ML20199D575
| ML20199D575 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/30/1997 |
| From: | Stevenson J External (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199D486 | List: |
| References | |
| 97.174-1, 97.174-1-R, 97.174-1-R00, NUDOCS 9711200364 | |
| Download: ML20199D575 (46) | |
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I Rpt. No. 97.174-1 Rev. 010/30/97 ;
OVERVIEW OF SAFETY RELATED LARGE BORE PIPING AND PIPING SUPPORT DESIGN AND -
CONSTRUCTION CURRENTLY EXISTING AT CRYSTAL RIVER-3 NUCLEAR POWER PLANT PREPARED FOR: ,
NUCLEAR OPERATIONS ENGINEERING FLORIDA POWER CORP.
15760 W. POWERLINE ST.
CRYSTAL RIVER, FL 34428-6708
'I 30 OCTOBER 1997 PREPARED BY:
J.D. STEVENSON, CONSULTING ENGINEER 9217 MIDWEST AVE.
CLEVELAND, OH 44125 jim *$$$Nh3a j
TAHLE OF CONTENTS OVERVIEW OF SAFETY RELATED LARGE BORE PIPING AND PIPING SUPPORT DESIGN AND CONSTRUCTION CURRENTLY EXISTING AT CRYSTAL RIVER 3 NUCLEAR POWER PLANT
1.0 INTRODUCTION
2.0 BACKGROUND
3.0 CURRENT PIPING STATUS 3.1 ISSUES 3.2 PLANT CONDITION 4.0 SUGGESTED PATli FORWARD 4.1 USE OF EXISTING CRITERIA 4.2 USE OF TlhiE lilSTORY ANALYSIS 4.3 USE OF R.G.l.61, R.G.I.122 AND SRP 3.7 CRITERIA 4.4 PERFORM A PILOT STUDY 4.5 CRITERIA CONCLUSIONS 5.0 RE3OLUTION OF CURRENT PIPING ISSUE 5.1 TECllNICAL ISSUES 5.2 DOCUhiENTATION ISSUES 6.0 SAFETY SIGNIFICANCE OF Tile CURRENT DESIGN AND INSTALLATION OF LARGE BORE SAFETY RELATED PIPING AT CR 3 6.1 filSTORICAL STUDIES TO DETERh11NE SAFETY SIGNIFICANCE AND DESIGN ADEQUACY OF NUCLEAR POWTR PLANT SAFETY RELATED PIPING 6.2 COhfPARISON OF Tile SAFETY SIGNIFICANCE OF INSTALLED SAFETY RELATED PIPING IN OLDER NUCLEAR POWER PLANTS AND TilAT FOUND INSTALLED IN CR-3 6.3 COhiPARISON OF Tile SAFETY MARGINS REQUIRED IN CURRENT DESIGN OR SAFETY RELATED PIPING IN NUCLEAR POWER PLANTS AND TilAT USED AT CR 3 7.0 SUhthiARY AND CONCLUSION
97cl74g overview.rpt Rev.0103097 OVERVIEW OF SAFETY RELATED LARGE BORE PIPING AND PIPING SUPPORT DESIGN AND CONSTRUCTION CURRENTLY EXISTING AT CRYSTAL RIVER 3 NUCLEAR POWER PLANT i BY: J. D. STEVENSON' la. 1NIAODUCTION The construction of nuclear power plant safety related piping to include administrative requirements, material selection and qualifications , design, fabrication, installation, examination
.and testing is currently the most time consuming and technically challenging single activity at a nuclear power plant. Nuclear plants of Crystal River. 3's vintage age typically have approximately 40. 000 feet o' safety related large bore *, This compares with only about six safety related building structures which require dynamic modeling and analysis which are less than 800 feet in total eight. For this reason, it is absolutely essential that simplification, efficient methods were used to design safety related piping.
This paper meant to place in proper perspective those issues currently associated with nuclear safety elated piping at Crystal River 3 consistent with adequate safety and the state of the art of process piping construction of nuclear power plants in the time frame Crystal River 3 was constructed.
In preparation of this paper, the fo!)owing reference material has been reviewed:
(1) Adams, T.M. and Stevenson, J.D., " Differential Design and Construction Cost of Nuclear Power Plant Piping Systems as a Function of Seismic Intensity and Time Period of Construction", Draft WRC Bulletin (Approved for publication)
(2) Stevenson, J.D., " Survey of Strong Motion Earthquake Effects on Thermal Power Plants in California with Emphasis on Piping Systems", Volumes 1 and 2 NUREG/CR-6239, U.S. Nuclear Regulatory Commission, November 1995 (3) Wais, E.A.," Evaluation of Piping and Support Documentation for Crystal River 3", Report No. 96 0-002 Wais and Associates, Inc., October 1996.
' Consulting Engineer,9217 Midwest Avenue, Cleveland, Ohio 44125 1
- . . , , . . - - - , . . . . y.
(4) Bitner, J. et. al., " Technical Position on Industry Practice", Welding Research ,
Council. WRC But 300, December 1984. !
(5) Branch, E.B. et. al., " Technical Position on Piping installation Tolerance", Welding Research Council. WRC Bul 316, July 1986.
(6) Rehn, D.L. et. al., " Position Paper on Nuclear Plant Pipe Supports", Welding Research Council. WRC Bul 353, May 1990. ,
(7) Seismic Task Groups, " Evaluation of Seismic Designs A Review of Seismic Design Requirements for Nuclear Power Plant Piping", NUREG 1061 Vol. 2, Report of the U.S.
Nuclear Regulatory Commission Piping Review Committee, April 1985.
(8) Seismic Task Group and Stevenson and Associates, " Summary and Evaluation of Ilistorical Strong motion Earthquake Seismic Response and Damage to Above Ground Industrial Piping" NUREG 1061 Vol. 2 addendum, report of the U.S. Nuclear Regulatory Commission Piping Review committee, April 1985.
(9) Piping Review Committee, " Summary- Piping Review Committee Conclusions and Recommendations", NUREG 1061, Vol. 5 Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, April 1985.
(10) Section 5.0, " Seismic Qualification Data, Revision 8, Crystal River Plant Unit No.
3, August 1995.
(11) MDG-1, " Piping Analysis Design Guide", Revision 1, Crystal River No. 3. Florida Power Corp., December 1992.
(12) Indian Point #2, Final Safety Analysis Report, Section 5, Dochet No. 50-247.
(13) II.B. Robinson, Final Safety Analysis Report, Section 5, Dochet No. 50 261.
(14) Point Beach, Final Safety Analysis Report, Section 5, Dochet No. 50-266.
(15) R. G.1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants", Revision 1, December 1973.
(16) R. G.1.61, " Damping Values for Seismic Design of Nuclear Power Plants", U.S. ,
Nuclear Regulatory Commission, October 1973.
(17) R.G.1.122 " Development of Floor Design Response Spectra for Seismic Design of Floor- supponed Equipment on Components, Revision 1. U.S. Nuclear Regulatory Commission, Feb.1978.
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(18) Subudhi, M. " Alternate Procedures for the Seismic Analysis of Multiple Supported Piping Systems" NUREG/CR-3811 Aug 1984.
(19) Bezier, P.," Response Margins Investigation of Piping Dynamic Analysis using the Independent Support Motion Method and PVRC Damping", NUREG/CR 5105, March 1988.
(20) Senior Seismic Review and Advisory Panel, "Part 1: Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Pl nts", DE92 019328 Sandia National Laboratory, June 1992. ,
(21) Weber, R.L Collece Physics,4th Edition. McGraw liill, Inc.1965.
(22) Stevenson and Associates,"Walkdown Procedure for Evaluation and Verification (Leak Tight and Structural Integrity) of Seismic Category 11 or Ill Non Safety Related Piping in Nuclear Power Plants", Rev. 3 Prepared for EPRI, January 1996.
(23) N. Newmark et.al. " Seismic Review of Dresden Nuclear Power Station Unit 2 for the Systematic Evaluation Program, NUREG/CR-0891, U.S. Nuclear Regulatory Commission, April 1980.
(24) Chang, T.Y. "Scismic Qualification of Equipment in Operating Nuclear Power Plants Unresolved Safety issue A-46", NUREG 1030, Nuclear Regulatory Commission, February 1987.
(25) Budnitz, R.J. et. al. "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants", NUREG CR 4334, U.S. Nuclear Regulatory Commission, August 1985.
(26) Sobel, P. " Revised Livermore Seismic Ilazard Estimates for 69 Nuclear Power Plant Sites East of the Rocky Mountains", NUREG 1488, Nuclear Regulatory Commission, October 1993 (27) Kennedy R. " Presentation at ASME BPVC Section til Subgroup on Design, Special Working Group on Seismic Rules Meeting, New York. NY, September 1997.
- LQ BACKGRCUND ~
At the time Crystal River-3 was constructed in the late IP60's to early 1970's, piping system design in particular was in a great state of flux. The industry was in transition from piping design which was rationally based largely on engineering judgement and experience gained from the successful operation, since the 1950's, of high energy and temperature piping systems in fossil power plants designed to the requirement of the ASME B 31.1 1955 and USAS B 31.1,0-1967
. construction codes. For high temperature (>200 'F) piping systems, designs were based primarily on assuring highly flexible systems to accommodate thermal movements due to elevated t,mperatures and sufficient deadweight support to restrict sag which effectively limited 3
. ~ _ . - - . _ _ _ _ _ _ _ _ ._ _
deadweight stresses to about 10 percent of normal code allowable (1500 to 2000 psi). Stress ;
analysis and the documentation of stress calculation of the piping was neither performed nor i considered necessary to assure an adequate design based on past experience with similar systems. !
Computerized stress analysis of piping in fossil power plants at the time was typically limited at most to three systems, main steam, reheat and feedwater.
The introduction of the seismic design requirement to nuclear safety related piping in the 1965-1971 time frame was a complication generally handled by adding supplemental lateral supports 4 every 4 to 6 deadweight span spacing depending on the relative seismic intensity of the site nd the location of the piping system in the building structure. The use of these lateral to vertaal support ratios were determined to keep stresses within the USAS B 31,1.0 allowable of 1.2 x 1.5 Sh or 1.8Sh which was the allowable stress typically used by the nuclear power industry in the 1966 to 1971 time frame when considering the hiaximum Hypothetical Earthquake, h1HE-(safe shutdown earthquake, SSE), load combined with deadweight and pressure on a generic basis (i.e., use of spacing tables and charts), in hot piping lateral suppons were often selected as snubbers (i.e., would move to accommodate thermal effects but would be rigid under suddenly applied seismic loads) so as not to interface with thermal movements.
Recent reviews of power plant piping behavior in several fossil power plants which have experienced earthquakes with at least 0.2 g zero period ground motions in Califomia * (twice the hillE at CR 3) have validated that iSh1E B 31.1 piping designed by very simple static procedures (or no seismic design at all) have performed very well in strong motion earthquakes.
There has been less than one pipe failure (loss ofleak tight integrity) per 50,000 feet of pipe at risk out of a total piping population of over 1,000,000 feet which have experienced at least 0.2 g Peak Ground Accelerations seismic input. This was for piping where more than 95 percent was not designed to resist any seismic load.
M CURRENT PIPING STATUS M issues The Wais repon (3) in a recera survey of the Crystal River-3 installed safety related piping, identific<l a number of large bore pipe (Do > 2 % ") design and design document issues which are summarized in Table 2 of this paper, hiost of the issues raised are generic to this vintage of nuclear power plants. Nuclear industry responses to these ssues have been primarily in the' form of WRC Bulletins (4,5,6). The issues in Table 2 that are noted by individual footnotes are addressed by the generic WRC Bulletins with the reference as given. The WRC bulletin discussions have also been supplemented by text in Section 5.0 of this report. In Table 3 additional piping systems issues applicable to CR 3 have been identified.
M Plant Condition 1 It must be emphasized that the current installed condition of the Crystal River plant large bore piping is similar to the condition of approximately 50 nuclear power plants in operation in the 4
U.S. today which obtained their Construction Pennits prior to 1970. A concise summary of the historical development of nuclear safety related piping applicable to the nuclear power plant at Crystal River's vintage can be found in WRC Bulletin 300, Section 4.2 and is provided as ,
Attachment A to this paper.
The state of the art of design of safety related large bore piping construction in nuclear power plants has changed signincantly since the early 1970's. These changes have been carefully reviewed by the NRC '" " with no resulting concem being identined regarding the continuing safety of piping designed and installed in the earlier plants.
In addition a walkdown of the CR 3 safety related piping did not result in the observation of any condition which would suggest that the safety related piping is its current "as installed" condition would not maintain its leak tight and structural integrity if exposed to the CR 3 design based earthquakes. This conclusion is also based on a comparison of the "as installed" condition of piping at CR 3 with the "as installed" condition and behavior of approgimately one million feet of piping in 8 power stations in California when they experienced earthquake input motions ,
which were at least twice the Mile (SSE) level carthquake (0.20g PGA) denned for CR 3.
- 10 SUGGESTED PATH FORWARD Given the number ofissues identined in Tables 2 and 3 of this paper associated with piping system design basis documentation and connguration management it is recommended that consideration be given to reanalysis and documentation of the design adequacy of safety related large bore piping in the Crystal River Nuclear Power Plant Unit 3. There are available at least 3 options associated with this reanalysis. The nrst option is to use the existing design basis criteria denned for CR-3. A second alternative is to use the existing design basis ground response spectra as the basis for generating a new design basis time history using criteria denned in R.G.1.122. The third attemative would be to adopt current Regulatory Guides, 1.60,1.61 and 1.122 and Standard Review Plan Section 3.7 seismic design criteria for the large bore reevaluation program.
LL Use of Existine Criteria In applying these original design bases criteria those issues identined by an *, as shown in Tables 2 and 3, would not be considered in the reanalysis because a) they were not part of the stafe of-the an at the time the original piping analysis was perfonned, and b) subsequent industry evaluation and experience "*" have indicated that these issues are not signincant contributors to the potential for pipe system failure, in my opinion, based on a brief plant walkdown and review of the conservatism in the acceptance enteria contained in Ref,11, use of the existing criteria in the reanalysis of the large bore piping would result in signi6 cant modincation to .
supports of such piping.
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M_ Use of Time Historv Analysis New design basis time histories would be used to general new Goor response spectra using the procedures given in R.G.1,122. The time histories generated would. meet all the current R.G.I.122 and SRP Section 3.7 guidance for the generation of design basis time histories This alternative would require some licensing changes to the existing design basis in that time history analysis, as well as response spectra dynamic analysis of piping would be permitted.
I believe that the new Coor spectra and new equipment spectra, if needed, would be less consc .ative than current door spectra. This atternative use of time history analysis would permit signincant reductions in the conservative inherent in the floor response spectrum analysis method.
In addition, average time history analysis procedure would result in equivalent seismic stresses in piping which are 55 percent of those produced by compatible response spectra "*.
M Use of R.G.I.60. R.G.I.61. R.G.I.1]2 and SRP 3.7 Criteri,a The use of current R.G.I.60,1.61,1.122 and SRP Section 3.7 would require significant licensing interaction with NRC, but should result in the least conservatism of the three alternatives, hence, the least potential ror physical modifications to the edsting piping.
4J Perform a Pilot Study In order to decided which of the 3 alternatives just described should be pursued, it is recommended a pilot study of 3 typical large bore piping systems (one cold and two hot) assumed
- r. 3 different plant locations, be perforn,ed to determine the consequences of the 3 alternatives.
For alternative 1 the following 3 locations and seismic inputs are suFgested:
- a. Ground, CR-3 Ground Response Spectrum - Figure I
- b. Reactor Building Interior, Elev.180.5, Fig 10"* increased by a factor of 2 for MHE (SSE) (4.0g).
- c. Auxiliary Bldg., Elev.143, Fig 12"* increased by a factor of 2 for MHE (SSE)
(l .Sg).
For alternative 2, at locations b. and c., generate new floor spectra using the spectrum to spectrum method based on the shape of the current CR-3 ground spectra for 0.5 percent piping damping and' 2 and 5 percent building structure concrete damping.
Also, develop time histories compatible with the design basis spectra: input and analyze the piping using a time history rather than response spectrum input. For abernative 3 repeat this process for R.G.1.60 and 1.61 compatible spectral shapes and damping values, r
In Figure 1 is shown a comparison of the CR 3 0.5 percent damped ground respoase spectrum compared to R.' G.1.60 shaped spectrum.fer 2.0 percent damped (piping - s12.0 inches in
' diameter), 3.0 percent emped (piping >l2.0 inches in diameter) and 5.0 recent (CCN411) damping. It should be noted from Figure I that the seismic spectral input from R.G,1.60 and 2.0 percent damping gives slight:y higher values (0.42g versus 0.36g) there the current CR 3 ground response spectrum for 0.5 percent damping; However, in the generation of floor spectra by the Biggs method and the lower bu9 ding and pipe damping used at CR-3_ resulted in significantly higher Door response spectra there than would be the case using methods defined in R.G.I.60, 1.61 and 1.122 UD for generation of floor spectra. Also, it should be noted that the allowable piping stre.a identified in the FSAR (1.2 Sh) for the SSE loaa case is nut consistent with allowable stress criteria typically used with the USAS B 31.1.0-1967 piping code 01 for SSE level seismic load case of 1.8 Sh. The large bore piping reanalysis would use th( rm 1.8 Sh acceptance criteria for evaluation purposes associated with the R.G.1.60, R.J. t.61 and R.G.l.122 SSE based spectral input (alt. 3)
In addition consistent with the use of the R.G.'s 1.60,1.61 SSE shaped spectra, the allowable stresses or load rating of MSS-SP-58 type supports and allowable stresses in AISC type support would be increased by the factor 1.6 instead of the current 1.33 increase.
The approach of using R.G.1.60 and 1.61 compatible criteria would also require some other modifications of the current SR 3 analytical approach such as using 3 components of E.Q. and combining them on an SRSS basis rather than Absolute Sum of each of the two camponents.
M Criteria Conclusions In my opinion, if the current CR-3 design criteria is used without employing one of the other two alternatives discucced herein, this will result in a number of instances where existing piping and supports are not in conformance with the current design basis. However, this is not to say that they would fail to meet currently propcsed operability criteria hence safe operation of the plant would not be an issue, it is strongly recomn ended that 3 typical large bore piping systems be analyzed at the 3 plant locations identified herein usir.g the 3 alternatives just identified in order to obtain data on the relative advantages and disadvantages of each alternative.
19 RESOLUTION OF CURRENT PIPING ISSUE L1 Technical Issues 5.1.1 Modeling of Values Model%g and analysis of valves in piping systems is of concern when considering 1) the seismic loadinh ca the valve body and operator and its abihy to operate or maintain its leak tight integrity as a result of the seismic loading, and 2) the loads the valve and operator has on the supporting piping as a result of the seismic excitation.
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5.1.1.1 Valve Qualification Valves are typically identified as inline components (i.e., they are supported by their attached piping). As a practical matter, most valves are located within one or two feet of one or more pipe supports. Hence the valve seismic excitation is usually limited by the motion of the pipe support (s) adjacent to the valve rather than the overall response of the piping system. In addition, ;
any load coming from the valve is transferred primarily to the pipe support rather than the piping due to the close proximity of the support.
Most piping valves including:
- Hand operated
+
Fluid operated (air and hydraulic)
- Motor operated
- Solenoid operated are seismically rugged as determined by a Bounding Response Spectrum. This spectrum as shown in Figure 2 was developed from real strong motion earthquake experience (1) (Peak Ground Acceleration > 0.2g). The Bounding Respone Spectrum is applicable as a limiting input motion to demonstrate that no valve seismic failures in real strong earthquakes have occurred provided certain caveats and geometry limitations are met" The 5.0 percent Bounding Response Spectrum very conservatively envelopes the CR-3 MHE (SSE) Ground Response Spectrum for 0.5 percent damping as shown in Figure 2. To assure no failure certain valve operator weight and geometry limitations are required to be met as shown in Figure 3."*
There are several caveats which also must be met as follows:
The valve body should not be made of cast iron. The intent of this caveat is to avoid the brittle failure mode of cast iron as evidenced by the poor performance of some cast iron components in earthquakes. Cast iron is not common in such applications in nuclear plants; therefore, it is not necessary to determine the material of the valve body unless it is likely that the body is made of cast iron. It is suggested that the material of a flanged valve be checked. In ,
such cases, if the valve is cast iron, the intent of this caveat is satisfied if seismic stresses in the valve body are low (for example, less than 5,000 psi).
The yoke of power operated valves should not be made of cast iron. The intent of this caveat is to avoid the brittle failure mode of cast iron as evidenced by the poor performance of some cast iron components in earthquakes. Cast iron is not common in such applications in nuclear plants. Therefore, it is not necessary to determine the material of the valve yoke unless it appears to the Walkdown Engineer that the yoke is made of cast iron. In such cases, if the yoke 8
I is cast iron, this caveat may be satisfied by performing a stress analysis of the valve for a 3g load
- applied at the center of gravity of the operator in the yoke's weakest direction. If the yoke stress
' is low (for example, less than 5,000 psi), then the intent of the caveat is satisfied.
The valve should be mounted on a pipe line of at least 1-inch diameter. This is the lower
+=
bound pipe size supporting fluid or motor operated valves currently in the earthquake experience
- value equipment class. The concern is that valves with heavy operators typically associated with power operated valves on small linu may cause an over stressed condition in the adjacent pioing.
' To satisfy the intent of this caveat a stress analysis (that accounts for the power operated valve t
operator eccentricity) may be used to'show that the pipe stress' adjacent to the valve is within -
acceptance criteria. There is no concern if the valve, the operator, and the line (if smaller than 1 inch) can be judged to be well supported and anchored to the same support structure. This judgement can often be made by a tug test where the tug load equals or exceeds a 3g loading on the value.
The valve actuator and yoke should not be independently braced to the building or other structure or supported by the structure unless the pipe is also braced to the same structure-immediately adjacent to the valve. The concem is that if the operator is independently supported from the valve and attached piping, then the operator may act as a pipe support during seismic motion and attract considerable load through the yoke and possibly fail the yoke or bind the shaft.
In addition, if both the operator and the valve / pipe are restrained, and if they are both not tied back to the same structure, then differential motion of support points may lead to high seismic loans and possible binding of the shaft. If either of these concerns are noted, then a detailed analysis should be conducted to demonstrate acceptable stresses and deflections.
Sufficient slack and flexibility should be present in attached line (e.g., cooling, air, electrical) to preclude a line breach due to differential seismic displacement of the equipment and the line's nearest support.
Since the Bounding Ground Response Spectrum envelops the CR-3 MHE (SSE) Design Basis Ground Response Spectrum by a very large margin, all valves meeting the caveat, geometry and weight limitations are qualified up to at least 40 ft above grade,159.0 ft. (162.0 ft. in Auxiliary Bldg.) elevation. Above the 159.0 ft elevation it is recommended that valve accelerations from
, a sample of piping analyses be compared to the 1.5 Bounding Spectrum spectral accelerations at -
~
the appropriate frequency of the piping systems. If the computed valve accelerations are less than 1.5 times the Bounding Spectrum acceleration, the valve design adequacy is assured *.
Alternatively, resultant stresses in the valve can be compared to Code allowable to determine design adequacy.
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l
-,- -.. - _. - ,-. . - -,- ..- -~ - - .. .-
5.1.1.2 Piping Qualification Adjacent to the Valves Figure 2 shows the USI.A-46 Bounding Response Spectmm for 5.0 percent damping enveloping the CR-3 MHE design basis Ground Response Spectrum for 0.5 percent damping. If valve geometry and weight limitations in Figure 3 are not exceeded, experience has shown there is no potential for failing the piping supporting the volve (2* .
Above the 159.0 ft elevation, the same spectral comparison used for valve evaluation adequacy is recommended to evaluate stress resultants in piping caused by eccentric valve operators. If the CR-3 floor spectral valves for 0.5 percent damping exceed the 1.5 times the 5.0 percent damped Bounding Spectrum (very conservative) then sample stress analysis of the piping should be performed to demonstrate design adequacy.
Because determination of eccentric valve operator induced stresses in the piping were beyond the state-of-the-ait at the time of the original design basis, it is recommended that when this eccentricity is considered explicitly the allowable primary stress under this loading condition be increased from 1.2 Sh to 1.8 Sh consistent with the criteria typically used in nuclear power plant design using the USAS B-31.1.01967 Code o2.uw ,
5.1.2 Zero Period Accelerations (ZPA) Consideration (Missing Mass Corrections)
The dynamic analysis of piping systems, in order to adequately represent the participating mass of the piping system hence the resultant seismic inertia stresses must consider most of the mass of the piping system participating in the dynamic re:ponse. It is impractical to require essentially all of the mass participating since this typically would require the inclusion of modes of response with frequencies greater than several hundred Hertz. Normally, 90 percent modal mass participation is considered sufficient to defme resultant seismic inertia stresses m. Historically, this effect has been addressed by including all response modes with frequencies equal to or less than 33 Hz.
Modern piping t nalysis computer programs include a missing mass correction and should be used in any reanaly:is of large bore piping.
5.1.3 Welded Attachments Computation of local pipe wall stresses due to lug, trunnion or stanchion type attachments in the time frame of the design and licensing of Crystal River-3 (CR-3) was not normally performed.
The piping Code of Record for CR-3 is USAS B31.1.0-1967. It does not provide specific rules for the evaluation of pipe stresses at welded attachments. In USAS B31.1.0 - 1967. General guidance is provided as follows:
" Consideration shall be 3i ven to the localized stresses induced into the piping component by the integral attachments.
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The design of hanger lungs for the attachment to piping for high temperature service shall be such as to provide for differential expansion between the pipe and the attached lug."
The method of consideration was lef1 to the judgement of the analyst or designer. In most cases, previous successful experience in service with similar attachment details were used as the basis of design. From a Code requirement standpoint, specific rules for evaluation oflocal stresses at piping attachments in ASME Section Ill Class 2/3 for nuclear piping were not provided until 1981 when ASME BPVC Code Case N 318 was published. Even today, specific rules have not been provided in the ASME/ ANSI B31.1 code.
Local pipe wall stresses induced by integral attachments are considered to be local stresses and are only a design concern as potential fatigue crack initiators. ASME B31.1 and ASME B&PV Section III Class 2 and 3 implicitly consider fatigue by reducing allowable piping stresses as a function of the number of cycles above 7000 cycles. The pseud, elastic stress allowable associated with low cycle (below 7000 cycles) fatigue were judged high enough to envelope local stresses induced by standard lug attaclunents, hence, were not explicitly considered in design.
Behavior of ASME B31.1 piping in power plants in California which have been subjected to strong motion ZPGA above 0.2g validate this assumption. No pipe lug or attachment failures due to earthquake effects (on over 1 million feet of pipe inspected) of at least twice the MHE design basis for CR-3 have been recorded (2). This is not to say that lugs don't occasionally fail due to high cycle thermal loads. However, such failures are quire rare even in the vast majority of cases where such localized stresses were not explichy considered in the design.
Typically, during the design of CR-3, enginee:s sized lugs and other integral attachments based on experience and performed no explicit stress analysis. If the area and section modulus of the attaclunent and the weld of the attachment to the pipe are qualified for the attachment load, at the interface between the lug and the piping wall the attachment can be considered acceptable.
Although, local pipe wall stresses were not specifically considered in the design of welded attachments, they are considered acceptable based on the procedures and techniques available during time frame of the design and licensing of CR-3. Any reanalysis oflarge bore piping could use fmite elements to analyze typical lug, trunnion or stanchion arrangements at CR-3 in a separate effects study in order to determine limiting load capacities at these welded attachments.
5.1 A Clearances Clearances including tolerances associated with piping installation are discussed in detail in Section 3.0 of WRC Bulletin 316* and 2.3.3 of WRC Bulletin 353m .
As built piping system clearances need to be evaluated by field walk downs. The required clearance depends on the relative flexibility of the piping system. In general,0.75 inches are sufficient to insure no adverse clearance problems with in line piping system components such 11 l
as valves. For passive piping components, the inertia velocities (V < 10 mph) developed in piping systems by the Maximum Hypothetical Earthquake (SSE) excitation are not sufficient to cause damage to such piping.
5.1.5 -Seismic Anchor Motions Seismic anchor (support) motions which induce secondary stresses in the piping system are not directly adeessed by the USAS B 31.1.0 and as such were not considered in the analytical evaluation of CR-3 piping. Within the same building structure seismic induced anchor motions were judged to be of such a small magnitude relative to the flexibility of the piping system as to have a negligible effect on design adequacy. Seismic anchor motion need not be considered if using the alternatives 1 and 2 original design criteria. If alternative 3 is used it is recommended that seismic anchor reactions need be considered only in those instances where the vector sum of all such motion at a support exceeds 0.75 inches.
5.1.6 Friction Forces 5.1.6.1 Thermal Indue;d Friction Forces Thermal movement indtced friction forces on supports are not addressed in USAS B 31.1.0 -
1967 and were not considered explicitly in the original analysis of CR-3 piping system and need not be considered if using the alternative 1 or 2 analytical procedure. Section 2.2.3.4 of WRC Bulletin 353M suggests that these thermal induced friction forces which are also deflection limited do not exceed 10 percent of the normal force on the contact surface. Any reanalysis oflarge bore piping using the attemative 3 analytical procedure should consider thermal induced friction forces in the support design.
5.1.6.2 Seismic Indv ! Friction Forces Seismic induced motien of piping on its support contact surface can also produce friction forces..
However, besides being deformation limited forces they are based on dynamic coefficients of friction which typically are 20 percent of the static coefficient". This would result in a dynamic friction force of about 2 percent of the normal contact forces, hence, need not be a design consideration.
5.1.7 Code of Record The piping design Code of Record is USAS B 31.1.0 - 1967. The piping support design Code of Record is AISC Building Specification. Manual,6th Edition, for piping supports designed by analysis and manufacturers catalog valve for piping standard supports as defined in MSS-SP-58.
12
5,1.8 Damping and Seismic Response Spectra Defmition Piping' system design damping used with the response spectra based on the CR-3 median shaped Ground Response Spectra defined in Figure 1 is 0.5 percent. This value of pipe system damping was commonly used with median shaped spectra at the time CR-3 piping was designed.
Since about 1974 a mean plus one standard deviation shaped spectra as defined in R.G.1.60 with 2 percent damping for piping equal to or less than 12 inches nominal damping and 3 percent damping for piping greater than 12 inches (R.G.1.61) has been used in SSE seismic design of safety related NPP piping.
Original floor response spectra generated for CR-3 have been generated using 5.0 percent building damping (except for the CR-3 concrete containment shell and containment internal structure which used 2.0 percent building damping).
In my reanalysis oflarge bore piping using the analytical alternatives 1 and 2, the 0.5 pipirg and 2.0 or 5.0 building damping should be used. For alternative 3, piping damping of 2 or 3 percent (depending on size) should be used with 7.0 percent building damping.
5.1.9 Differential Thennal Modes - USAS B31.1.0.-1967 Piping The thermal analysis is in accordance with the USAS B 31.1.0. - 1967 piping code does not require explicit thermal modes of analysis to determine fatigue usage factors as is required by ASME BPVC Section III NB Class 1 piping analysis. The fatigue concern is controlled by limiting the total number of cycles to 7000 above which the allowable stiesses in the piping are reduced by an f factor as determined by the code. In accordance with the Code of Record (USAS B31.1.0.-1967), there is no requirement to perform thermal mode analysis other than to estimate the total number of cycles and reduce the allowable stress if necessary.
5.1.10 Uplift of Rod Hangars Rod hangers are designed to carry the normal downward weight load of the piping but are not considered to restrain the piping in any other direction. Due to the clearancre in attached parts of the hanger and the potential for buckling if subjected to compression they are not assumed to restrain upward movement of the pipe.
Since rod hangers typically have been provided to restrain the pipe only in the downnard direction, the possibility exists that during a seismic event the pipe could lift off the suppor'.
From an analytica! standpoint this lift-off results in a nonlinearity which cannot be accor.nted for in routine linear analysis. This lift-off could increase the loads on adjacent svppor'.4 and effectively modify the dynamic response of the system.
13 l
l l
1 l
Because CR-3 is a low seismic zone plant it is expected that the upward seismic load on hanger supports will not exceed, in most cases, the downward load of gravit- ln those cases where vertical upward seismic loads are determined in hanger suppons to exec.d the gravity load, the hanger may be removed from the analytical model and the seismic analysis repeated until non-excessive upward seismic load in determined in a support. This approach while not rigorously correct will effectively provide conservative estimates of both seismic stresses in the pipe and loads on supports. Altematively time-history or non-linear analysis could be perfonned to compute resultant pipe stresses and support loads.
This issue was not considered in the original CR-3 seismic analysis and need not be considered in any reanalysis oflarge bore piping using altematives 1 and 2 analytical procedures, but should be considered if alternative 3 is used.
5.1.11 Support Mass Effects in general, supports such as strut:,, sway braces and snubbers where the piping being restrained is relied upon to carry the dead weight of the restraint should be avoided.
In any event, pipe supports should be designed to avoid requiring the pipe to carry excesi e mass from the support, especially in directions other than the direction the support is b-provided to carry load. Table 4 provides recommended maximum values for ' support mass" abv.
which the support mass should be considered explicit!y on the piping model as recommended in Ref. 6. Again, this was not a design issue during the initial piping analysis, and need not be considered if analytical altemative 1 and 2 are usei 5.1.12 Uncinched U-Bolts Acting as Two Way Restraints In general, U-Bolts can be used as two way pipe restraints as shown in Figure 4 provided the total gap in the direction of restraint between the U-bolt and pipe is equal to or less than values shown. In Table 5 are stiffness capacities and directional load capabilities of U-bolts based on tests
- and WRC Bulletin 353
- as shown in Table 5.
5.1.13 Strut / Snubber Angularity Based on analytical evaluations described in Section 5.10 of WRC Bulletin 316 (5) a tolerance of +\- 10 is pemiitted on the nominal angle orientation of struts and snubbers. In instance where the installed angularity exceeds this tolerance the piping system should be reanalyzed using the "as is" str t snubber orientation.
5.1.14 Coi.nputer Program Verification Verification of computer programs used to perform any future analysis or reanalysis of CR 3 safety related piping used as a design basis for such piping shall be in accordance with the requirements of the current issue of the CR-3 Quality Assurance Program Manual.
14
5.1.15 Uses of " Rigid" Versus " Actual" Support Stiffnesses Consistent with the state of the art where CR-3 safety related piping was originally dynamically seismically analyzed, all piping supports were modeled as rigid. Any future dynamic seismic analysis or reanalysis of large bore piping systems may consider horizontal supports rigid where at least 80 percent of the modal effective mass of the piping systems can be shown to have natcral frequencies less than the resonant peak of the applicable Response Spectrum. In such cases it is acceptable to model all such supports as rigid since this assumption would have the conservative effect of shifling the response towards the peak of the spectra and thereby increase scismic loads over what they would otherwise be if support flexibility were considered. Vertical supports are assumed rigid since they are always required to carry at least a 1.0g load, and their relatively close spacing (i.e. one quarter to one sixth lateral support spacing for a MHE (SSE) 0.lg peak ground acceleration plant) typically results in seismic vertical reaction induced piping stresses less than 5 percent of the stresses induced by horizontal motion.
For loadings other than dynamic where the response frequencies of the piping system do not effect the loads seen by the system either rigid or actual support stiffness may be used.
5.1.16 Overlap Criteria Overlap Criteria is generally necessary for two reasons,1) to break up the analytical model to meet analytical model size restrictions imposed by the model size limitation of computer program being used, and 2) termination of the model of small branch piping and Seismic Class 3 portions (non seismic) of a piping system interfacing with Class 1 portions.
The first reason in general is no longer a significant concern because modern computer programs are capable of modeling piping systems between anchors without the necessity to break the model at intermediate points with undefined boundary conditions. However, an overlap criteria is still necessary when analyzing existing piping system models when the model was terminated at points with unknown boundary conditions. In such instances there should be sufficient overlap with adjacent piping system models that the computed system force: and reactions at location of interest do not vary by more than 10 percent.
With respect to reason two, it is sufficient to model the branch piping or non-seismic portion to include only three orthogonal supports. Altematively, if the branch piping is small enough e.g.
the ratio of small bore branch piping moment ofinertia to the large bore run piping is less than 1/25, then the small bore pipe can be modeled simply as a restraint of the large bore pipe. When the ratio is less than 1/50 then the effect of branch pipe on the run pipe can be ignored. Since the non seismic portion of a piping system is not required to be seismically restrained, there often are nct three orthogonal restraints close to the interface with the seismic Class 1 portion of the system. In such instances, it is permissible to terminate the Class 3 portion of the piping system after at least 5 dead weight supports have been included in the non-seismic portion of the model.
See attachment B for a justification of this modeling procedure.
15
5.1.17 Stress Intensification Caused by Branch Lines Intersecting with Run Line Current large bore run line piping analysis generally do not show location of connected branch lines, hence the local effect on stress calculation at the connection. Future large bore at alysis and reanalysis of run piping s!,ould include the mass and stiffness effects of attached branch piping until at least 3 orthogonal supports are encountered or alternatively the momen: ofinertia ratios are less than 1/25 or 1/50 in which case the modeling procedure discussed in Section 5.1,16 may be used.
5.1.18 Use of Unstable Support Configurations To date, no unstable safety related piping support configurations have been found at CR-3.
During the reconciliation of the "as built" to "as designed" phase of the large bore piping reanalysis project any unstable support configurations will be identified and modified as necessary.
5.1.19 Axial Trunnhn Supports on Elto <s This is a special case of Section 5.1.3 where the welded attachment is in the form of a trunnion welded to an elbow side. This is of special concern because the trunnion is attached to the highest stressed portion of the elbow. This issue will be dispositioned in the same manner as described in Section 5.1.3.
5.1.20 Anchor Design Piping anchors are generally designed to resist 6 components of load coming from the pipe associated with axial and shear loads and bending and torsional moments as shown in Figure 5.
During the recondition phase of the Large Bore Piping Reanalysis Program pipe anchors will be reviewed to assure they are capable of providing restraint which is assumed in the analysis. If not, either the pipe anchor will be modified to provide the restraint assumed in the analysis or the analysis will be modified to represent she installed restraint.
5.1.21 Cut Off Temperature for Flexibility Analysis The USAS B 31.1.0 - 1967 Piping Code (Code of Record) does not give a specific temperature limit at which a flexibility analysis is required. Most piping designers have used flexibility analysis limits which range between 150 F and 200 F. The current Crystal River Unit 3 Piping Criteria (11) has established a 150 F as the maximum operating temperature above which e piping flexibility analysis would be required.
It is proposed to raise the Crystal River Unit 3 piping flexibility analysis maximum operating temperature limit from 150
- F to 200 F. This change is in recognition that piping initial installation temperatures at Crystal River, located in mid Florida, are significantly higher than installation temperatures most power plant locations in the U.S. Inherit in the 200 F limit is the 16
i assumption that the initial installation temperature is 80
- F and the differential temperature is 120 F in order to determine resultant restraint of free end displacement loads on the support and a 160 F temperatures range from a lower bound 40
- F temperature to 200
- F temperature in the piping.
In order to demonstrate the suitability of this change in temperature criteria, a sample of 5 typical elevated piping systems should be analyzed assuraing a 40* F to 150
- F temperature range and a 70
- F installation temperature and then reanalyzed assuming a 40 F to 200 F temperature range and an 80
- F installation temperature. It is anticipated that the temperature change ii.duced stresses will increase with the increased temperature limits, but will still be well within the allowable established by the B 31.1.0 - 1967 and AISC-67 Code and Specification.
In performing these analyses it would be permissible to use the actual stiffness at support and azzle locations rather than assume these locations are rigid.
5.1.22 Pre-Loaded Rod Hangers This observation was listed in the Wais Report. However, it is not clear how it is distinguished from the Section 5.10 issue except that it tends to relieve the uplift concern and increase the secondary tensile stress in the rod hangers. It should have no effect on the ultimate tensile strength capacity of the rod nanger.
5.1.23 Spring Hangers are Considered Act i ve in the Seismic Pipe Stress Analysis it is common pipe design practice to igaore spring hanger restraint in seismic analysis of piping.
Such hangers consisting of relatively very soft springs have little effect on the dynamic response of the system, hence, are usually deleted from the analysis for model simplification purposes.
However, such springs are active during seismic excitation, hence their inclusion in the model cannot be considered an error.
5.1.24 Excess Number of Snubbers in Table 1 (1) are shown the average number of snubbers installed in nuclear power plants as a function of the era the plant was constructed. While an accurate count of snubbers has not been performed for CR-3, it is estimated that CR-3 has approximately 250 snubbers which would' place it in line with NPP of the same vintage.
5.1.25 Some Snubbers Installations Consist of Dual Snubbers It is preferable to use a single snubber at a support installation on the piping in order to eliminate any concern with respect to relative load distribution to the multiple snubbers. However, it is not a design requirement any more that multiple or redundant fixed supports are not permitted. In the limit individual members of a redundant support system are expected to behave compositely and shed load to other members of the support system such that support will act compositely.
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5.1.26 Nozzle Loads Nozzle load allowable are divided into the categories,1) passive and 2) active. For nozzle loads on passive components such as run piping, vessels, heat exchanges and tanks at CR-3 the allowable stress in the nozzle due to extemally applied loads to the nozzle for the load case including the MHE(SSE) are limited to 1.2 times the normal allowable stress as defined in USAS B 31.1.0 - 1967 for the nozzle material.
For active components such as valves, pumps and turbines vendor specified nozzle load limits are used. In the eve:a that such information is not available NEMA Std. SM 201958 or API ui0 Stds limits may N used to demonstrate nozzle load adequacy.
Nozzle loads will be evaluated as part of the Reanalysis of Safety Related Large Bore Piping Project.
5.1.27 Equipment Rigidity (Floor Versus Equipment Response Spectra)
In instances where piping is attached to flexible components (fun 6amental frequency of the attached components = 25 Hz) as defined in the FSAR, either the component shall be included in the piping analysis model or an equipment response spectrum defined at the point of piping attachment. This evaluation will be conducted as part of the Reanalysis of Safety Related Large Bore Piping Project.
12 DOCUMENTATION ISSUES 5.2.1 Hard Copy of Pipe Stress Calculations in Files Does Not Always Represent Latest Revision of the Calculation The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assurance Program which will eliminate this issue.
5.2.2 Reconciliation of Calculations with I&E Bul. 79-14 Results Is Not Always Complete The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assur.mee Program which will eliminate this issu6:
5.2.3 Missing Support Calculations The Reanalysis i Safety kelated Large Bore Piping Project will be conducted in accordance with the current en.on of the CR-3 Quality Assurance Program which will eliminate this issue.
I8
5.2.4 Some Stress Calculations Have a Number (in excess of 5) of Discrepancies The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assurance Program which will eliminate this issue.
5.2.5 Seven Different Computer Programs Have Been Used for Pipe Stress Analysis This condition is indicative of the rapid change in piping analysis procedures which have occurred since the initial construction of CR-3 piping. Existing commercially available Nuclear Quality Assurance qualified computer programs are being evaluated with the intent of using such a program as part of the Reanalyu of Safety Related Large Bore Pipe Project.
5.2.6 Uncontrolled Documents The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assurance Program which will eliminate this issue.
5.2,7 Accessibility of Pertinent Documents The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assurance Program which will eliminate this isse-5.2.8 Consistency Between Design Packages The Reanalysis of Safety Related Large Bore Piping Project will be conducted in accordance with the current edition of the CR-3 Quality Assurance Program which will eliminate this issue.
60 SAFETY SIGNIFICANCE OF THE CURRENT DESIGN AND INSTALLATION OF LARGE BORE SAFETY RELATED PIPING AT CR-3 It can not be stated too strongly that there is Hg safety concern with respect to the "as designed" and "as installed" safety related large bore piping at CR-3. Any reanalysis of large bore safety related piping at CR-3 would be to better document the design adequacy of the piping not because there is any concern with regard to the design and installation adequacy or safety of such piping which is much more conservative than would be required by currently applicable criieria.
fd HISTORICAL STUDIES TO DETERMINE SAFETY SIGNIFICANCE AND DESIGN ADEOUACY OF WCLEA.R POWER PLANT SAFETY RELATED PIPING Starting in 1978 the U.S. NRC sponsored a Systematic Evaluation Program to evaluate the
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seismic adequacy of structures, mechanical and electrical distribution systems and mechanical and electrical components in older operating nuclear power plants. There were 11 power plants included in that study all of which received construction permits before 1967. A Senior Seismic Review Team drawn from the Nuclear Industry was convened and charged with the responsibility 19
of " estimating the safety of the selected older nuclear power plants relative to those designed under current (1978) standards, criteria, and procedures and to recommend generally the nature and extent of retroiltting to bring these plants to acceptable levels of capability if they are not already at such levels.""U This review did develop some generic seismic issues with regard to active mechanical and electrical components and electrical distribution systems (cable trays) as vell as vertical tank and heat exchange supports. However, detailed evaluation of nmples of piping in these plants indicated there were no safety concerns with respect to such piping except
'or the need of piping flexibility between adjacent structures. The results of this Systematic Evaluation Program ultimately lead to the Unresolved Safety Issue A-46"O seismic review of all older nuclear power plant safety related to activunechanical and electrical equipment, cable trays and supports for tanks and heat exchangers. Excluded from the detailed A-46 effort was safety related piping because of the experience gained in the Systematic Evaluation and other programs.
This conclusion regarding the seismic safety ofinstalled safety related piping in older nuclear power plants was confirmed in NUREG/CR-4334"') in August 1985 which except for the evaluation of the piping flexibility between buildings did not recommend a need for detailed walkdown evaluation of piping when the design basis peak ground acceleration is below 0.5g.
This value of 0.5g is 5 times the SSE equivalent peak ground acceleration at CR-3 of 0.lg.
fd COMPARISON OF THE SAFETY SIGNIFICANCE OF INSTALLED SAFETY RELATED PIPING IN OLDER NULCEAR POWER PLANTS AND THAT FOUND INSTALLED IN CR-3 A brief walkdown of large bore safety related piping installed in CR-3 similar to those walkdowns performed for piping in the SEP plants and in approximately 20 other A-46 plants did not result in the observation of any safety significant issues. Lateral support spacings were somewhat longer than in plants typically designed for 0.lg OBE or 0.2g SSE 1 due primarily to the fact that the CR-3 plant safety related SSC consistent with the very low seismicity site are designed to only an 0.05g OBE and 0.10g SSE equivalent loading.
It should also be noted that CR-3 has the second lowest probalistically defined mean design basis ground acceleration levels of all nuclear power plants in the U.S. CR-3 has a 0.65x10"/yr probability"" for a mean peak ground acceleration of 0.lg. This probability value is more conservative by a factor of more than 2.0 as compt ed to most nuclear power plants operating in the U.S. today and is less than the NRC's A-46 reevaluation criteria of 10"/yr, a criteria was compared to the B31.1.0 designed piping it addition, CR-3 large bw. mfety r ated observed in 8 power plants in t.- )rnia with approximately 400,000 feet oflarge bore pipe at risk"D. These plants have experie..ced at least one of a total of 10 strong motion earthquake peak ground accelerations in excess of 0.2g (twice the CR-3 SSE equivalent) which have effected nuclear power plants in California since 1952. A total of 6 large bore pipe failures cccurred in more than 400,000 feet of piping at risk in these 8 power plants. It should also be imderstood that most of the piping in these plants was not designed to resist any seismic load. The large bore safety related piping at Crystal River 3 is installed with considerably more seismic lateral l
l 20 1
l 1
I support ten that observed at the 8 power stations in California which successfully resisted l seismic leve3 at least twice that used in design of CR-3 piping.
As a result of these installation comparisons it can be concluded that CR-3 safety related large bore piping is safe in a relative sense when compared to other nuclear power plants in the U.S.
and is safe in an absolute sense when compared to lar);e bore power plant piping which has expedenced seismic excitations at least twice that which CR 3 piping is required to resist.
QJ COMPARISON OF THE S AFETY MARGINS REOUIRED IN CURRENT DESIGN OF SAFETY RELATED PIPING IN NUCLEAR POWER PLANTS AND THAT USED AT-CR-3 From a safety significance or design basis probability of failure stand point, existing CR-3 safety related large bore piping is much more resistant to seismic induced pipe failure than new nuclear power piping design currently required by the ASME and NRC pipe design criteria.
A task committee of the Design Subgroup of the Subcommittee on Nuclear Power (Section III) of the ASME Boiler and Pressure Code has been reviewing piping component seismic test data for the past two years in an attempt to establish design margins associated with current design cf safety related nuclear power plant piping. The allowable stress in ASME Class 2 and 3 piping is limited to 3.0S, as required Fy the NRC (since 1994 ASME has permitted a value of 4.5S,).
With the allowable stress limit of 3.0S, the test results indicate that there is a margin of at least 2 against a 10-2/yr probability of failure induced by a design basis earthquake (SSE) using a R.G.I.60 i'ased (spectral shape) and ASME Code Case N411 (damping).<2n 4
When couple,i with the conditional mean probability of a design basis earthquake of 10 /yr this results in an estimated overall seismic induced probability of failure of safety related piping of 4 4 between 10 to 10 /yr. This failure mode thus would contribute much less than one percent to the overall probability of core melt in nuclear power plants which typically ranges between 10" to 10'/yr in operating nuclear power plants. This failure contribution is considered negligible and therefore an acceptable design basis.
With regard to margins associated with CR-3 large bore piping, the seismic it.put as a function of spectral sitape and damping values is approximately equal to the current R.G.1.60 and ASME Code Case N-411 criteria as shown in Figure 1. When conve-ted to floor spectra for design purposes the R.G.I.60 and Code Case N-411 compatible criteria typically yield floor spectra magnitudes which range from 0.67 to 0,5 of those used in the design of CR-3 piping.
The mean probability of occurrence of the SSE equivalent carthquake as defined by the peak-ground acceleration for CR-3 is 0.65 x 10"/yr '2D which is considerably less than the NRC's A-46 reevaluation criteria of 1 x 10"/yr. In addition, the allowable stress in CR-3 safety related large bore piping under the SSE equivalent earthquake is 1.2S, as compared to the 3.0S, currently permitted by both the ASME and the NRC design criteria for the SSE.
21
CR-3 piping design, because of the excessive conservatism contained in both the definition of the SSE equivalent zero period ground acceleration, the design basis floor spectra and the allowable piping stresses, results in piping failure probabilities which are several orders of magnitude less than currently required by ASME and NRC piping design criteria and therefore has resulted in an extremely safe design.
23
SUMMARY
AND CONCLUSION After review of the issues identified in Table 2 and 3 and Section 5 of this report, it can be concluded that CR-3 large bore safety related piping has been designed to a much more conservative criteria and installed in a manner similar to safety related piping of some 50 other operating nuclear power plants of the same vintage. The seismic safety of safety related piping in such plants has been validated by NRC studies conducted in the mid 1980's N and by walkdowns of power plant piping systems designed to the same construction code as used in CR-3 which have experienced actual earthquake motions at least twice as intense as that specified for design of CR-3"0 However, the detailed documentation of the design basis and configuration management of"as built" piping systems does not meet current quality assurance expectations for safety related nuclear power plant piping at CR-3, For this reason, it is recommended that a reanalysis oflarge bore safety related piping be performed in order to document design adequacy of the existing piping and piping supports.
Of the 27 technical issues identified in Tables 2 and 3, and addressed in Section 5.1, Nine of them have been determined to beyond the sta'e-of the-art of the Code of Record used in the construction of the safety class pipe. The other ssues would be addressed as part of a reanalysis effort or in separate effects studies. In additica to the comprehensive large bore project there should be 4 separate effects studies conducted which include:
(1) Evaluation of Wcided Attachmen s (2) Effect of Branch Pipe on Run Pipe (3) Overlap Evaluations:t1 (4) Nozzle Load Allowables
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in order to validate the assumptions and provide limiting loads or capacities as input to a reanalysis project.
2 See Attachment B for a preliminary evaluation.
22
Table 1- Typical Quantities of Piping, Piping Supports and Unit Manhours for PWR Nuclear Power Plant Designed for a 0.lg OBE PGA (From Ref.1)
Construction Period - Commercial Operation ist Era 2nd Era 3rd Era 1967-74 1974 80 1981 - Present (900 MWe) (1100 MWe) (1200 MWe)
A. TOTAL FEET OF PIPE 125800 233600 319700
- 1. Safety Related
- a. Large Bore 35600 70400 107000
- b. Small Bore 50200 93200 112700
- 2. Non Safety Related 40000 70000 100000 B. TOTAL NUMBER OF 11850 23520 30210 SUPPORTS
- 1. Safety Related
- a. Dead Weight 6980 12270 14,020 "'
- b. Seismic I - Fixed 1990 5630 7,080 * )
ii. - Snubber 200 950 1500
- 2. Non Safety Related 2680 4670 6660 C. UNIT MANHOURS SAFETY RELATED ,
- 1. Piping MH/11
- a. Engineering 1.2 2 7
- b. Crails _
- 2. Support MH/Sup
- a. Engineering 4 15 38
Table 1 Notes:
- 1. 1st Era - Turnkey Era -
2nd Era - Regulatory Guide, Formal Quality Assurance and Pipe Break Era-
- 3rd Era "As Designed" and "As Constructed" Reconciliation Era
- 2. These quantities are best estimate values. A coefficient of variation of
+0.25 and .15 is suggested to define variability.
- 3. The quantitles shown include approximately 1000 additional rigid seismic supports and 300 snubbers which would be required on non safety related piping to accommodate the 2 over 1 issue.
- 4. The. support quantities shown do not assume the use of a significant number of dual supports, deadweight plus seismic nor does it assume the use of a significant number of ganged supports where a single support stnicture is used to support several piping systems.
- 5. For safety related and seismically supported 2/1 piping, the average ratio of the number of deadweight to lateral supports in the first era is 3.2. In the second and third era, the ratio dropped to approximately 1.8.
24
.- _ .. - ,. - _- ~. _ --__ . - - . - - - - - - . -. - ..-. . --- .
1 Table 2 i
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- ISSUES OBSERVED RELATED TO PIPING AND P2ING SUPPORT DESIGN -
The following list contains issues that were observed during the review of Crystal River-
.3. piping and piping . supports by Dr. E. Wais. * -
-*- Explicit modeling of valves; WRC Bul 300 Sect 3.2
+-
-x
_ Zero Period Acceleration (ZPA) consideration (missing mass correction)
- Analytical evaluation of welded attachments; WRC Bul 353, Sect 2.3.5
- Clearances; WRC Bul 316, Sect 3.0; WRC Bul 353, Sect. 2.3.3
- Seismic Anchor Motions (SAMs); _ .
- Consideration of friction forces; WRC.Bul 353, Sect. 2.2.3.4 x +
Code of Record x +
- Damping; WRC Bul 316 x-
- Seismic response spectra definition
- - Differential thermal modes-B 31.1 piping
- __ Uplift on rod hangers; WRC Bul 353, Sect 2.3.6
- Support mass effects; WRC Bul 353, Sect 2.3.4
- Uncinched U-bolts acting as two way restraints; WRC Bul 353, Sect 2.4.5 x + a
- Strut / snubbers angularity; WRC Bul 316, Sect 5.10 x -
Computer program verification x ** Use of " rigid" versus " actual" support stiffness; WRC Bul 353, Sect 2.3.2 x .
Overlap Criteria and Seismic Class I and II Interface Trunc: tion x -
Stress Intensification Factor (SIF) selection x +
Use of nonstandard (unstable) supports; WRC Bul 353, Sect 2.4 x ** Axial trunnion supports on elbows; WRC Bul 353, Sect 2.3.5 x -
Anchor desigt x -
Temperature cut off for requiring flexibility analysis Rod Hangers have been preloaded by tightening the turnbuckle on the rod to offset upward loads.
-x -
Spring hangers are considered active in the seismic pipe stress analysis There is an excess of snubbers.
- Some supports consist of dual snubbers.
-x -
Missing support calculation x: -
The hard copy of the pipe stress analysis calculations in the files does not always represent the latest revision of the calculation.
=
x +
. The reconciliation of the calculations with the Bulletin 79-14 results is not always
- complete.
' x Some stress- calculations have a number (in excess _ of five) of identified discrepancies.
x -
Spring hangers are considered active in the seismic pipe stress analysis.
+
- x. -A total of seven different computer programs have been used for pipe stress analysis.
- indicates an issue that was not part of the state of 'he art at the time CR-3 piping was analyzed.
adicates an issue which would be addressed in whole or in part as a part of the large bore reanalysis Jnless strut or snubber installed angularity exceeds 10? of that used in the analytical model.
25 am- e 'm-se
= -4've/ +r 4 % r -- +-"'Wm m ei 3 'w+w.-
m t+ eir- etn4w -w y g +pr -
9 y-+m-g" +--7rrs-T'-
Table 3 r,
OTHER ISSUES RELATED TO PIPING AND PIPING SUPPORTS AT CR-3 x
- Uncontrolled Documents x
- Accessibility of Pertinent Documents x
- Consistency Between Design Packages x+ + Nozzle Loads x
- Equipment higidity (Floor Vs. Equipment Response Spectra) x
- Unanalyzed Piping x*
- Branch Line Effects on Large Bore Pipe
' indicates an issue that was not part of the state-of-the-art at the time CR-3 was designed.
x indicates an issue which should be addressed as part of the large bore reanalysis.
+ nozzle loads were typically transmitted to the component vendor for his evaluation ad acceptance.
26
Table 4' Recommended Limit on Added Support Weight Before it Should be -
Considered in the Piping Analysis Size Sched. B31.1 Span Weight of Max Added L (ft) Span W (Ib) Weight Wi(Ib) 2 1/2 40 11 87 10 80 105 13 160 127 15 4 40 14 228 27 80 280 34 160 371 45 8 40- 19 954 114 80 1201 144 160 1719 206 12 STD 23 2244 269 X 2584 310 160 4489 539 16 STD(30) 27 3826 459 X(40) 4302 516 160 8125 975 20 STD(20) 30 6128 735 X(30) 6810 817 160 14003 1680 ~
24 STD(20). 32 8914 1070 X 9772 1173 160 21401 2568 27
Table 5 - U-Bolt Lateral Stiffnesses and Inline U-Bolt and Lateral level Capacity for the Load Case Including-the SSE Size Lateral Stiffness S S E Load Capacity lbs.
Rod (in.)xPipe (in.) (lb/in.) . ("
- 1/4xl 54,498 640 220 3/8xl-1/2 91,259 1600 520
. 3/8x2 46,274 1600 380 1/2x2-1/2 83,248 2930 800 1/2x3 46,141 2930 620 1/2x4 21,710 2930 460 5/8x6 4800 610 5/8x8 4800 460 3/4x10 7180 630 7/8x12 10000 850 (1) Based on linear elastic analysis for material with FY=36ksi.
(2) Lateral load capacity limited by Laboratory Test to .024 inches lateral deflection.
For lateral load limited to 0.125 inches multiply values given by 0.75.
l 28
FIGURE 1.
FLORIDA POWER CORPORATION MULTIPLE DAMPlNG VALUES CRYSTAL RIVER #3 NORMALIZED AT 0.1G GROUND RESPONSE SPECTRA HORIZONTAL MHE (SSE) 0.6 _
0.5 _
$ 0.4 , [,sx
~
5 g
E
/
l/
a "Ok N snk g 0.3 --
j -
sj ~
t 3 g _
(
g N e 0.2-:
o
)
)
J \ - --
0.1 pr/
p] W %
- Y [
0 ,
1 0.1 ,
1 5 10 100 FREQUENCY (Hz)
DAMPING: R.G.1. 60 +2% -X-3% -El-5.0%
CR-3 GRS G1/2%
.. 1.4 e
5% DAMPING 1.2 - EXCEPT AS INDICATED m-2 9
F-1- i 4
12: 1.5 X Bounding Spectrum W
- J 0.8 -
m O
O j 0.6 - Bounding 4 Spectrum cc
$ 0.4 - .36
.3_5-w .24 CR-3 Ground Response Spectrum:
a W .2 '16 0.5% Damped 15 0.2 - e 5.0% Damped 0 i i i i i .
0 5 10 15 20 25 30 35 FREQUENCY (Hz)
FREQUENCY (Hz) 2.0 2.5 7.5 8.0 10 12 16 20 28 33 1,5 X Bounding Spectrum (g) .98 1.2 1.2 1.13 .90 .80 .68 .59 .53 .50 Bounding Spectrum (g) .65 .80 .80 .75 .60 .53 .45 .39 .35 .33 0.Sti Damping CR-3 MHE Design Spectrum (g) .36 .36 - - .24 - - .13 . .10 5.03e Damping CR 3 MHE Design Spectrum (g) .19 .18 - - .11 - - .10 - .10 FIGURE 2. Seismic Capacity Bounding Spectrum 5% Damped Based on Earthquake Experience Data (Source: Reference 5) Compared to CR-5 Design Spectra 5 and 0.5 Percent Damped 30-
VALVE OPERATOR CANTILEVER LIMITS 120 Applicable to Motor and Piston Operated Valves
Applicable to Air Operated Diaphram 1 100_ and Spring Operated Pressure '
a Relief Valves 750# "
O F
Cc in This Area Experience O 80- Cannot Be C;ed For 750#
Q Valve Qualification y See Note (1) ct O 60- ;---- 400y O l F i CE 45- '------ ' in This Area Valves Are Qualified By
% 40- ! 200# Experience z i 30-
$ 25-l 100#
g 75#
- o. 20-c.
O i .
a 0' 2 4 6 8 10 12 14 16 18
~
PIPE DIAMETER (in)
(1) Approximate Maximum operatorWeights Given forVarious Ranges of Pipe Diameter FIGURE 3. Valve Operator Qi alified Cantilever Length Limits for Power or Flui; Operated Valves 31
.. .: 9 2- 93
%- "77Mh n-7( jf n x
-m - ~
/
92 93
' wi g '
A N s 4..
[/
1
' h -Q lh
. f. 'd. p*.
n-1 J0 n T W w o'
. .'y'o,
'. . N
..y - .-9 ,:
,g
.. , 'h:'-
w 0.0 < (g 2 + 93 ) s 0.25 in.
0.0 < g3 s 0.125 in.
FIGURE 4. U-Bolts Used as Two Way Restraints 32'
m r
,- Fi - = Axial Force From Attachment -->
or Branch Pipe Along 11 Axis.-
F2 = Shear Force From Attachment.- r or. Branch Pipe Along 2 2 Axis -
- Fa = Shear Force From Attachment -
or Branch Pipe Along 3-3 Axis BRANCH PIPE
/ OR ATTACHMENT D1 M3= Torsional Moment From ,/
- Attachment or Branch Pipe -
About the 1-1 Axis- 1 EQUIPMENT SHEL'_
OR RUN PIPE M2= Bending Moment From F1 Attachment or Branch Pipe 2
-About the 2-2 Axis- 3 s, J M2 F2
- M3 = Bending Moment From 'N.N , M1 Attachment or Branch Pipe About the 3 3 Axis t-1 P M3 . -
/' l Fa \
2' l-3
\l '
l A' .'.
1
',/. l\ \
l V ,/ \
f
/ '\
t2 ,e ,/
k
\. !,, '; -
' L
'n .3'\j ,
d2 's \.//
D2 -
\
i FIGURE 5. External Attachment or Branch Line Loads on Equipment Shell or Run Pipe
-33
ATIMCIIMENT A piping system. The design of other safety llistuleal 7hvelopment of the Nueicar class piping systems was normally Piping Design Process performed on a quasi static basis using lateral restraint loads based on the peak of Common practice associated with the ground response statically applied to pipe and pipe support design in the mess distribution of the pipe in conventional high energy or high s:veral instances, pipe lateral st.pports were temperature design and in me nuclear provided on spacings so as to insure the industry prior to the late 1960's was for fundamental local frequency response (as a the A/E to prepare single line piping layout simple beam) would be in the unsmplified overlays basert on judgc.nent and practical recponse spectra range. Static loads on experience. These single line drawings supports were also detennined by were provided to the manufacturers of simplified quasi static analysis. General standard supports. The manufacturers application of floor instead of ground usually performed the piping analysis and response spectra did not begin until 1970.
designed the supports in general, limited Not until the development of flexibility analyses were perfonned on the generally available dynamic analysis piping system and special supports were algonthms staning around 1968 was it designed using c:.aple stetic and strength of possible to perform dynamic analysis of material fonnulas. Component standard piping systems on a routine basis. Such support were quali6ed by load rating dynamic analyses vwre typically 'imited to which determined experimentally the load reactor coolant, main steam and feedwater at which the support would fail. *ihis load systems. All other seismic piping system wss typically divided by a factor of 5 and analysis at the time was limited to quasi-the resultant load became the static based on inertia forces sssociated manufacturer's catalog or MSS SP 58 e;ther with the peak or the rigid range of design value for the support, the ground response spectra. About 1969, Until 1967, the design of nuclear actual stress analysis of supports began.
piping was per' 'ed primarily u ing rule Prior to 1969, piping suppons were of thumb layou ign procedares and pre- typically qualified by load rating or generic analyzed piping nguration data in analysis of the catalog pipe support with a tabelar form. safety factor of 5.
Starting with the publication of the The initial publication of ANSI B31.1-1967 Code and the general ANSIB31.7 in 1969 and its subsequent availability of static computer programs incorporation as NB 3600 for Class I such as MEL-21, it became practical to piping and NC and ND 3600 for Class 2 perform static stress analyses of complex and 3 piping systems in ASME Section ill piping systems subjected to thennal 1971 Ccde required a more rigorous expansion, analysis of piping systems. In 1975, the in the late 1960's, scismic static initial publication of ASME Section ill NF analysis was performed indepenoently of extended rigorous analysis to individual desd weight and thennat analysis. The pipe supports, first application of dynamic seismic These new Codes along v'ith other evaluation was on the reactor coolant regulatory requirements have called the 34
Attachment Alntnjd analytical effon associated with qualification of ASME Safety Class piping and supports e grow steadily. In 1967, this total effort was estimated at $70,000 or 3,500 rnanhours for a nominal 800 MWe nuclear power plant. Today, the effort has grown 1,000 fold to $70,000,000 or 1,750,000 manhours.
[
35
f i
r i
1 -ATTACHMENT B f Effect of Seismic Class 3 Rod Hung Piping on Seismic Class 1 Interface Piping and j Support Levels !
P B.I.0 Introduction i
For Crystal River 3 NPP there is an FSAR commitment to model the Seismic Class 3 portion of a piping system with the Class 1 portion of the line. This modeling of the Class 3 l portion of the line is limited to three onhogonal supports in order to include the loading ,
effects of the Class 3 line on the Class 1 3 interface pipe and suppon. In many instances it l
- is impractical to perform the Class 3 line modeling because this Class 3 portion of the line ;
which is not designed to be earthquake resistant often consists of very long straight runs -
which are supported by_ vertical rod hangers. A computer study was conducted in order to- ,
determine the effective loading on interface piping and supports when the attached rod hung '
i support, pandulum effect is and is not considered. ;
5 B.2.0 The Analytical Model j Initially a 4 hanger suppon model for a 4.0" nominal diameter schedule 40 straight j pipe filled with water was prepared for static analysis by the CAE PIPE piping analysis ;
program. The boundary conditions include a fixed left end which represented the boundary between the Class 1 and Class 3 piping segments with three different conditions assumed for the right end boundary (fixed, pinned, free) of the Class 3 piping. These three conditions <
represent any additional piping and supports which may be present down stream from the ;
modeled Class 3 piping segments as shown in Figures 2.1 to 2.3.
The hanger supports shown are assumed to carry lateral load by pendulum action by -
the relationship:
Eg. (B.1) !
K,= M where:
K,= lateral stiffness of suppon in lbs/in. i W=the weight of the pipe acting on a single hanger in Ibs.
L= length of the hanger in inches .
i (two cases were -"Muated L-36.0 in, and L=72.0 in.)
i 36 I I
w-- -g y ygv*-ar- ge y- - w g yMPgr ye Tw--*tv'-Ft-*'t--W-"Y"'t-Wtr*h-*yte gr y+ w- w' ir yr% uwywga--'y *
- A constant 1000 lb. lateral load was applied at the centroid of the modeled segment which was 35.0 0 from the left fixed end. Additional models were developed for six and eight l hanger support systems with the same three right end boundary conditions and hanger lengths and 1000 lb. lateral load applied at the centroid of the modeled segment.
11.3.0 - Analytical Results and Conclusions The analytical results are summarized in Table 3.1. The results clearly show that if at least 4 hanger supports e.re in:!nha in the Class 3 piping segment models and the average hanger rod length is nt c. ore than 36 inches, then the reaction loads on the interface were at least equal to what the reaction levels would be if the pendulum effect was ignored and the ,
right boundary of the Class 3 piping segment is fixed or pinned (cha acteristic of 3 way restraint). For hanger rods with an average length of no more than 72 inches the reaction forces would be at most 15 percent higher than if the pendulum effect was ignored and right boundary condition is fixed or pinned.
The boundary moment at the left fixed end of the 4 hanger arrangement (average 36 inches long) if the pendulum effect were ignored is 8750 in. lbs. The maximum moment computed for the 4 supports with pendulum action was 10256 in. . lbs., which is 1.17 times the limiting moment value using the three orthogonal support right end condition. For the 72 inch long case 14165/8750=1.61. In this case the use of 6 vertical hanger model with a right free end condition would be necessary to conservatively determine left end mo,1ents on the interface piping.
?
)
. 17
Table B 1 Seismic Class 3 Interface Loads Summary of Results ;
A. Loads on Interface Support Assumed Fixed 36" Long lianger Case 1 - 4 Ilanger Supports X Y Z M, hty hi, Free Right End 0 122 416 0 10256 35 ,
Pinned Right End 0 118 500- 0 9551 18 Fixed IUght End 0 118 403 0 7277 18 No Pendulum Effects PL/4 500 8750 with P=1000 lbs. L=35.0 ft.
Case II 6 Ilanger Supports Free Right End 0 113 233 0 7256 0 Pinned Right End 0 113 245 0 7423 5 Fixed Right End 0 113 240 0 6706 5 No Pendulum Effects PL/4 500 12250 with P=1000 lbs. L=49.0 ft.
Case 111 8 Ilanger Supports Free Right End 0 114 83 0 4468 1 Pinned Right End 0 114 79 0 4341 1 Fixed Right End 0 114 87 0 4332 1 No Pendulum Effects PL/4 500 15750 P=1000 lbs. L=63.0 ft.
B. Loads on Interfacc Supports -- Assumed Fixed - 72" long IIanger _
Case 1 - 4 llanger Supports X Y Z M, M, M, Free Right End 0 122 520 0 14165 35 ,
Pinned Right End 0 118 570 0 11035 18 Fixed Right Ewl 0 118 447 0 7942 18 3 No Pendulum Effect 500 8700 ,
l 1
38 l
Case 11. 6 Ilanger Supports X Y Z M. M, M,
',.' Free Right End 0 113 331 0 10354 4 l Pinned Right End 0 113 373 0 10590 5 Fixed Right End 0 113 354 0 8734 5
- No Pendulum Effect 500 42250 l Case 111 - 811 anger Supports ;
Free Right End 0 114 184 0 7633 1 Pinned Right End 0 114 185 0 7658 1 Fixed Right End 0 114 189 0 7201 1 No Pciaulum Effect 500 15750 t
M 39
iA- - n- ~-~- =A w _s - LA,---- a - ., a. m 4,s.a,y -w-- a s4-n& S- -,4-,.1 & 4 rm l
,. FIGURE B.1 l l
t i
s 9*~
a
~~g w~ sL,:_ s Qy ~ + . e ..<
g p>x/ '
4 ~~
dg s -
~
\' ~sL(~s
,,,- \
40
. _ - _ . .. . . .-. . .. - = . -. . . . ..
P FIGURE B.2 q
..i i
^
EXAMPLE PROBLEM 2 PINNE y.
a
\'N14'0" g >^4
'N y N' \ 4 ,go 1808 N*' ,
y s\ ,0" N1 -
% '\.14 ' 0 "
s.,\s DWT RUN- PIPE UGT. ONLi' LENGTH (FT'IN") FORCES.MONENTS(LB.FT-LE 41 y w te me r ,--ep e ,e-r w-o- w- m-w.v-' , ~ - , + -
t FIGURE B.3 s
I r
EXAMPLE ~ PROBLEM 3 FIXED <-
Y
[.
,448g" gy q
\
d' n M 4,8,, #'"
L
- p
- g" Yp 1000
'sq .
N N14' 0" -
\ 1 N,.\'4,0" 1
s.,,s, i DWT RUN- PIPE UGI. ONLY LENGTH (FT'IN") FORCES, MOMENTS (LB.FT-LB) 42
ATTACHMENT D FPC's Large Bore Piping and Pipe Support Action Plan i
, -- ...,y._-_ .. , ,.. ._ _ _ , . . . . . - , _ _ - . _ , . , . . _ _ , _ . . . - - . , _ , . . , . - , . - - , ,
e Larne Bore Pinian and Pine Suonert Action Plan I Pceliminary Short Term Action Items: ,
l Task - Description Projected j
- Completion l 1 Develop Master List - Required for scope defmition. Mark-up Flow Diagrams to 1/9/98 I capture all large bore safety related piping. Review against FSAR safety related component list and PI 305 drawings. ;
2 Develop Walkdown Criteria and Procedure - resiew of NRC Bulletin 7914 (and 1/23/98 {
supplements) walkdown attributes should be included (pipe run geometry, support i function, location, clearances, embedments significant lumped masses, etc.) and ;
expanded as required. =l 3 Retrieve Documents from Parsons - Retricyc any support calculations in Parsons' 1/30/98 !
possession and not on file at CR 3 (mainly anchorage cales). l 4 Review Licensing Commitments - A review oflicensing commitments made related 1/30/98 {
to piping and supports will be perfonned. This includes any FSAR and SER i commitments as well as commitments made through other NRC correspondence. L
- 5 Develop a Reconciliation Design Guide - In order to consistently apply analytical 2/13/98 q techniques and methodologies developed during this program, a Reconciliation Design l Guide will be developed. The document will be a reconstruction of the original j analytical techniques employed during the liceming and constmetion of CR 3 (and used by Gilbert Commomvealth) and will provide guidance in the development of reconciliation calculations. In addition to a historical reconstruction, this criteria will ,
incorporate results from the licensing commitment resiew. ;
6 lasue Project Scoping Report At the completion of the items above, a summary 2/27/98 ,
scoping report will be developed that will formally define the smpe, cost and scheduled completion of this piping and support reevaluation cfrort. :
7 Third Party Assessment of Task Numbers 1 - 6 Ongoing 4
Preliminary Long Term Actions:
Outage / Cycle Description Projected Period Completion Fuel Cycle 10 ' 1) Walkdown areas outside of Reactor Building (RB) and cold pipe and update September (Jan 98 to Sep 99) drawings. 1999'
- 2) Begin/ continue analysis of piping and supports. -i
- 3) Design Modifications identified for Code Compliance.
- 4) Install Modifications that can be donc non outage.
Outage 1IR 1) Start RB and flot Pipc Walkdowns then update drawings. October *
(Oct 99) 2) Install Modifications that are outage related. 1999
' Fuel Cycle Ii 1) Complete walk 6 owns of areas outside of RB and cold pipe then update drawings. October ,
(Nov 99 Oct 01) 2) Design Modifications identified for Code Compliance. 2001
- 3) Install Modifications that can be donc non-outage. !
Outage 12R 1) Complete RB and llot Pipe Walkdowns. November (Nov 01) 2) Install Outage Related Modifications. 2001 Fuct Cycle 12 1) Complete all analyses and modification packages. December ;
(Dec 01 - Dec 03) 2) install modifications that can be completed non-outage. 2003 Outage 13R 1) Complete remainder of code compliance modifications. January (Jan 04) 2004 rw e- v-- wi m ----e-e.cm-iWa 4 -.r----*-w = se m t 'o-s--ee-w-+ia .<+v--m-u-+-c-awve of=>se-m- , -----,evr r,= .- -es-+--,-s a4-.-'-e ao y, =p r'.i -e c-r-, --9, *--Te ir *'Trw v V '-"!