, and Ormond Bea:,n_1 & 2). The Monticello condenser design data is similar to or bounded by data for the two experience data sites. The Monticello SSE ground spectrum, which is the demand spectrum for the condenser, is enveloped by the Moss Landing spectrum per Figure 3-1. The Ormond Beach estimated PGA demand due to the February 21,1973 Point Mugu earthquake was 0.20g. This exceeds the Monticello PGA of 0.12g. The Monticello condenser design data is also well represented by the data presented in Reference [1], Appendix D, Table 4-3. The comparison verifies that the results of the Reference [1] evaluation for structural integrity are applicable to the Monticello condenser.

The Monticello condenser anchorage consists of eight guided supports with one support located at each corner of the two condenser shells. At each support, the condenser base bears against a steel plate shear lug that is welded to an embedded sole plate.

The shear lugs rigidly resist lateral loads but are arranged 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 l base are 2.75 inches in diameter to allow for thermal growth. Figures 5-5 and 5-6 show guided support layout and details.

22

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 l ductile behavior. Capacities were derived from equations for capacities of anchorage elements defined in codes such as AISC Manual of Steel Construction, ACI-349.

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

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

shear lug load path at a support. Based on a detailed evaluation, the rigid-behavior capacity of the condenser anchorage was determined to be 0.1Sg. The capacity is ,

controlled by the direction transverse to the turbine axis. The shear lug load path  !

capacity parallel to the turbine axis is 0.169. A ductile-behavior anchorage capacity was i 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.

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 structures with substantial internal 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 I 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 i are above corresponding values for the experience data sites shown in Figures 4410 and ,

4-11 of Appendix D of Ref.1 i 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 i- and contrary to the experience data.

l The condenser was also subject to a walkdown inspection which was summarized in a i Screening Evaluauon Work Sheets (SEWS). Some surface cracking of embedment

, grout was observed at support locations. The condenser was declared an outlier L

pending repair of the grout. This grout was repaired during the recent refueling outage at 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 spatial interaction checklist, notes, and attached pictures (if available). The SEWS identify the determination of whether the item is acceptable or 23

1 1

l 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.

l I

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 fait safe as in the case of air- and solenoid-operated valves.

i f

( 24 l

(_

t.

I 1

l Table'5-1 Summary of Piping Properties for the Monticello Leakage Path Piping Walkdown Pipe Size Pipe Pipe Pipe O D/t 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.S 0.3 12 A106B  ;

1 160 1.315 0.25 5 A106B i 2913-2 10 80 10.75 0.593 18 A106B I 2 160 2.375 0.344 7 A106B I 1-1/2 160 1.9 0.281 7 A106B f l 2913-3 18 80 18 0.938 19 A672, Gr. 70  !

l 10 80 10.75 0.593 18 A672. Gr. 70 l 2913-4 4 80 4.5 0.337 13 A106B 2913-5 16 STD 16 0.375 43 A538/A106B

12 STD 12.75 0.375 34 A53B/A1068 l

10 STD 10.75 0.365 29 A53B/A1068 '

2913-6 16 STD 16 0.375 43 A53B/A106B 12 STD 12.75 0.375 34 A53B/A106B

10 STD 10.75 0.365 29 A53B/A106B j l 2913-7 6 40 6.625 0.28 24 A53B/A1068  !

3/4 80 1.05 0.154 7 A53B/A106B l 2913-8 6 40 6.625 0.28 24 A53B/A1068 3/4 80 1.05 0.154 7 A53B/A106B l

2913-9 3 160 3.5 0.438 8 A106B 3 STD 3.5 0.216 l

16 A53B/A106B j 2 160 2.375 0.344 7 A1068 1 80S 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/A106B i 12 STD 12.75 0.375 34 A53B/A1068 10 40 10.75 0.365 29 A53B/A106B 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 l

6 80 6.625 0.432 15 A106B j 4 80 4.5 0.337 13 A106B  !

2913-13 24 80 24 1.22 20 SA106B l 6 120 6.625 0.562 12 SA106B l 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 I 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 A103B 2913-16 4 120 4.5 0.438 10 A106B i 3 160 3.5 0.438 8 A1068  !

25 l

L__ -__ _. l

I Table 5-1 Summary of Piping Properties for the Monticello Leakage Path Piping Walkdown Pipe Size Pipe Pipe Pipe OD/t Material Package NPS (in) Schedule OD (in) Wall (in) ASTM /ASME Designation 2913-17 1 160 1.315 0.25 5 A1068 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 XXH 1.315 0.358 4 SA106B 1/2 XXH 0.84 0.294 3 SA106B 2913-21 4 120 4.5 0.438 10 SA1068 2913-22 1 80 1.315 0.179 7 A1C6B 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 A1068 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 8.625 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) 4 40 4.5 0.237 19 CS(1) 3-1/2 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) 2913-27-4 1-1/2 40 1.9 .145 13 CS(1) 3/4 40 1.050 .113 9 CS(1) 2913-28 1-1/2 160 1.9 0.281 7 A106B 1 160 1.315 0.25 5 A106B 2913-29 14 STD 14 0.375 37 A53B/A106B 10 40 10.75 0.365 29 A53B/A1068 26

Table 5-1 Summary of Piping Properties for the Monticello 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/A1068 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 6.625 .28 24 A106B 2 80 2.375 .218 11 A106B 2913-34 1-1/4 40 1.660 .140 12 A1068 5 40 5.563 .258 21.5 A106B 5 40 5.563 .258 21.5 Cast iron (3)

_2G13-35 2 80 2.375 .218 11 A1068 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 3/4 80 1.050 .154 7.0 A1068 1/2 - Tubing .065" Wall .5 .065 7.7 SS(2) 5/8 - Tubing .065" Wall .625 .065 9.5 SS(2) 5/8 - Tubing .065" Wall .625 .065 9.5 A213-304L 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 A1068 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 1

27

Table 5-2 Seismic Experience Piping Data [1]

Pipe Size Pipe Pipe Pipe O D/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 0.375 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 0.28 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 1 1/2 160 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 16 40 16.00 0.5 32 16 N/A 16.00 0.902 18 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 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 21 N/A 21.00 3.793 6 1

28 i

4

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 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 4 10 40 10.75 0.3650 29 l 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 0.5620 12 6 40 6.625 0.2800 24 4 80 4.500 0.3370 13 4 40 4.500 0.2370 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 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 1 1/2 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 j 3/4 40 1.050 0.1130 9 l

l 29 l

e

Table 5-3 Dit Range Comparison Nominal Pipe Size Monticello Experience Data (NPS)(ID) Piping D/t Ranges Piping D/t 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 31/2 13 4 10-19 8-19 5 15-22 6 9-24 9-24

, 8 27 10-31 l 10 18 10-29 12 31-34 10-34 l 14 37 32-37 16 32-43 11-43 I

18 19-48 5-41 l 24 20 23-35 Table 5-4(a) Predominant Materials of the Experience Data Material ANSI 831.1 Allowable Stress, psi ASTM Designation A53B 15000 A106B 15000 A335 14000 A120 (1)

A139 12000

! (1) Stress allowables not provided by B31.1. B31.9 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 A53 B 15000 A106 B 15000 312-304 15900 376-316 17000 312-304L 13700 B42 - Copper 6000 W (1) This is the lowest value for B42 Copper given in the B31.1 Code.

30 l

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

l Walkdown Pipe Type Maximum Minimum Maximum Minimum l 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 Suggested B31.1 Pipe Size] Suggested Suggested Support Suggested Support Support Spacing Support Spacing (2) Spacing (LVSSR-Max) Spacing (2) (LVSSR - 'l l Min) 2913-1 LB 1.5 1 4.2 1 2913-2 SB 1.5 .5 3 .5 i 2913-3 LB 1 1 3 1

l. 2913-4 LB 2.2 (1) 1.5 7 1 l 2913-5 LB 1.5 1 3 2 l 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 l 2913-10 LB 1 N/A 1.5 1 2913-11 LB i 1.25 5.25 2 2913-12 LB 1.5 <1 2.75 1 i 2913-13 LB (3) (3) (3) (3) 2913-14 LB (3) (3) (3) (3) 2913-15 SB 1 <1 1.6 <1 2913-16 LB 2 1 2.5 1 2913-17 LB 1.5 <1 6 <1 l 2913-18 SB 1.5 1.3 - 5.5 1.3 2913-19 LB (3) (3) (3) (3)

I 2913-20 SB 1.5 <1 2 1 l 2913-21 LB 1.5 1 1.5 1 2913-22 SB 1 .5 1.5 1 2913-23 LB 1.5 1 6 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) l 2,-3 2913-27-4 SB 1.5 1 5 2 l

2913-28 SB 1 1 3 1 2913-29 LB 2 <1 2.7 2.7 2913-30 SB 1 <1 2 1 2513-31 SB 1.5 1 5.0 2 2913-32 LB 1 1 3 2 2913-33 LB,SB 1 1 2 1 2913-34 LB,SB 1 1 2 1 31

\

j Table 5-5 MNGP Span Ratios in Comparison to ANSI B31.1 Suggested Deadweight Spacing Walkdown Pipe Type Maximum Minimum Maximum Minimum Package. SB = Small Vertical Vertical Lateral Lateral l Bore (<2.5") Support Support Support Support j LB= Large Actual Actual Actual Actual Bore (>2.5") Spacing Spacing Spacing Ratio Spacing 1

[ Based on Ratio to Ratio to to B31.1 Ratio to '

Predominant B31.1 B31.1 Suggested B31.1 Pipe Size] Suggested Suggested Support Suggested l Support Support Spacing Support 1

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 consideration of spring hangers.

(2) Spans inc!ade consideration of modified or added supports. l I '

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

(4) This was a worse case system and was qualified by detailed analysis. l l

l 32

Table 5-6 Nominal Suggested Vertical Deadweight Spans per ANSI B31.1 Suggested B31.1 Deadweight Spans (ft)

Monticello Outside Pipe Water Service Steam. Gas or

Nominal Pipe Diameter (in) Air Size ** (in) Service 3/4 1.050 6* 8*

1 1.315 7 9 1 1/2 1.900 9* 11*

2 2.375 10 13 3 3.500 12 15 31/2 4.000 11* 12*

4 4.500 14 17 5 5.563 16 19*

6 6.625 17 21 8 8.625 19 24 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 small amounts of 1/2" piping and liC tubing (1/8",1/4",1/2", 5/8" and 3/4") not presented in this table.

33

\

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

Parameter Monticello Moss Landing Ormond Beach j 6&7 1&2 l Manufacturer Worthington Ingersoll Rand Southwestern l 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 MP: 210,000 ft' 435,000 ft' 210,000 ft' i LP: 189,000 ft 2 l l Shell Material ASTM A285C ASTM A285C ASTM A285C I I

Shell Thickness  % inch  % inch  % inch Operating Weight HP: 1,900,000 3,115,000 lbs. 1,767,000 lbs.

Ibs.

LP: 1,800,000 l Ibs. l Tube Material Type 304 S.S. Al-brass 90-10 Cu-Ni Tube Size 1 inch 1 inch 1 inch i Tube Length 36 to 40 feet 65 feet 53 feet l Tube Wall Thickness 18 to 22 Bwg 18 Bwg 20 Bwg 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 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

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

l 1

l Table 5-8 Summary of Concerns and Resolution l

Identifier Concerns Resolution Package 2913-4 Spatialinteraction Loose equipment moved or restrained Package 2913-5 Loose hanger Hanger repaired l Package 2913-4 (a) Broken U-Bolt (b) Missing U- (a) Replaced (b) Installed (c) l Bolts (c) Spatialinteraction Potential target conduits determined j to be not required for normal or accident conditions Package 2913-11 (a) Lack of Late ral Restraint (b) (a) Support modified (b) Repaired i

Loose rod har' Jer (c) Short rod (c) System qualified assuming this rod l hanger (d)Pe urly supported l&C line hanger failed (d) Reroute /resupport line l Package 2913-12 (a) Loos' J-Bolt (b) Loose rod (a) Repaired (b) Repaired (c) U-Bolt hanger added l (c) Additionallateral support required 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 I ' Trays (c) Tubing needs added l lateral / vertical restraint (2 places) (d) Lead blecks restrained l (d) Spatialinteraction for SV-2 and 17-104 Package 2913-20 (a) Missing U-Bolt (b) Spatial (a) U-Bolt installed (b) Block wall i interaction braced l Package 2913-22 Spatialinteraction Crane rail demonstrated to be seismically adequate Package 2913-24 (a) Lateral support required (b) (a) Support added (b) Piping l Short rod hanger qualified assuming hanger would fail.

l Package 2913-26 (a) Lack of seismic support (b) (a) Line 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 l Package 2913- Lack of lateral support & spatial Seven new pipe supports added l 27-2 interaction concerns Package 2913- Lack oflateral support Three supports added 27-4 Package 2913-28 Missing support Support reinstalled l E-2A, E-2B, E-4 Anchorage Bracing was added to reduce anchor l l loads l T-33 Anchorage Bracing and anchors were added V-813 Anchorage Plates added 35

Table 5-8 Sursmary of Concerns and Resolution identifier 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 l

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.

2913-27-4 Piping Overspans Added three supports 2913-40 Inadequately supported Re-support the tubing system 36

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. Alllarge bore piping (>2 in diameter) including all possible leak paths to the condenser.

Displacements at all small bore (2" and under) connections to the large 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 Concerns 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 assumed 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 of the piping system 2917-30 Oxygen injection Piping Although the line was well supported, the ASTM B42 materialis not represented in the experience database of references [1]

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

37 L____ _ ____ __

Table 5-10 Summary of Equivalent Static Analyses Conducted i i

Based for the Equivalent Static Number of Equivalent Static Analyses Analyses conducted for this Reason Evaluate SAMs 2 l Evaluated Spatial Interactions 8 Evaluate Local Vulnerabilities 2 Determined Support loads for Evaluation 5 Table 5-11 Summary of the Detailed Support Qualifications Basis of the Qualification Number of Supports Evaluated Rod Fatigue Concerns 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 Table 5-12 Maximum Stress Levels for the Steam Seal System Location Equation 4.1b Equation 4.2 Analysis Pipe Developed Allowable Developed Allowable Node Point Size (in) Stress (ksi) Stress (ksi) Stress (ksi) Stress (ksi) 105 8 7.1 36.4 15.8 36.0 42 12 14.3 36.4 15.9 36.0 99 10 4.1 36.4 15.3 36.0 545 3 22.1 36.4 16.3 36.0 575 8 27.1 36.4 9.2 36.0 685 10 27.6 36.4 4.2 36.0 Notes:

Maximum Equation 4.1a stress is 4.9 ksi at node point 545. Allowable stress is 15.0 ksi.

Maximum Equation 4.1b stress is 27.6 ksi at node point 685. Allowable stress is 36.4 ksi.

Maximum Equation 4.2 stress is 16.3 ksi at node point 545. Allowable stress is 36.0 ksi.

Equation 4.3 stress evaluation not required as there are no seismic anchor motion effects on the main line.

( 38 L-__---------

Large Bore Vertical Support Span Ratio of Monticello j Subject Piping Compared to the Experience Data '

i e l 1

E i O Experience Data g g Monticello Data

.a 5

z

% %_ I _s .

= m 9 9 3 3 7 a

VSR Figure 5-1 l

Small Bore Vertical Support Span Ratio of the Subject l Monticello Piping Compared to the Experience Data i

I h O Experience Data j o y Monticello Data l 5 -

I E i m e o e m N- $ $ I a a VSR Figure 5-2 39

r -

Large Bore Piping Comparision of the LVSSR of the Subject Monticello Piping to the Experience l Data j l

7_ _ _ _ . _ _

• l E l 3e i 1

$ O Experience [hta o EManticello Data I u j g -

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^

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n y 6

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Figure 5-3 l 1 Small Bore Piping Comparision of the LVVSR of the Subject MonticelloPiping to the Experience '

l Data e

5 u>i M O Experience Data 8 -

E Monticello Data 2 -

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z

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Figure 5 6: MNGS Condenser support details from Worthington DR-127368 Rev. B, Support B is similar to Support A 42-L

6.0 References

[1] NEDC-31858P, Revision 2, General Electric, "BWROG Report for increasing MSIV Leakage Ratic Limits and Elimination of Leakage Control Systems," September 1993, (principally Appendix D thereof).

[2] EPRl/SQUG, " Generic Implementation Procedure (GIP) for Seismic Verification of

' Nuclear Plant Equipment," Revision 2, February 1992.

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

[4] "Monticello Nuclear Generating F'lant 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

l l

43

Attachment 3 Seismic Calculation Package i

J This package includes the following calculations.

96C2913-C-020 Limited Analysis for Baseplate Qualification 96C2913-C-016 HP and LP Turbine Condenser 96C2913-C-015 Gang Hanger Support

.m._dE._:_. .m _ _ _ _t%...___ ._._;