ML19241B279

Rept on Reanalysis of Safety-Related Piping Sys,Revision 1, by S&W

ML19241B279
Person / Time
Site:	Beaver Valley
Issue date:	07/11/1979
From:	DUQUESNE LIGHT CO.
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NUDOCS 7907130413
Download: ML19241B279 (100)

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3EAVER VALLEY POWER STATION, UNIT 1 REPORT ON THE REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS FOR BEAVER VALLEY UNIT 1 DUQUESNE LIGHT COMPANY ORIGINAL - JUNE 15, 1979 REVISION 1 - JULY 11, 1979 541 042 Stone & Webster Engineering Corporation Boston, Massachusetts g- g)P 7907130 %4

BEAVER VALLEY POWER STATION, UNIT 1 If a seismic event which results in accelerations greater than acceleration

~

level of 0.0lg occurs during the period of interim operation, the plant vill be shut down for inspection of those piping systems and supports which have not been shown to be fully acceptable for the OBE case. As discussed in the FSAR, Section 5.2.8.1, the accelerometers are initiated and recording started at a setpoint of 0.01 g acceleration. All seismic monitoring instrumentation is demonstrated operable in accordance with the test methods and testing frequencies specified in Table 4.3-4 of the Technical Specifications. The seismic instrumentation vill be checked prior to startup.

This report addresses details of the analysis work, results of pipe and support analyses to date, presents a discourse on conservatisms, and discusses other topics within the scope of the reanalysis task. The report represents all work to date and is in addition to other submittals previously forwarded since the Order to Show Cause.

The seismic reanalysis is based on piping analysis programs, SHOCK 3 and NUPIPE, that use methodology currently acceptable to the NRC. The results to date indicate that the subject systems will be able to perform their intended safety functions under the maximum seismic conditions specified in the Final Safety Analysis Report. The reanalysis effort has 1strated the conservative nature of the original seismic analysis. The piping systems have been found to be impacted only slightly after thorough, rigorous reanalysis.

541 043 1-2 Revision 1

3EAVER VALLEY FO*JER STATION, U!iIT 1 Results to date also show that no piping of any size vill have to be replaced or repaired.

Abbreviations used in this report are defined in Table 1-1.

54! 044 1-3 Revision 1

BEAVER VALLEY F0k'ER STATION, UNIT 1 TABLE l-1 A3BREVIATIONS S  : Pressure Stress S

DL : Deadicad Stress Sh: All wable Stress at Maximum (Hot) Temperature SOBET = Total Stress under OBE Condition S # *

DBET S #

• DBEI S : Allowable Stress a

S :

y Yield Stre Sth S = Ultimate Strength u

S g : Thermal Stress 541 045 1 of 1 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 SECTION 3 RESPONSES TO NRC LETTERS AND ADDITIONAL QUESTIONS The following four questions were raised by NRC personnel during a visit to the Beaver Valley Unit 1 project at S&W, June 5-7, 1979. Each NRC question is followed by the responsa.

NRC Ouestions

1. Indicate the frequency range over which the new SSI-ARS is not enveloped by the previous spectra. Discuss the effect this has on conponents, equipment, and piping analyzed to the old spectra.

Reseense The problems listed below with the system piping frequency and period use the old ARS curve as the run of record. A review of the curves included in this section which indicated a comparison between the peak spread SSI curve vs the old ARS curve shows that none of these problems except as noted fall into the period range where the SSI curve is not enveloped by the old ARS curve.

541 046 3-1

BEAVER VALLEY PO'a*FR STATION, UNIT 1 Problem Frequency Period No. (cycles / seconds) (seconds) 100 9.16 .11 179 10.87 .09 215 5.42 .18 101 4.98 .20 3063 9.51 .11 204 13.17 .08 785 3.78 .26 157 13.42 .07 158 23.95 04 212 10.47 .10 228 8.71 .11 229 9.41 .11 2112 3.31 .30 610 16.22 .06 612 16.66 .06 3011 3.87 .25 1 5.73 .18 Problems 785, 3011, and 1 presently fall into the area where the SSI curve is I

not enveloped by the old ARS.

,P.. 047 3-2 Revision 1

BEAVER VALLEY p0WER STATION, UNIT 1 A review of Problem No. 785 (Feedvater System) indicates that 72.6 percent of the allowable seismic OBE stress was attained using the original amplified response spectrum. Therefore, a substantial increase, 1.38 times for the OBE, would still be acceptable. The portion of the SSE curve for the horizontal earthquakes that exceeds the acceleration values of the original ARS is not seen by the piping system. For the vertical earthquake, the increase in acceleration is 20 percent which would still result in acceptable stress levels. For the DBE case, the horizontal accelerations increase 1.4 times and the vertical accelerations increase 1.2 times. These values are seen by the piping system and would result in stress levels below the allovable stress.

Problem No. 785:  ; S SLP + SDL = 4832 h = 15000 S p+SDL OBET: 14397  ; 1.2 S h 18000 S p+SDL + DBET: 20232  ; 1.8 S h

27000 A review of Problem No. 3011 (Residual Heat Removal System) indicates that 98 percent of the allowable seismic OBE stress was attained using the original amplified response spectrum. For the DBE case, only 62 percent of the allovable stress was exhausted. Therefore, an increase of 1.02 times for the OBE and 1.61 times for the DBE case would be acceptable. A comparison of the original ARS with the SSI-ARS indicates that the acceleration values of the SSI-ARS not bounded by the original ARS vere in a frequency range not experienced by the piping system. The 54i 048 3-3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 only exception to this is for the DBE case where the acceleration values of the Y-direction earthquake increased. The acceleration values for the first and second modes increased 1.17 and 1.5 times, respectively. This is within the 1.61 allowable increase given above. The contribution of the Y-direction earthquake is minor due to the rigidity of the system to the vertical response.

Problem No. 3011: S p+SDL " 'l 3 h = 14950 1.2 S S p+SDL + OBET

= 17656  ;

h  : 17940 S p+SDL l' 3 h  : 26910 DBET A review of Problem No. 1 (River Water System) indicates that only 54.3 percent of the allowable seismic OBE stress was attained using the original amplified response spectrum. Similarly for the DBE case, 46.5 percent of the allowable was used. Therefore, a substantial increase, 1.84 times for OBE candition and 2.15 for the DBE condition, would be acceptable. For the OBE case, a ecmparison of the original ARS and the SSI-ARS indicates a slight increase in acceleration values for the SSI curve. This increase is only for the Y-direction earthquake, which does not contribute heavily to the overall response of the system.

For the DBE case, the comparison of the curves indicated an increase in the X , Y , and Z-direction earthquake acceleration values. However, the 54! 049 3-4 Revision 1

BEAVER VALLEY PO'n*ER STATION, UNIT 1 stress margin (7.15] readily accommodates this increase. Problem No. I does not have any supports.

Problem No. 1: " = 15000 S p+SDL h I

• S p+SDL OBET "

l

• h I

S p+SDL DBET h

The following ARS for the intake structure have not been peak s; read; however, the problems (157, 158) using these curves have been reviewed and the system frequency is well beyond the spread peak.

A review of procedures used for the qualification of Seismic Category I equipment and the potential effect of SSI-ARS indicates that the original plant qualifications basis is conservative and that increased margins of safety would generally result from the use of SSI-ARS. This conclusion is coi# irmed by comparison of the original plant ARS with SSI-ARS and by review of procedures and seismic data used for the original equipment qualification basis.

Procedures used for the qualification of Seis=ic Category I equipment are described in 3VPS FSAR Section 3.2.2. These procedures resulted in qualification programs being i=plemented for balance-of-plant equip =ent.

Mechanical equip =ent was principally qualified by static analysis YS 1 0 0;' Revision 1 3-5

BEAVER VALLEY TuRER STATION, UNIT 1 techniques and instrumentation and electrical equipment by shake table tests.

The original plant ARS was conservatively used for both analytical and test qualification programs. A review of the original plant ARS on a building-by-building and elevation-by-elevation basis indicated that peak resonant responses occurred below 10 H: and that amplification of grcund motion principally occurred below 20 H: for structures housing Seismic Category I equipment. For each building a " cutoff frequency" was selected (i.e.. 10 or 20 Hz) in order to identify seismic acceleration levels above and below the cutoff frequency for calculational purposes. The "g" level identified below the cutoff frequency was a minimum of 1.3 times (Ref.

FSAR Quartion 3.15) the peak ARS response. At the cutoff frequency the rigid range g value was conservatively selected. Equipment having a natural frequency below the cutoff frequency was qualified to an equivalent static acceleration of 1.3 times the peak ARS response. When equipment frequency characteristics were rigid (above the cutoff frequency) the maximum rigid range g values were used. For tested equipment, the maximum rigid range g levels were conservatively used for qualification.

A comparison of the ARS used for the original plant design with the SSI-ARS indicates that the original plant ARS are consersative based upon the 54i 05i O

3-6 Revision i

BEAVER VALLEY PC'JER STATIOd, UNIT 1 above seismic specification of static g values for qualification by static analysis and testing. Seismic Categolf I equipment was qualified on this conservative basis.

Seismic qualification of Seismic Category I equipment may also be established by the response spectrum modal analysis or seismic testing (Test Response Spectra) techniques. For these options, the ARS used for the original plant design provide the appropriate seismic definition for qualification. In this regard it is noted that peaks of the SSI-ARS are significantly lower than the peaks of the original plant ARS. The SSI-ARS peaks occur in the 2 to 5 Hz region for all structures evaluated :nd there is little amplification of maximum floor acceleration above 10 Hz. In some isolated cases the SSI-ARS curves exc eed the original plant ARS in the low frequency region (below 5 Hz) distant from peak original ARS responses. This breaching of the original ARS would only potentially affect equipment whose natural frequency is below 5 Hz. One item, the outside recirculating spray pumps, was found which exhibited natural frequencies below 5 Hz. This component was qualified by dynamic analysis using the original plant ARS. It was concluded to be seismically qualified on the basis of a significant reduction of the primary modes response. Seismic Category I equipment which exhibits natural frequencies in excess of 5 Hz cannot be affected.

541 052 3-7 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 Based on this discussion it is concluded that the ARS used fo'r original plant design provide an acceptable basis for qualification of Seismic Category I equipment.

Components loaded by piping systems are reviewed by analytical techniques described above. Each components nozzle is first reviewed to assure local c omp o".e nt integrity. Loads for all nozzles were combined with the component seismic response to assure adequacy of component supports (near term). All components required for near term have been qualified to their revised loadings. Each component's seismic response was not revised to reflect changes due to SSI consideration. This is extremely conservative e

and facilitated an expeditious review of nozzle load data.

2. Indicate which code or what criteria is used for the evaluation of local stresses and whether anything different from the original analysis is being done in this respect.

Resoonse Local stresses are those induced at welded attachments to pipe, such as lugs or trunnions. Criteria for local stress evaluation are established through application of Velding Research Council Bulletin 107 (b'RC-107).

541 053 3-8 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 This method of a7.alysis is consistent with the original analysis performed.

3. Indicate whether eccentricities, e.g., valve center of gravities, are accounted for in the piping analyses.

pesconse The eccentricity of the operators on 411 motor-operated and air-operated val res is included in the pipe stress analysis / review.

4 If interim operation is proposed, indicate how ISE Bulletin 79-02 vill be addressed p-ior to startup for any support which contains base plates and concrete expansion anchor bolts, which are not found to be completely acceptable.

Essponse Duquesne Light Company has a program underway for inspecting base plata and anchor bolts in the plant. Those supports which at this time are not completely acceptable have been included as priority items for this inspection.

541 054 3-9 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 NRC Letter The following are responses to questions raised in an NRC letter (Appendix C,)

f rom Mr. D.G. Eisenhut to Mr. C.N. Dunn of Duquesne Light Company.

1. Indicate whether both OBE and DBE seismic stresses always include stresses due to seismic anchor movemer.ts (if any) and show how they are combined; e.g., sum of the absoluta values. Is anything being done differently now than was done in the original SHOCK 2 analysis? Your answer should include an explanation of the second paragraph of page B. 2-2 of the FSAR.

Resoonse For the reanalysis effort, the effects of the seismic anchor displacements have been evaluated statically and separately from the inertia effect.

Stat'.c analysis is performed for each direction of relative displacement and for each e.arthquake, leading to a total of six evaluations. Internal moments resulting from the three e'aluations for each earthquake are combined by SRSS on a component level and are then combined with the inertia effects by absolute summation, also on a component level. 1.54.s procedure differs from the SHOCK 2 procedure in that the SHOCK 2 program utilized a single static analysis for each earthquake that incorporated the anchor movements in each of three directions simultaneously with the 541 055 3-10 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 equivalent inertia forens resulting from the intramodal and then the intermodal summatio. procedures of SHOCK 2.

Calculated stresses in Table 4-1 include the effect of ancho.1 displacement combined with inertia effects with the resulting response then combined and deadicad and pressure stresses to form the total stress which is compared to the allovable stress, as follows:

S p+SDL OBET I

• h S,j + SDL + DBET I

• h problem No. 120 (River Water System) has been evaluated for the DBE case as follows:

S 7 r SDL + DBEI I

• h 4c the time the Beaver Valley 1 procedures were formulated, the 331.1 code did noc address seise.ic design in the sense of providing detailed rules f1r stress determination and load combinations. Further, the code did not deal with Normal, Upset, Emergency, and Taulted ~- ess limits. Since that time, development of 331.7 and ASME III have addressed these rules and limits.

541 05C 3-11 Revision 1

BEAVER VALLEY PO5TER STATION, UNIT 1 Current rules allow two significant departures from the original techniques utilized on Beaver Valley Unit 1.

A. An option is provided for Upset Conditions whereby the anchor displacement effect can be considered in equation 9 along with deadweight, pressure, and seismic inertia effects or they may be combined with thermal expansion effects and aealuated under equation 10. -

B. For Emergency and Faulted Conditions, the codes require evaluation of only the primary portion (inertia effect) of the seismic loadings and do not require that the anchor displacement effect be considered, since it is secondary in nature. Also allowed is a Faulted Stress allowable of 2.4 Sh, which was not stated in the Beaver Valley Unit i licensing documents; the equivalent value utilized was 1.8 S

• h

2. State how support stiffness is being accounted for in the current reanalysis effort and whether anything different from the original analysis is being done in this respect.

541 057 3-12 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

Response

Reanalysis efforts are utilizing two programs, SHOCK 3 and NUPIPE. If SHOCK 3 is utilized, supports and restraints are modeled in the manner of SHOCK 2 as rigid members, essentially allowing zero deflection in each restrained direction. When NUPIPE is utilized, representative spring stiffnesses are input in each restrained direction.

Consistent support stiffnesses are used for each problem.

3. Provide the acceptance criteria used in the design of the pipe supports, including veld and bolt sizing criteria, and indicate any deviations from criteria originally used (except criteria established in addressing ISE Bulletin 79-02). Also, state your intention to comply, prior to facility startup, with I&E Bulletin 79-02 for all cases where loading on a pipe support increases as a result of the piping reanalysis and the support reevaluation indicates chat any part of the support is not within the applicable acceptance criteria.

541 058 3-13 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

Response

Acceptance criteria used in the design of pipe supports are shown in Table 3-1. Allowable loads for drilled-in-concrete anchor bolts are shown in Table 3-2. These criteria are being utilized for the reevaluation effort except under the conditions of Section 2 which addresses interim startup conditions.

Duquesne Light Company has a program underway that addresses the fellowing items as a plan of action to comply with IE Bulletin 79-02 for those pipe supports requiring modifications based upon pipe stress analysis described in this report.

a. Where pipe support reanalysis results in new supports, the base plates and anchor bolts shall be designed incorporating IE Bulletin 79-02 criteria.

b. Where pipe support reanalysis results in modifications to existing supports, the base plates and anchor bolts shall be evaluated incorporating IE Bulletin 79-02 criteria.

c. Field inspections shall be perfo'rmed on those existing base plates being modified in order to ensure bolt integrity.

541 059 3-14 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

4. Discuss the impact the current piping stress reanalysis effort has on the FSAR pipe break criteria. Indicate whether postulated pipe break locations could or have change (d) as a result of the reanalyses and, if so, what you propose to do in the event a break location previously not designed for must be postulated.

Resoonse The reanalysis performed to date to the licensed acceptance criteria indicates that stress patterns have not changed significantly since

=aximum stresses occur at points of stress intensification, such as elbows and branch connections.

A detailed review of these problems indicates that the first five highest stress points occur at points of stress intensification. They also occur in those areas where the lines are fully restrained by pipe whip restraints and therefore no additional restraints are required.

FSAR Section 5.2.6.3 states that break locations have been postulated for only the main steam and feedwater inside containment and Appendix D of the FSAR states that breaks need only be postulated in the main steam and feedvater systems outside containment.

54i 060 3-15 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 NRC Letter The following are respontes to questions raised in a second NRC letter, dated May 25, 1979 (Appendix G) also from Mr. D.G. Eisenhut to Mr. C.N. Dunn of Duquesne Light Company.

1. All pipe runs analyzed with SHOCK 2 must be identified.

Reseoqie Appendixes A and 3 identify problems originally analyzed with SHOCK 2.

Appendix A lists those problems addressed for interim startup and Appendix B lists those problems to be analyzed in the long term.

2. Request the following full size drawings:

RM-213 RM-27A, B RM-29A, B,C,D RM-37A RM-39A, 3 -

541 061 3-16 Revision 1

BEAVER VALLEY POWIR STATION, UNIT 1 Resoonse Full size drawings were provided to the NRC by StW during the meeting at S&W on June 5, 1979.

3. Reanalysis of the primary component cooling water heat exchanger discharge piping.

Lines: 18"-WR-14-151 Q3 18"-WR-15-151 Q3 18"-WR-16-151 Q3 30"-WR-17-151 Q3 Failure of any of these lines would result in flooding of redundant safety related equipment.

Resoonse These lines have been added to the problems for interim startup. Problem No. 121 includes:

18"-WR-14-151-Q5 18"-WR-15-151-Q3 541 062 3-17 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 1 '-WR-16-151-Q3 30"-WR-17-151-Q3 Problem No. 122 includes:

30"-WR-17-151-Q3 4 Reanalysis of the following lines located in the intake structure.

30"-WR-171-151-Q3 ,

30"-WR-172-151-Q3 e

30"-WR-175-151-Q3 18"-WR-154-151-Q3 12"-WR-177-151-Q3 10"-SWW-14-151-Q3 10"-SWW-1-121M Failure of any of these lines coul' result in possible flooding of safety related pumps. The asterisked line, unlike the other lines, was not considered safety-related during the plant design and was never seismically analyzed. This line runs above and adjacent to River Water Pump 13 and can only be isolated from the seismically designed piping by a 541 063 3 '.3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 manually operated butterfly valve which is normally open during plant operation.

Response

These lines have been added to the problems for interi= startup. Problem No. 152 includes:

30"-WR-171-151-Q3 30"-WR-172-151-Q3 30"-WR-175-151-Q3 Problem No. 160, which overlaps problem No. 159, includes:

18"-WR-154-151-Q3 Problem No. 161 includes:

12"-WR-177-151-Q3 541 064 3-19 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 Problem No. 165 includes:

10"-SWW-14-151-Q3 10"-SWW-1-121

5. The cooling water discharge lines from the emergency diesel generator cooling system heat exchangers downstream of the normal open isolation valves are not seismically qualified. These lines are located in the diesel generator compartments and their failure could impact on the operation of the emergency diesels. A seismic analysis should be performed on these lines.

Response

The cooling water discharge lines, which are less than 6 inches, were not analyzed on SHOCK 2 but were hand calculated and seismically supported based on standard spacing between supports.

6. The discharge lines of the quench spray pumps have not been proposed for reanalysis.

10"-QS-3-153-Q3 10"-QS-4-153-Q3 41 q/c 4i bUJ 3-20 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 8"-QS-22-153-Q3 8"'QS-23-153-Q3 Justify that reanalysis of the above lines is not necessary.

Resoonse The discharge lines of the quench spray pumps were seismically analyzed on NUPIPE for the DBE plus water hammer loads previous to the present reanalysis effort; consequently, these lines were not included in this reanalysis effort. The OBE case for which the SHOCK 2 run is the calculation of record will be rerun in the long term reanalysis.

7. The recirculation spray piping both inside and outside containment with the exception of the lines listed below is not being reanalyzed. Justify that reanalysis of the recirculation spray system is not necessary.

12"-RS-5-153-Q3 12"-RS-7-153-Q3 12"-RS-8-153-Q3 541 066 3-21 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 Reseense The recirculation spray lines were seismically analyzed on NUPIPE previous to the present reanalysis; consequently, these lines vers not cluded in this reanalysis effort. The 03E case for which the SHOCK 2 run is the calculation of record will be rerun on the long term reanalysis.

8. Verify that the discharge lines from the control Yoom air condition condensers, the charging pumr coolers, and line 6"-WR-53-151-Q3 have been seismically analyzed by an acceptable method. These lines are part of the river water system.

Res onse The discharge lines were not analyzed on SHOCK 2, but were hand calculated and seismically supported based on standard spacing betusen supports.

Additional NRC Ouestions The following questions were raised during a telephone conversation among Duquesne Light Company, Stone & Webster, and NRC personnel on June 28, 1979.

54i 067 3-22 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

1. It appears that Table B-2 contains some problems which should be addressed in the short term.

Resoonse The problems previously included in Table B-2 have been rereviewed in depth and, as a result, the short-tern effort has been revised to include the following: -

Problem 213 Problem 2113 Problem 616 Problem 651 Problem 652 Problem 653 Problem 301 (Comprised of Problems 308, 3007, 3008, 3013 and 3014)

The following problems have been found to be checks of the hand calculations of .ecord and have been transferred to Tabla B-3:

541 068 3-23 Revision i

BEAVER VALLEY POWER STATION, UNIT 1 310 3021 3131 312 3031 341B 3035 655C 3043 840 3100 965 3127 Problem 139 was voided because the line was not r. quired to be seismically supported.

2. For those problems not includeu in the interim scope, what is the consequence of a failure?

Response

The systems which are not included in the interim scope are (1) component cooling water system outside containment, (2) fuel pool purification and cooling system, and (3) quench and recirculation spray system.

The component cooling water (CC) system outside containment has been evaluated using the short-term criteria with the following results:

541 069 3-24 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 Problem 171, the supply to the CC heat exchangers has one support (M-65) out of 15 which has a local overstress of 4 percent.

Problem 270, the discharge from the CC heat exchangers has one support (H-56) out of 13 which has a local overstress in a lug of 78 percent. It is considered that, if a DBE vere to occur, failure of these two supports would not cause a system rupture or a resultant loss of function.

The fuel pool cooling and purification system is presently isolated since there is no srent fuel being stored.

The quench and recirculation spray systems have been completely analyzed for DBE and water hammer loads using NUPIPE. The. OBE case will be run in the long term.

3. How have stress intensification factors been applied at branch connections during the reanalysis?

Response

Appropriate stress intensifications from B31.1 have been applied to the run pipe at reduced outlet branch connections. Branches which are uncoupled have been evaluated for the effects of the movements of the run 541 070 3-25 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 pipe using appropriate stress intensification. The thermal and seismic displacements of the run pipe are applied at the branch with the stresses being determined by the use of a flexibility nomograph. The stresses are then compared to code allowables.

4. In the SSI Report, where do the building displacements come from? Which data sets were used and what are the bases for their selection?

Reseense The building acceleration and displacement profiles, illustrated in Figures 4-11 and 4-12 of the Report on " Soil-Structure Interaction in the Development of Amplified Response Spect:a for Beaver Valley Power Station Unit 1," are maxi =a from the time history responses at each mass point in the structural dynamic model and are determined automatically by the FRIDAY computer program. They are based on the FSAR earthquake, the strain-compatible free-field soil properties from the final iteration of the SHAKE computer program, and a structural damping ratio of 0.02. This is consistent with the basis used for generation of Amplified Response Spectra (ARS) and conservative with respect to soil properties associated with broadened and ' bumped' ARS, referred to under Item 7, Section 9.5 of the report. Displacements calculated on this basis are, therefore, reasonable for use in the reevaluation of piping systems.

5x41 D71 3-26 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

5. Provide a general statement relative to the sele. tion of an amplified response spectrum at the highest support location versus the center of gravity of the piping system.

Reseense Ap p endi: B2.1 (page 32.2) of the FSAR states that Beaver Valley Unit i dynamic piping stress analysis is based on a response- spectra curve closest to, but higher than, che center of gravity of the piping system.

However, the procedure that is being implemented on the reanalysis effort, that is, to use the amplified response spectra at the highest pipe support elevation is always conservative, because the ARS at the highest support

@ location vill always result in higher acceleration levels than at the center of gravity.

For the reanalysis to date, only two problems have used ARS curves which have been applied just above the center of gravity of the piping system.

These two problems are the pressurizer relief valve discharge lines (833) and the pressurizer spray line (1200); both systems encompass a large elevational change from termination to termination. In these two cases, it has been deemed to be more reasonable to use an ARS curve close to the center of gravity of the system, rather than at the highest support location.

Sh\

3-27 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

~

TABLE 3-1 PIPE SUPPORT ACCEPTANCE CRITERIA Load Combination Tension SheaT Column Buckline Welds Maximum of.

DL + TH + 03ET 0.8 Sy 0.513 Sy (web) Note (1) 0.3 Su 2I 0.53 sy DL + TH + DBET Note (1): Colunn buckling criteria are established by Euler equations and are a function of (Idd in accordance with Table 1-36, p 5-84 of AISC.

r I

}k\ 0 1 of 1 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 3-2 DRILLED-IN-CONCRETE ANCHOR BOLT ALLOWABLE LOADS

1. Red head self-drill type S, and type JS installed in 3,000 psi concrete; see Attachment A, Tables I and II, respectively.

For reductions in allowable loads due to closer spacing, see Attachment A, Tables III and IV, respectively.

2. Star slugin compounded cinch anchor bolts and ring vedge cinch anchors; see Attachment A, Tables V and VI, respectively.

3. Hilti or Phillips vedge type anchor bolts are as follows:

Allevable Loads

• Solt Tension Shear Diareter (lbs) (!bs) 3/8" 950 1150 1/2" 2185 2180 5/8" 2145 2845 3/4" 3525 3800 7/8" A&00 4585 1" 5710 6780 The one-third increase does not apply to drilled-in-concrete anchor bolts.

4 Anchor bolt tension and shear interaction equation:

[T M [S \Y3 4

i T ~

- 1. 0 (A (SA)

Where T/TA and S/SA are the ratios of the actual over the allevable for tension and shear, respectively.

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BEAVER VALLEY POWER STATION, UNIT 1 TABLE 3-2 (Cont)

ATTACHMENT A TABLES III & IV 6

PULLOUT CAPACITIES OF PHILLIPS RED HEAD CONCRETE ANCHORS AS AFFECTED BY SPACING ,

In compliance with the request of the client, Doberne is Elgenson conducted a series of tests to develop the information used in this report. The test facilities of the Smith-Emery Company, an independent testing laboratory, were used. .

'Ibe purpose of these tests was to determine the load holding characteristics of Ihillips anchors under various spacing arrangements.

Results

1. When the spacing between adjacent anchors reaches a distance equal to several times the anchor diameter, there is no loss in capacity. The following table shows the minimum center-to-center spacing that could be used with each anchor without causing a loss in individual capacity. -[ Q g g Anchor Bolt Size l1/4" \5/16" l 3/8" 1/2" 5/8" I 3/4" l 7/8" Minimum Spacing 3" 3-1/4" 4" 5" 6" 7" 8" for 100% capacity

2. When the center-to-center spacing, as shown in the above table is reduced, the capacity of the individual anchor decreases.

' Die following table shows center-to-center spacing correspone".ng to a 208/. reduction in individual anchor espacity.

TAfbt.ETI Anchor Bolt. SLze 1/4" 5/16" L 3/8" 1/2" 5/8" 3/4" 7/8" Minimum Spacing 2" 2-1/2" 3" 3-1/2" 4" for 80% Capacity bl/2" 1-5/8" Dimensions of blocks used for tests were 8" x 8" x 16" with an average compressive strength of 2650 psi.

541 076 q

T3//qp He,

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a ce a ,-

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 3-2 (Cont) ATTACHMENT A TABLES V & VI i '"

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BEAVER VALLEY POWER STATION, UNIT 1 0 SECTION 4 PIPE STRESS RESULTS A total of 120 pipe stress problems have been identified for reanalysis and l are being analyzed by Stone & Webster Engineering Corporation in Boston, Massachusetts. -

The pipe strass reanalysis consists of substituting the SHOCK 3 or NUPIPE code for the SHOCK 2 cade. SHOCK 3 is a current seismic code that calculates both intramodal and intermodal seismic forces using a modified square root of the sum of the squares (SRSS) technique and an SRSS technique, respectively, rather than an algebraic summation. The NUPIPE Program utilizes modal response combinations as follows:

Intermodal - SRSS for combination, grouping for modal combination (where closely spaced modes are combined by absolute sum).

Intramodal - SRSS for direction combination.

50 CN 4-1 Revision 1

BE4VER VALLEY POWER STATION, UNIT 1 Field verified piping fabricator isometric drawings provide the basis for program inputs for the pipe stress reanalysis.

Additionally, in some cases, piping is analyzed utilizing amplified response spectra (ARS) that are developed using soil structure interaction techniques (SSI-ARS). The resultant stresses and loads are used to evaluate piping, supports, nozzles, and penetrations. These techniques are discussed in Section 8.7. -

of the ~ 120 SHOCK 2 problems, 93 have been reanalyzed and are within allevable f stress valuss. Table 4-1 lists the problems including the peak stress values for the SHOCK 3 and NUPIPE pipe stress runs.

Stresses were computed by the SHOCK 3 sr NUPIPE program using different mass models and in some cases different ARS than the original calculations. More importantly, the reanalyses were based on field-verified, as-built conditions which in some cases differ significantly from the original design conditions.

For these reasons, the originally calculated stresses are not comparable to the new stresses.

Table 4-2 summarizes the nozzles and penetrations evaluated under the reanalysis program. Of a total of 87 nozzles on problems within the scope of the interim e: fort, 82 have been evaluated and found to be acceptable, and 5 541 079 4-2 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 are contained in problems for which the final pipe stress analysis is not complete but are expected to be acceptable based on reanalysis.

The SHOCK 2 stress problems contained in the interim effort include 50 penetrations, all of which have been evaluated and found to be acceptable.

Summary During the period between the initial issue of this report and this revision, 30 additional problems have been rerun on NUPIPE using the SSI-ARS curve. All of the abovs 30 problems have been reanalyzed and were found to be within allavable stress limits.

541 080 4-3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 PIPE STRESS REEVALUATION

SUMMARY

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method Reactor Coolant 653A 12.547 7.671 NUPIPE/SSI-ARS 18,820 8,189 653B 19.200 10.084 NUPIPE/SSI-ARS 28,800 10,575 653C 19.200 15.060 NUPIPE/SSI-ARS 28,800 17,244 833 & 8 17.220/10.2005 12.420828 NUPIPE/SSI-ARS 25,830/28,200 17,300 1200 19.200 12.690 NUPIPE/SSI-ARS 28,800 16,424 1201 19.200 9.711 NUPIPE/SSI-ARS 28,800 10,442 Safety Iniection 391A 19.080 15.425 SHOCK 3/SSI-ARS 28,620 18,228 2112 22.500 20.754 SHOCK 3 33,750 25,002 610 18.586 11011 NUPIPE/SSI ARS 27,8?8 2,328 613 21.180 9.802 NUPIPE/SSI-ARS 31,770 14,336

61. 19.500/20.280) 17.585 NUPIPE/SSI-ARS 29,250/30,400 16,069 15 17.340/19.200' 7.214 SHOCK 3/SSI-ARS 26,010/28,800 8,123 541 081 1 of 8 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

O System Allevable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method 1011 20.850 8.245 SHOCK 3/ S SI -ARS 31,275 14,087 301 19,200 7.722 NUPIPE/SSI-ARS 26,800 8,712 213 20.388 5.837 NUPIPE/SSI-ARS 30,587 5,437 2113'** 20.388 4.078 NUPIPE/SSI-ARS 30,582 4,443 -

Quench -

Sorav 211 22.500 1.807 SHOCK 3/SSI-ARS 33,750 2,653 212 22.500 10,441 SHOCK 3 9 228 33,750 22.500 11,639 12.149 SHOCK 3 33,750 16,589 229 22.500 11.810 SHOCK 3 33,750 15,987 Recirculation Scrav i

i 612 18.796 1.366 NUPIPE/SSI-ARS 28,193 1,434 Charging and

• Volume Control 100 18,660 15.220 SHOCK 3 27,990 15,468 102 13 660 6,289 SHOCK 3/SSI-ARS 27,990 6,621 007-54t 2 of 8 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (psi) Stress Method Residual Heat Removal 255A 17.940 Lpli NUPIPE/SSI-ARS 26,910 6,041 256 17.160 11.841 NUPIPE/SSI-ARS 25,740 15,063 14 17.940/19.200 8.740 SHOCK 3/SSI-ARS 26,910/28,800 10,376 3011 17.940 17.656 NUPIPE 26,910 18,148 616 18.300 8.498 NUPIPE/SSI-ARS 27,450 11,500 Component Cooling Water 301. 18.000 6.?95 NUPIPE/SSI-ARS 27,000 10,271 303 18.000 6.906 NUPIPE/SSI-ARS 27,000 10,377 304 18.000 _f.836 NUPIPE/SSI-ARS 27,000 11,108 305 18.000 5.835 NUPIPE/SSI-ARS 27,000 8,330

- 306 18.000 4.246 NUPIPE/SSI-ARS 27,000 3,077 307 18.000 6.133 SHOCK 3/SSI-ARS 27,000 7,780 180E'88 18.000 7,55- NUPIPE/SSI-ARS I 27,000 7,370 541 093 3 of f, Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method 181E'8' 18.000 7.516 NUPIPE/SSI-ARS 27,000 15,168 170C'85 18.000 4.339 NUPIPE/SSI-ARS 27,000 5,031 171588 18.000 5.472 NUPIPE/SSI-ARS 27,000 5,667 172 888 18.000 8.319 NUPIPE/SSI-ARS

- 27,000 13,734 173D's> 18.000 5.107 NUPIPE/SSI-ARS 27.000 5,994 174D'** 18.000 3.036 NUPI"E/SSI-ARS 27,000 4,115 175B'8' 18.000 5.197 NUPIPE/SSI-ARS 27,000 5,414 176At*> L8,000 3.505 SHOCK 3/SSI-ARS 27,000 3,777 177'8' 18.000 9.223 iHOCK3/SSI-ATS 27,000 13,707 179Cis* 112222 15.703 NUPIPE/SSI-ARS 27,000 15,797 179485 18.000 2.081 NUPIPE/SSI-ARS 27,000 2,995 183'85 18,000 9.320 nUPIPE/SSI-ARS 27,000 11,336 184*85 18.000 10.133 NUPIPE/SSI-ARS 27,000 11,734 186A'8' 18.000 15.703 NUPIPE/SSI-ARS 27,000 15,797 270A 18,000 15.703 NUPIPE/SSI-ARS 27,000 15,797 54i 084 4 of 8 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method 215 18.000 16.311 SHOCK 3 27,000 26,731 217 18.000 15.751'28 NUPIPE/SSI-ARS 27,000 23,924 930 18.002 15.751 NUPIPE/SSI-ARS 27,000 23,924 931 18.000 15,751 NUPIPE/SSI-ARS 27,000 23,924 214 18.000 14.740 NUPIPE/SSI-ARS 27,000 25,774 River Wate:

1 18.000 10.801 SHOCK 3 .

27,000 13,759 30 18.000 4.830 NUPIPE/SSI-ARS 27,000 7,576 31 18.000 4.830 NUPIPE/SS_-ARS 27,000 7,576 32 18.000 5.363 NUPIPE/SSI-ARS 27,000 8,390 33 18.000 5.169 NUPIPE/SSI-ARS 27,000 8,241 140 18.000 13.758 SHOCK 3/SSI-ARS 27,000 16,349 384 18.000 6.156 SHOCK 3(5) 27,000 8,512 157 18.000 1.884 NUPIPE/SSI-ARS 27,000 2,011 158 18.000 1.976 NUPIPE/SSI-ARS 27,000 2,090 5 of 8 Revision 1

BEAVER VALLEY PO'JER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem *fo. (esi) Stress Method 159'6' 18.000 10.443 NUPIPE/SSI-ARS 27,000 17,277 128 18.000 10.760 NUPIPE/SSI-ARS 27,000 12,562 127 18.000 13.384 NUPIPE/SSI-ARS 27,000 15,970 125 18.000 10.760 NUPIPE/SSI-ARS 27,000 12,562 -

124 18.000 13.384 NUPIPE/SSI-ARS 27,000 15,970 123 18.000 10.861 NUPIPE/SSI-ARS 27,000 17,797 e

120 18.000 8.020 NUPIPE/SSI-ARS 27,000 <r>

126 18.000 13.384 NUPIPE/SSI-ARS 27,000 15,970 216 18.000 6.047 NUPIPE/SSI-ARS 27,000 9,989 203 18.000 2.4a4 NUPIPE/SSI-ARS 27,000 4,189 2031 18.000 8.260 NUPIPE/SSI-ARS 27,000 9,699 152 18.000 4,950 NUPIPS/SSI-ARS 27,000 6,032 121 18.000 6,068 NUPIPE/SSI-ARS 27,000 8,354 122 18.000 8.030 NUPIPE/SSI-AAS 27,000 14,096 165 18.000 4.950 NUPIPE/SSI-ARS 27,000 6,032 6 of 8 541 086 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method 652 18.000 1.231 NUPIPE/SSI-ARS 27,000 1,394 653 18.000 1.495 NUPIPE/SSI-ARS 27,000 1,624 Main Steam 658 22.500 10.248 SHOCK 3/SSI-ARS 33,750 12,025 6590 18.002 9.977 SHOCK 3/SSI-ARS 27,000 11,108 101 18.000 16.917 SEOCK3 27,000 18,277 659 22.500 10.544 SHOCK 3/SSI-ARS 33,750 12,570 660 22.500 1J,121 SHOCK 3/SSI-ARS 33,750 13,304 3063 22.500 12.289 SHOCK 3 33,750 16,481 Feed Water 204 18.000 2.952 SHOCK 3 27,000 3,761 783 19.000 9.361 SHOCK 3/SSI-ARS 27,000 11,624 784 18.000 16.853 SHOCK 3/SSI-ARS 27,000 13,726 785 18,000 14.397 SHOCK 3 27,000 20,232 261 18,000 10.479 SHOCK?/SSI-ARS 27,000 13,585 087 g 541 7 of 8 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-1 (Cont)

System Allowable Reanalysis and Stress Maximum Reanalysis Problem No. (esi) Stress Method Diesel Generator Exhaust 651 12.960 1.20( NUPIPE/SSI-ARS 19,440 1,717 Notes: SSI-ARS : Amplified response spectra developed using soils structure interaction techniques Stresses shown are Operational Basis Earthauake (OPE) Stresses Design Basis Earthquake (DBE) Stresses

<*8 TP304/IP316 allowables

'28 After modification

'** Froblems are no longer within scope of short-term reanalysis effort. See Appendix B.

'98 Problems 213 and 2113 include S *

• DL LP + S OBEI DL+ LP DBEI only.

Being rerun with SSI-ARS.

(b) Problem 159 includes Problems 160 and 161.

878 Evaluated for the DBE case only.

n

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{ k '(

8 of 8 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 TABLE 2 N0ZZLE AND PENETRATION

SUMMARY

No. Acceptable Total No. After Pipe No. Requiring System / of Nozzles / Stress Re- Further Re-Problem No. Penetrations analysis Analysis Reactor Coolant 653A 6/0 6/0 0/0 653B 8/0 8/0 0/0 653C 8/0 8/0 0/0 833 1 8 4/0 4/0 0/0 1200 1/0 1/0 0/0 1201 0/0 0/0 0/0 Safety Iniection 391A 1/0 1/0 0/0 2112 0/0 0/0 0/0 610 2/2 2/2 0/0 613 0/0 0/0 0/0 615 2/3 2/3 0/0 15 1/0 1/0 0/0 1011 0/0 0/0 0/0 301 0/2 0/2 0/0 213 0/0 0/0 0/0 2113 0/0 0/0 0/0 541 089 1 of 6 Revision 1

BEAVER VALLEY POWIR STATION, UNIT 1 TABLE 4-2 (Cont)

No. Acceptable Total No. After Pipe No. Requiring System / of Nozzles / Stress Re- Further Re-Proble9 No. Penetratiens analysis Aralvsis Ouench Sorav 211 1/0 1/0 0/0 212 1/0 1/0 0/0 228 1/0 1/0 0/0 229 1/0 1/0 0/0 Recirculation Sorav 612 2/2 2/2 0/0 Charging 8 '

Volume control I

100 2/0 l 2/0 0/0 102 1/0 1/0 0>0 Residual Heat Removal 255A 6/0 6/0 0/0 256 0/0 0/0 0/0 14 1/0 1/0 0/0 3011 0/0 0/0 0/0 616 0/1 0/1 0/0 Component Cooling Water 302 1/1 1/1 0/0 303 1/1 1/1 0/0 541 090 2 of 6 Revision 1

BEAVER VALLEY PO'JER STATION, UNIT 1 TABLE 4-2 (Cont)

No. Accepta' ale Total No. After Pipe No. Requiring System / of Nozzles / Stress Re- Further Re-Problen No. Penetrations analysis Analysis 304 1/1 1/1 0/0 305 1/1 1/1 0/0 306 0/1 0/1 0/0 307 0/1 0/1 0/0 180E 2/0 2/0 0/0 181E 2/0 2/0 0/0 170C 3/0 1/0 2/0 171 6/0 6/0 0/0 172 0/0 0/0 0/0 173D 0/0 0/0 0/0 174D 0/0 0/0 0/0 1753 0/0 0/0 0/0 176A Oe 0 0/0 0/0 177 1/0 1/0 0/0 178C 1/0 1/0 0/0 179 1/0 1/0 0/0 183 3/0 3/0 0/0 184 2/0 2/0 0/0 186A 0/0 0/0 0/0 270A 3/0 3/0 0/0

/

215 0/4 0/4 0/0 217 0/4 0/4 0/0 5z1 091 3 of 6 Revision 1

SEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-2 (Cont)

No. Acceptable Total No. After Pipe No. Requiring System / of Nozzles / Stress Re- Further Re-Problem No. Penetrations analysis Analvsis 930 0/1 0/1 0/0 931 0/1 0/1 0/0 214 0/1 0/1 0/0 River Water

~

1 4/4 4/4 0/0 30 1/1 1/1 0/0 31 1/1 1/1 0/0 32 1/1 121 0/0 33 1/1 1/1 0/0 -

140 1/0 1/0 0/0 384 1/0 ' 1/0 0/0 157 0/0 0/0 0/0 158 0/0 0/0 0/0 159 3/0 3/0 0/0 128 0/0 0/0 0/0 127 0/0 0/0 0/0 125 0/0 0/0 0/0 124 0/0 0/0 0/0 123 0/4 0/4 0/0 120 0/4 0/4 0/0 126 3/0 3/0 0/0 541 092 216 1/1 1/1 .e0 ,

4 of 6 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 4-2 (Cont)

No. Acceptable Total No. After Pipe No. Requiring System / of Nazzles/ Stress Re- Turther Re-Problem No. Penetrations analysis Analysis 203 3/0 3/0 0/0 203', 0/0 0/0 0/0 152 0/0 0/0 0/0 121'28 3/0 0/0 0/0 122 0/0 0/0 0/0 165 0/0 0/0 0/0 652'2' 1/0 0/0 0/0 653'28 1/0 0/0 0/0 Main Steam 658 1/1 1/1 0/0 6590 000 0/0 0/0 101 0/0 0/0 0/0 659 1/1 1/1 0/0 660 1/1 1/1 0/0 3063 0/0 0/0 0/0 Teed-Fater 204 3/0 3/0 0/0 783 1/1 1/1 0/0 784 1/1 1/1 0/0 785 1/1 1/1 0/0 261 0/0 0/0 0/0 541 093 5 of 6 Revision 1

BEAVER VALLEY PC'JER STATION, UNIT 1 TABLE 4-2 (Cont)

No. Acceptable Total No. 'fter Pipe No. Requiring System / of Nozzles / Stress Re- Further Re-Problem No. Penetratioqs analysis Analysis Diesel Generator Exhaust 651 0/0 0/0 0/0 l NOTES:

Not within the short term reanalysis effort. -

'28 These problems recently added to the interim scope. Results of the reanalysis are not available at this time.

541 094 6 of 6 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 SECTION 5 PIPE SUPPORT RESULTS Table 5-1 summarizes the pipe supports evaluated in the reanalysis program.

There are 696 pipe supports on lines within the interim reanalysis effort; of these, 50 8 have been evaluated and found acceptable and 7 have been modified to be acceptable. A support is considered acceptable if all the load components are lower in magnitude than those for which the support was originally designed. If some load components are greater than the original design load components, the support is reanalyzed using the new loads. Of the total 188 supports requirit.g reanalysis, 68 have been found to be acceptable based on DBEI+DL, 111 have not been accepted at this time. Of the 111 unacceptable supports, 76 have not been evaluated at this time due to their recent addition to the reanalysis effort. There is sufficient analytical information available for the remaining 35 supports to exercise engineering judgment in determining whether the unacceptable condition will become acceptable.

1. The use of ASME III Section NF faulted allowable stress values for structural members <

)

6 h '\

5-1 Revision 1

BEAVER VALLEY p0WER STATION, UNIT 1

2. The use of one time load for snubbers

3. Use of DBEI plus dead load If a support is unacceptable using any of the above approaches, a modification is required. Table 5-2 identifies those supports where acceptance is based on he future use of the options listed above. Hardware modifications and additions are discussed in Section 6. -

With respect to item 3 above, acceptance criteria for pipe support design and analysis are presented in Table 3-1 of this report. Is a casis for interim startup of the Beaver Valley Unit i facility, supports which do not meet these criteria vill be reevaluated using the allavables of ASME III, Subsection NF, Appendix XVII and Appendix F for the design basis earthquake (DBE). The load combinations and a summary of significant allowable stresses to which evaluation will be made under these ASME criteria appear in Table 5-3.

Support designs which are not in accordance with either of these criteria will be suitably modified against the acceptance design criteria of Table 3-1 prior to interim plant operation.

Base plate design criteria and anchor bolt pullout and shear allowable loads are addressed in Section 3. The seismic support loadings which will be 541 096 5-2 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 O

utilized for the NT evaluation vill be the result of either SHOCK 3 or NUPIPE evaluations using SSI-ARS.

Summary The pipe support reanalysis effort which took place between the original issue and Revision 1 of this report includes accepting 97 supports; 68 based on DBEI+DL and 29 based on long-term criteria. Also, one additional modf.fication was necessary for the 14" RH' line off the reactor coolant loop.

541 097 5-3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 5-1 PIPE SUPPORTS

SUMMARY

No. Presently Total No. Acceptable No. Acceptable Modifications System / of Based on for Interim or Additiors F_rJ;Lblem No , sueports Reanalysis ooeration Recuired Reactor Coolant 653A 2 2 0 0 653 16 12 0 0 653C 8 8 0 0 83388 15 15 0 1 1200 18 15 3 0 1201 19 19 0 0 Safety Iniection 391A 11 11 0 0 2112 8 8 0 0 610 2 2 0 0 613 5 5 0 0 615'2' 11 6 5 0 15 11 7 '4 0 1011 19 16 3 0 301'88 56 0 56 0 213'** 16 0 0 0 2113'** 16 0 0 0 Ouench Sprav 541 B98 212 3 3 0 0 1 of 6 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 5-1 (Cont) 9 No. Presently Total No. Acceptable No. Acceptable Modifications System / of Based on for Interim er Additions Eroble9 No. Suoports Peanalysis Operation Recuired 228 0 0 0 0 229 0 0 0 0 Recirculation SoraY 612 0 0 0 0 Charzine Volume -

Control 100 9 9 0 0 102 8 8 0 0 Residual Meat e Removal 255A 3 4 4 0 256 6 6 0 0 14 15 15 0 0 3011 11 10 1 0 616'** 7 0 0 0 l Comoonent Cooling Water 302 23 23 0 0 l 303 23 23 0 0 304 33 31 2 0 305 30 29 1 0 306 11 10 1 0 307 10 10 0 0 l 541 099 2 of 6 Revision 1

BEAVER VALLEY FO'4ER STATION, UNIT 1 TABLE 5-1 (Cont)

No. Presently Total No. Acceptable No. Acceptable Modifications System / of Based en for Interim or Additions Proble9 No. Supports Reanalysis Operation Recuired 180E' 5 5 0 0 181E* 5 4 1 0 170C 17 16 1 0 171 15 11 4 0 172 13 12 1 0 173D 15 14 1 0 174D 20 16 4 0 175B 6 5 1 0 176A 5 5 0 0 177 9 9 0 0 178C 14 10 4 0 179 8 8 0 0 183 9 8 1 0 184 14 11 3 0 186A 6 3 3 0 270A 10 6 4 0 215 8 6 2 0 217 10 10 0 1 l

930 3 3 0 0 931 2 2 0 0 214 5 5 0 0 54i 100 3 of 6 Revision 1

BEAVER VALLEY F0WER STATION, UNIT 1 TABLE 5-1 (Cont)

No. Presently Total Fo. Acceptable No. Acceptable Modifications Syste / of Based on for Interim or Additions Problem No. Supports Reanalysis Ooeration Recuired River Water 1 0 0 0 0 30 2 2 0 0 31 2 2 0 0 32 2 1 1 0 33 2 2 0 0 l 140 2 2 0 0 384 5 5 0 0 l 157 3 3 0 0 158 2 2 0 0 159'88 8 8 0 0 l 128 1 0 1 0 127 10 4 6 0 l 125 12 10 2 0 124 13 12 1 0 123 15 12 0 3 120 11 5 6 0 126 7 7 0 0 216 2 2 0 0 203 16 15 1 0 l 2031 9 9 0 0

. 121 15 0 0 0 l 541 101 4 of 6 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 5-1 (Cont)

No. Presently Total No. Acceptable No. Acceptable Modifications System / of Based on for Interim or Additions proble9 No. Succorts Reanalysis Operation peauired 122 19 0 0 0 165 1 0 0 0 152 8 7 1 0 652 1 0 1 0 653 2 0 2 0 Main Steam 658 6 6 0 0 l 6590 3 3 0 0 101 4 4 0 0 659 2 1 1 0 660 7 7 0 0 3063 0 0 0 0 Feedvater 204 15 15 0 0 783 9 9 0 0 784 6 6 0 0 l

785 3 3 0 0 261 6 6 0 0 Diesel Generator Exhaust 651'"> 2 0 0 0 NOTES:

8 Supports are no longer in scope for interim startup. See Appendix 3.

5 of 6 541 102 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TA3LE 5-1 (Cont) 82' Problem 615 contains nodifications (NPSH) ccheduled for installation during the first refueling outage. Therefore, it will only be analyzed for interim operation.

'88 Problem 159 includes Problems 160 and 161.

Problem recently added to scopa and result not available.

Analyzed based on D3EI+DL only.

541 jg3 6 of 6 Revision 1

i BEAVER VALLEY POWER STATION, UNIT 1 TABLE 5-2 ENGINEERING EVALUATION OF REMAINING SUPPORTS Problem Support No. No. Overstress Condition Resolution SAFETY INJECTIOV SYSTEM 615 A37 Pad to Run Pipe Weld over- Will be Acceptable Based on stressed DBEI+DL R61 Frame Overstressed Will be Acceptable Based on DBEI+DL HSS-211 Member Overstressed Will be Acceptable Based on DBEI+DL HSS-212A Snubber Overloaded Will be Acceptable Based on DBEI+DL HSS-2123 Snubber overloaded Will be Acceptable Based on DBEI+DL -

15 H2 Member / Base Plates / Bolt Will be Acceptable Based on Pullout ~

DBEI+DL H8 Lug to Pipe Weld overstress/ Will be Acceptable Based on Bolt Pullout DBEI+DL H102A Snubber overloaded Will be Acceptable Based on DBEI+DL H102B Snubber Overloaded Will be Acceptable Based on DBEI+DL 1011 R13 Member overstressed Will be Acceptable Based on DBEI+DL R14 Member Overstressed Will be Acceptable Based on DBEI+DL R16 Member Overstressed Will be Acceptable Based on DBEI+DL

_ RESIDUAL HEAT REMCVAL SYSTEM 255 H11 Member overstress Will be Acceptable Based on DBEI+DL H16 Member Overstress Will be Acceptable Based on DBEI+DL H21 Member Overstress Will be Acceptable Based on DBEI+DL H22 Member overstress Will be Acceptable Based on DBEI+DL 3011 H10A.4 Weld Overstressed/ Bolt Will be Acceptable Using Pullout SSI-ARS Curve 541 104 1 of 3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

~

TABLE 5-2 (Cont)

Problem Support No. u__ Overstress Condition Resolution COM*0NENT C00LI'7G UATER SYSTEM 304 R182 Bolt Pullout / Member Will be Acceptable Based on Overstressed DBEI+DL R10D.6 Local Stress Will be Acceptable Based on DBEI+DL 305 R176 Bolt Pullout Will be Acceptable Based on DBEI+DL 306 R264 Local Stress /Trunnien Will be Acceptable Based on l Overstress DBEI+DL -

l 215 R201 Bolt Pullout Will be Acceptable Based on DBEI+DL R203 Member Overstressed Will be Acceptable Based on DBEI+DL The following problems are act required for interim startup:

(~ '

170,171,172,173,174 175,176, 2

177,178,179,180,181,183,184, 186 & 270. Refer to Appendix B.

RIVER UATER SYSTEM 127 H56 Me=ber Overstressed Will be Acceptable Based

. on'DBEI+DL ,

H63 Member Over5 tressed Uill be Acceptable Based on DBEI+DL 125 H57 Bolt Pullout Will be Acceptable Based on DBEI+DL H49 Bolt Pullout Will be Acceptable Based on DBEI+DL

,_ 124 H28A Weld Overstressed Will be Acceptable Based

/n on DBEI+DL 1 .. . 4 D i! i e3 CJ l 2 of 3 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1

~

r TABLE 5-2 (Cont)

Problem Support No. No. Overstress Condition pesolution CSMPONENT COOLING UATER SYSTEM 304 R182 Bolt Pullout / Member Will be Acceptable Based on Overstressed DBEI+DL R10D.6 Local Stress Will be Acceptable Based on DBEI+DL 305 Rl76 Bolt Pullout Will be Acceptable Based on DBEI+DL 306 R264 Local Stress / Trunnion Will be Acceptable Based on  !

Overstress DBEI+bL -

f 215 R201 Bolt Pullout Will be Acceptable Based on DBEI+DL R203 Member Overstressed Will be Acceptable Based on DBEI+DL The following problems are not required for interim startup:

9 170,171,172,173,174,175,176, 177,178,179,180,181,183,184, 186 & 270. Refer to Appendix B.

RIVEP UATER SYSTEM 127 H56 Member Overstressed Will be Acceptable Based

- on'DBEI+DL ,

H63 Member Overstressed Will be Acceptable Based on DBEI+DL __

125 H57 Bolt Pullout Will be Acceptable Based on DBEI+DL H49 Bolt Pulleu. Will be Acceptable Based on DBEI+DL 124 H28A Wald Overstressed Will be Acceptable Based j tj i 106 [

2 of 3 Revision 1

L \VER VALLEY Pok'ER STATION, UNIT 1 Pollowing reanalysis of Problem No. 833, an additional snubber was designed tnd vill be installed to alleviate a pipe overstress occurring under upset (OBE) and faulted (DBE) conditions.

Similarly, an additional snut was designed and will be installed in Problem No. 217 to allesiate a pipe overstress occurring under the same conditions.

Three supports in Problem No. 123 vill be modified, one to make the as-built condition agree with the original design, one to strengthen a marginal original design, and one to alleviate an overstressed weld in the support resulting from seismic uplift forces.

Sim.larly, four supports in Problem No. 653B will be modified, three to make the as-built condition agree with the original design, and one to alleviate an overstressed member in the support resulting from seismic forces.

bN) 6-2 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 9

SECTION 7 LONG TERM REANALYSIS PROGRAM The long term reanalysis program will consist of preparing completely documented calculational packages, utilizing the NUPIPE computer program with amplified response spectra (ARS) based on soil-structure interaction (SSI),

for the problems identified in Appendixes A and B. In addition, the problems associated with the quench and recirculation spray systems will be analyzed for the operating basis earthquake.

ANCHOR MOVEMENT CRITERIA Pipe stress analysis for I? aver Valley Unit I was performed in accordance with the ANSI B31.1 Power Piping Code - 1967. In formulating load combinations to meet paragraphs 102.3.3(a) and (d), seismic anchor displacement effects were included with seismic inertia effects to form total seismic response for the D3E case.

Inclusion of the DBE anchor displacement effects in combination with the DBE inertia effects is not a requirement of current codes, neither ANSI B31.1 oc ASME III NC or ND3600, since the displacement effect is secondary in nature.

541 108 Revision 1 7_t

BEAVER VALLEY POWER STATION, UNIT 1 Under the long term piping analysis criteria established for Beaver Valley Unit 1, anchor displacement effects need not be combined with the inertia effects of the DBE event when evaluating primary stresses in the cystem.

541 109 7-2 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 7.

Piping and supports in general are conservatively designed, even when no dynamic seismic analysis is performed. Fossil-fueled power plants, refineries, and process plants have survived major earthquakes in California, Alaska, Guatemala, and other locations with little or no piping damage. This experience includes earthquakes considerably larger than the DBE for Beaver Valley Unit 1.

The experience with piping performance in earthquakes is reviewed in detail in a report included here as Appendix H.

In addition to the conservatisms listed above, which are inherent in any

, design of nuclear facilittes, there are additional conservatisms specific to the Beaver Valley uns.t.

These conservatisms are not theoretical concepts, but indeed are real and existing margins of safety. To quantify these conservatisms is difficult, but this in no way negates the sound conservative premise on which the reanalysis effort is based. These additional conservatis=s are discussed below.

8.1 STRESS LIMITS The analyses and reanalyses of Seismic Category I piping systems are based upon the conservative stress licit of 1.8S h under the limiting faulted or DBE loading conditions. The present ASME Section III Code specifies the piping stress limit to be 2.4Sh under the Faulted DBE Condition. Only the quench and 4 4 st. l a A 8-3

3EAVER VALLEY POWER STATION, UNIT 1 recirculation spray systems were redesigned and reanalyzed in 1975 for the D31 condition including water hammer loads using 2.4Sh as an allevable. In July 1978, the NUREG/CR-0261 reportw used the limit moment theory to address the Code rules, and it was established that gross plastic deformation may occur when primary stress exceeds 1.5 to 2.0 times the yield strength (5 )7 of piping material, but for stresses below these values, functional capability was maintained.

For 3eaver Valley, Unit 1, the majority of carbon steel piping material is of SA-106 Grade B steel. Using the lover limit of 1.5S y from NUREG/CR-0261 and representative properties of SA-106 Grade 3 steel, the added margin of conservatism is the ratio (1.5S /1.8S y h ), which ranges from 1.4 at 650'T to 1.94 at 100*T.

The Beaver Valley Unit 1 pipe stress reanalysis calculations have included the seismic stress due to anchor displacements in the D3E condition. Inclusion of the anchor movement stresses was not explicitly required by ANSI Code 331.1, used for the original design, and is not required by current 1979 codes, for the faulted D3E condition. Addition of this stress component is a significant conservatism for the long term reanalysis.

M E.C. Rodabaugh and S.E. Moore, " Evaluation of the Plastic Characteristics of Piping Products in Relation to Code Criteria," NURIG/CR-0261, July 1978 8-4 54l i!!

BEAVER VALLEY POWER STATION, UNIT 1 8.4 TIELD VERITICATION OT AS-3UILT CONDITIONS The documentation of as-built conditions for Beaver Valley Unit 1 began in September 1974 and was completed prior to startup. The effort was canned by Pipe Stress Analysis and Pipe Support Designers uno walked all Category I piping systems to ensure compliance with the stress analysis summaries (MSKs).

All Category I piping was checked for piping configuration, pipe support location, and pipe support type. The results of this effort were documented and reported en Southwest Fabricating iso =etric drawings which then became part of the permanent plant record. These isometric drawings supersede the RP series drawings. Duquesne Light Co. persennel verified the accuracy of a portion of these isometric drawings during March and April of 1979 subsequent to the shutdown order.

8.5 ENGINEERING ASSURANCE A comprehensivs and extensive Engineering Assurance program has ' en developed and applied to the reanalysis activities. A detailed project procedure was developed chat includes provisions for design control, documant control, and interface controls. Each new proje-t procedure developed received a full review and approval by the S&W Engineering Assurance (EA) staff.

bi 8-6 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 APPENDIX A SYSTEMS AFFECTED FOR IN~ERIM STARTUP The reanalysis for interim startup includes those lines originally computer analyzed with tLe SHCCK2 code and which are necessary for safe shutdown. In order to evaluate safety system lines which have interconnecting ties with ncnsafety system lines, the reanalysis was extended to include linas attached to the safety systems, past the first automatic trip valve or the first normally closed manual valve, to the first piping anchor.

Problem FSAR System Line Number No. Fig. No.

Reactor Coolant 29"-RC-4-250lR-Q1 653A, 4-1 (RC) 27.5"-RC-6-250lR-Q1 31"-RC-5-250lR-Q1 8"-RC-29-250lR-Q1 8"-RC-27-250lR-QL 31"-RC-8-250lR-Q1 653C 4-1 29"-RC-7-250lR-Q1 27.5"-RC-9-250lR-QL 8"-RC-37-250lR-QL 8"-RC-39-250lR-QL 14"-RC-86-250lR-Q1 4-1,4-2 h

29"-RC-1-250lR-Q1 6533 4-1 27.5"-RC-3-250lR-Q1 31"-RC-2-250lR-Q1 8"-RC-17-250lR-Q1 8"-RC-19-250lR-Q1 12"-RC-111-602 833 & 4-2 6"-RC-100-602 8 6"-RC-101-602 6"-RC-102-602 6"-RC-108-602 6"-RC-104-1502-QL 6"-RC-97-1502-Q1 6"-RC-98-1502-Q1 6"-RC-99-1502-Q1 3"-RC-105-1502-QL 3"-RC-106-1502-Q1 3"-RC-107-1502-QL 4"-RC-71-1502-Q1 1200 4-1,4-2 4"-RC-72-1502-Q1 4"-RC-71-1502-Q1 1201 4-1,4-2 4"-RC-72-1502-Q1 hb A-1 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 Problem TSAR System Line Number NO- Fiz. No.

Safety Injection 6"-SI-41-153W-Q2 100 9.1-1 (SI) 6"-SI-42-153W-Q2 6"-SI-40-153W-Q2 12"-SI-110-602-Q1 653C 6.3-2 12"-SI-ill-1502-Q2 4-1, 6.3-2 6"-SI-40-153W-Q2 102 9.1-1 6"-SI-44-153W-Q2 8"-SI-2-153W-Q2 8"-SI-2-153W-Q3 2112 6.3-1 8"-SI-2-153W-Q3 1011 6.3-1, 9.1-1 12"-SI-5-153A-Q2 610 6.3-1, 12"-SI-6-153A-Q2 6.3-2 12"-SI-7-153A-Q2 12"-SI-8-153A-Q2 12"-SI-13-153A-Q2 12"-SI-6-153A-Q2 613 6.3-1, 12"-SI-1-153W-Q2 6.3-2 10"-SI-15-1502-Q1 615 10"-SI-16-153W-Q2 10"-SI-17-153W-Q2 10"-SI-18-1502-Q1 10"-SI-26-153W-Q2 10"-SI-27-153W-Q2 10"-SI-28-1502-QL 6"-SI-32-1502-QL 6"-SI-33-1502-Q1 .

6"-SI-34-1502-QL 615 6.3-1, 6"-SI-40-153W-Q2 6.3-2 6"-SI-44-153W-Q2 12"-SI-121-1502-QL 15 6.3-2,4-1 12"-SI-108-602-Q2 12"-SI-1-153W-Q3 391A 6.3-1, 6.4-1 8"-SI-2-153W-Q3 6.3-1 12"-SI-101-1502-Q1 14 6.3-2,4-1 10"-RH-23-1502-QL 9.3-1 12"-SI-120-602-Q2 10"-RH-16-602-Q2 6"-SI-30-1502-QL 30L 6.3-2 l A-2 .

}fi ,yeyision1

BEAVER VALLEY F0WER STATION, UNIT 1 Proble TSAR No. _ Eie. No t Line Number

_ System 6"-SI-29-1502-Q1 6"-SI-20-1502-QL 6"-SI-19-1502-Q1 6"-SI-32-1502-Q1 6"-SI-33-1502-Q1 2113 6.3-1 6"-SI-40-153W-Q2 213 6.3-1 6"-SI-44-153W-Q2 211 6.4-1 12"-QS-2-1533-Q3 Quench Spray (QF) 212 6.4-1 12"-QS-1-1533-Q3 228 6.4-1 12"-QS-1-1533-Q3 229 6.4-1 12"-QS-2-1533-Q3 612 6.4-1 Recirculation Spray 12"-RS-7-153A-Q2

' 12"-RS-8-153A-Q2 (RS) 12"-RS-5-153A-Q2 100 9.1-1 Charging and Volume 6"-CH-63-153W-Q2 6"-CH-67-153W-Q2 Control 8"-CH-15-153W-Q2 (CH) 102 6"-CH-15-153W-Q2 6"-CH-68-153W-Q2 255A 9.3-1 12"-RH-6-602-Q2 Residual Heat Re= oval 12"-RH-9-602-Q2 (RH) 12"-RH-12-602-Q2 10"-RH-4-602-Q2 10"-RH-5-602-Q2 10"-RH-7-602-Q2 10"-RH-8-602-Q2 10"-RK-10-602-Q2 10"-RH-19-602-Q2 256 9.3-1 12"-RH-9-602-Q2 12"-RH-12-602-Q2 10"-RH-L6-602-Q2 10"-RH-17-602-Q2 6"-RM-20-602-Q2 3"-RM-13-602-Q2 3011 9.3-1 10"-RH-16-602-Q2 541 ,

' Revision A-3

BEAVER VALLEY PC'JER STATION, UNIT 1 Problem TSAR

_ No. Ti?. No.

Systen Line Number 6533 4-1,9.3-1 14"-RM-1-1502-Q1 14"-RH-2-602-Q2 14"-RH-18-602-Q2 653C 6.3-2 10"-RM-24-1502-Q1 14 9.3-1, 10"-RM-23-1502-Q2 6.3-2 6"-RH-14-152-Q2 616 9.3-1 l 302 9.4-4 Component Cooling 18"-CC-118-151-Q3 18"-CC-116-151-Q3 303 9.4-4 304 9.4-4 18"-CC-114-151-Q3 18"-CC-130-151-Q3 305 9.4-4 8"-CC-255-151-Q3 306 9.4-3 8"-CC-256-151-Q3 8"-CC-257-151-Q3 6"-CC-261-151-Q3 8"-CC-476-151-Q3 ,

6"-CC-258-151-Q3 307 9.4-3 6"-CC-265-151-Q3 8"-CC-259-151-Q3 8"-CC-260-151-Q3 8"-CC-517-151-Q3 215 9.4-4 6"-CC-519-151-Q3 24"-CC-125-151-Q3 18"-CC-489-151-Q2 18"-CC-490-151-Q2 18"-CC-529-151-Q3 18"-CC-530-151-Q3 6"-CC-488-151-Q2 6"-CC-526-151-Q3 4"-CC-487-151-Q3 4"-CC-525-151-Q2 3"-CC-486-151-Q3 3"-CC-523-151-Q2 2"-CC-485-151-Q2 2"-CC-524-151-Q3 24"-CC-266-151-Q3 6"-CC-518-151-Q3 217 9.4-3, 24"-CC-ll2-151-Q3 9.4-4 24"-CC-113-151-Q3 k llf A-4 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 Problem TSAR System Line Number l No. Fie. No.

6"-CC-510-151-Q3 9.4-3, 6"-CC-511-151-Q3 9.4-4 6"-CC-512 *51-Q3 6"-CC-482-151-Q3 18"-CC-483-151-Q3 1C"-CC-484-151-Q3 18"-CC-527-151-Q3 18"-CC-528-151-Q3 8"-CC-517-151-Q2 214 9.4-3 6"-CC-481-151-Q2 930 9.4-4 6"-CC-511-151-Q3 6"-CC-480-151-Q2 931 9.4-4 6"-CC-510-151-Q3 Chilled Water 8"-CW-8-151 216 9.4-3 (CW) 8"-CW-9-151 214 9.4-3 e

River; Water 6"-WR-il7-151-Q3 203 10.3-5 (WR) 14"-WR-64-151-Q2 30 9.9-1A 14"-WR-82-151-Q2 31 9.9-1A 14"-WR-89-151-Q2 32 9.9-1A 14"-WR-87-151-Q2 33 9.9-1A 8"-WR-228-1,4-Q3 140 9 . 9 - 1.

8"-WR-229-151-Q3 8"-WR-230-151 8"-WR-231-151 8"-WR-234-151-Q3 214 9.4-3 14"-WR-63-151-Q2 1 9.9-1A 14"-WR-65-151-Q2 14"-WR-86-151-Q2 14"-WR-88-151-Q2 A-5 f ffR,evision1

SEAVER VALLEY FORER STATION, UNIT 1 Froblem FSAR Line Number No. Fiz. No.

System 14"-WR-25-151-Q3 12C 9.9-1A 14"-WR-26-151-Q3 14"-WR-27-151-Q3 14"-WR-28-151-Q3 24"-WR-29-151-Q3 14"-WR-21-151-Q3 123 9.9-1A 14"-WR-22-151-Q3 14"-WR-23-151-Q3 14"-WR-24-151-Q3 24"-WR-19-151-Q3 24"-WR-20-151-Q3 24"-WR-19-151-Q3 124 9.9-1A 24"-WR-187-151-Q3 24"-WR-20-151-Q3 125 9.9-1A 24"-WR-186-151-Q3 24"-WR-7-151-Q3 126 9.9-1A 24"-WR-8-151-Q3 24"-WR-9-151-Q3 18"-WR-11-151-Q3 18"-WR-12-151-Q3 18"-WR-13-151-Q3 24"-WR-19-151-Q3 127 9.9-1A 24"-WR-20-151-Q3 128 9.9-1A 24"-WR-99-151-Q3 157 9.9-1A 158 9.9-1A,3 24"-WR-100-151-Q3 159 9.9-1A,3 20"-WR-1-151-Q3 20"-WR-2-151-Q3 20"-WR-3-151-Q3 20"-WR-4-151-Q3 20"-WR-5-151-Q3 20"-WR-6-151-Q3 24"-WR-99-151-Q3 24"-WR-100-151-Q3 18"-WR-154-151-Q3 12"-WR-177-151-Q3 216 9.4-3, 8"-WR-227-151 9.9-lA 30"-WR-171-151-Q3 9 9-13 5 457 ,

ll8

• A-6 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 Problem PSAR Eine Number No. Fir. No.

System 30"-WR-172-151-Q3 30"-WR-175-151-Q3 10"-SWW-14-151-Q3 165(2' 9.9-13 10"-SWW-1-121 18"-WR-14-151-Q3 121 9.9-1A 18"-WR-15-151-Q3 18"-WR-16-151-Q3 30"-WR-17-151-Q3 122 9.9-1A 6"-WR-155-151-Q3 384 9.9-13 6"-WR-214-151-Q3 652 RM-53A 6"-WR-215-151-Q3 653 RM-53A Main Steam 3"-SDHV-1-601-Q2 101 10.3-1 (MS) 3"-SDEV-2-601-Q2 3"-SDMV-3-601-Q2 4"-SDRV-4-601-Q2 32"-SHP-56-601-Q2 658 10.3-1 32"-SHP-57-601-Q2 659 10.3-1 32"-SHP-58-601-Q2 660 10.3-1 4"-SHP-19-601-Q2 6590 10.3-1 4"-SHP-20-601-Q2 4"-SMP-21-601-Q2 6"-SAE-1-601 6"-SAE-2-601 6"-SAE-3-601 32"-SHP-56-601-Q2 3063 10.3-1 32"-SMP-57-601-Q2 32"-SHP-58-601-Q2 32"-SHP-22-601-Q2 32"-SHP-23-601-Q2 32"-SHP-24-601-Q2 10"-SSVD #.01 10"-SSVD-2-60.

10"-SSVD-3-601 10"-SSVD-4-601 10"-SSVD-5-601 10"-SSVD-6-601 10"-SSVD-7-601 10"-SSVD-8-601 541 119 A-7 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 t Problem FSAR Svete9 Line Number No. Fig. No.

10"-SSVD-9-601 10"-SSVD-10-601 10"-SSVD-il-601 10"-SSVD-12-601 10"-SSVD-13-601 10"-SSVD-14-601 10"-SSVD-15-601 Main and Auxiliary 4"-WAPD-3-601-Q3 204 10.3-5 Feedwater (FW) 4"-WAPD-4-601-Q3 4"-WAPD-5-601-Q3 4"-WAPD-6-601-Q3 6"-WAPD-1-601-Q3 6"-WAPD-2-601-Q3 16 -WFP D-2 2-6 01-Q 2 783 10.3-5 16"-WTPD-24-601-Q2 784 10.3-5 16"-WTPD-23-601-Q2 785 10.3-5 16"-WFPD-9-601-Q2 0261 10.3-5 16"-WFPD-13-601-Q2 9 16"-WFPD-17-601-Q2 6"-WD-23-151-Q3 203 10.3-5 6"-WD-24-151-Q3 6"-WD-25-151-Q3 6"-WD-26-151-Q3 4"-WD-27-151-Q3 4"-WD-41-151-Q3 8"-WD-22-151-Q3 2031 10.3-5 6"-WD-25-15L-Q3 6"-WD-26-151-Q3 Diesel Generator 22"-OL-55-151-Q3 651 RM-53A Exhaust (OL) i NOTES:

Problems 160 and 161 are included within the scope of the reanalysis effort for problem 159.

'2' Problem 165 has been analyzed o n- NUPIPE as part of the Beaver Valley Unit 2 stress analysis effort.

These lines are identified on the flow diagrams included in Appendix C.

A-8 I i 2 0 "'"** * "

BEAVER VALLEY POWT.,R STATION, UNT.T 1 In addition to the problems referenced above, a number of other computer analyses were alco performed for Beaver Valley - Unit 1, using the SHCCK2 code. These haca been excluded from the scope of the reanalysis for interim startup and are discussed in Appendix 3.

r .s ,

J 'i l l.

A-9 Revision 1

3EAVER VALLEY FCWER STATION, UNIT 1 APPENDIX 3 FROBLEMS TO BE REANALYZED IN THE LONG TERM 4

The probicms described in Tables 3-1, 3-2, and 3-3 are within the scope of the long *.erm effort. These problems are identified on the flow diagrams included in Appendix D.

1. Prinary Cc9 enent Cooline Water System The primary compontnt cooling water (CC) system is used during normal operation and coolds vn to remove heat from various pri=ary plant components; however, safe shutdown of the reactor (i.e., hot standby) can be achieved without dependence on the CC system. By use of other systems the plant can be maintained in this condition indefinitely while restoration of the CC system is being accouplished.

The responses to the FSAR questions listed in the referene.es below discuss in detail the effects of various CC system pipe breaks and their impacts on the operation of the plant. These discussione de:enstrate the capability of the plant to maintain a safe shutdown condition without the availability of the CC system. The response to TSAR Question 9.33 541 1,,

t . . . .

3-1 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 describes the method of repair and the expected repair times for cracks As mentioned in that response, the and breaks of various severities.

worst case repair (the addition of a piece of pipe) can be made within 5 days.

In summary, problems a.dociated with the CC system outside the containment system is are included in the long-term reanalysis effort because the CC not required either t. attain or maintain a safe shutdown condition.

Table B-1 identifies the CC problems to be addressed in the long term.

References:

TSAR Section 9.4.

TSAR Questions 9.2, 9.10, 9.11, 9.13, 9.33, 9.34, and 9.35.

2. Other Safety Svstems Table B-2 identifies the SHOCK 2 problems that are within the scope of the are not required for safe long-term reanalysis effort; thesa lines shutdown.

3. Hand Calculations Table B-3 identifies SHOCK 2 problems that are not within the scope of the interim startup or long-term reanalysis effort; these SHOCK 2 runs are only 541 7'73

- - Revision 1 3-2

BEAVER VALLEY P;WER STATION, UNIT 1 check calculations of manual hand calculations. They are identified here only to shov the scope of the original SHOCK 2 effort.

4. Superseded Calculations The following SHOCK 2 runs have been superseded by a problem presently within the interim and long term reanalysis effort.

Superseded New Problem SHOCK 2 Run Number 122A 122 312 840 657 785 916 217 1012 391 7110 341B 6230 310 .

5. Seismically Sueoorted Non-0 Lines Thu following lines are not safety related but have been seismically supported as designated by an "E" in the line desi; nation table.

3-3 Revision L 541 174

BEAVER VALLEY PC?iER STATION, UNIT 1 2"-CV-1-154 2"-SHPD-5-601 2"-SRPD-6-601 2"-SEPD-7-601 2"-SKPD-8-601 1/4-SS-163-N9 1/4-SS-173-N9 ,

1/4-SS-174-N9 B-4 54l 775 m

aevaan 1

BEAVER VALLEY POVER STATION, UNIT 1 TABLE B-2 SAFETY SYSTEMS TO 3E ANALYZED IN THE LONG TERM Problem FSAR System Line Nun'oer ,,

No. Fi2. No.

k Fuel Pool Cooling 6"-FC-4-152-Q3 104

& Purification System 6"-FC-5-152-Q3 (FC) 6"-FC-8-152-Q3 105E 9.5-1 6"-FC-9-152-Q3 10"-FC-1-152-Q3 198B 9.5-1 6"-FC-2-152-Q3 6"-FC-31-152-Q3 l 4"-TC-10-152 *07

. 9.5-1 4"-FC-11-152 6"-FC-14-152 6"-FC-17-152 6"-FC-32-152 I

Quench Spray 10"-QS-4-1533-Q3 614 6.4-1 (QS) 10"-QS-3-1533-Q3 617 6.4-1 4"-QS-6-1533-Q3 210 6.4-1 10"-QS-4-1533-Q3 ,

4"-QS-5-153B-Q3 218 6.4-1 10"-QS-3-153B-Q3 I

Recirculation Spray 4"-RS-14-1533-Q2 611 6.4-1 (RS) 10"-RS-10-153B-Q2 4"-RS-15-153B-Q2 l

River Water 30"-WR-175-151-Q3 153 9.9-1B (WR) l O

541 126 _. . . _ _

1 of 1 Revision 1

3EAVER VALLEY POWER STATION, UNIT 1 TABLE B-3 HAND CALCULATIONS CHECKED 3Y SHOCK 2 Problem Systen Line Number No.

High Pressure Steam 3"-SHP-26-601-Q2 3043 (SHP) 3"-SHP-31-601-Q2 Steam Generator Auxiliary 3"-WAPD-13-601-Q3 207 Feedvater Pu=p Discharge (WAPD) 3"-WAPD-11-601-Q3 208 Generator Water Blowdown 3"-WGCB-8-601-Q2 309, (WGC3) 3"-WGCB-12-601-Q2 3017, 6220, 3002, 3018, 6216 3"-WGCB-4-601-Q2 310 3"-WOC 3-4-601-Q2 3100 Fuel Fool Cooling and 6"-TC-12-152-Q2 301 Purification System (FC) 6"-FC-17-152-Q2 655C Charging and Volu=e 3"-CH-125-1503-Q2 911, Control System (CH) 260, 3001 2"-CH-97-1502-Q1 200 2"-CH-141-1503-QL 220 2"-CH-100-1502-Q2 230 2"-CH-186-152-Q2 2"-CH-1-1502-QL 240 2"-CH-96-1502-Q1 250 2"-CH-23-1502-Q1 300 2"-CH-143-1502-Q1 350 2"-CH-149-1502-Q1 2"-CH-145-602-QL 2"-CH-2-602-Q1 2"-CH-3-602-Q1

= . .

9 ,

' < i u. /

1 of 4 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE 3-3 (Cont)

Problem System Line Number No.

2"-CH-4-602-Q1 2"-CH-146-152-Q3 3/4"-CH-115-1502-Q2 380 2"-CH-2-602-Q2 702

- 2"-CH-148-602-Q2 703 3"-CH-106-153W-Q2 901, 3135 3"-CH-107-153W-Q2 3135 3"-CH-108-153W-Q2 704, 3135 3"-CH-110-153W-Q2 704, 3"-CH-ill-153W-Q2 3057 3"-CH-il4-152W-Q2 3129, 3044 4"-CH-14-153W-Q2 3057 3"-CH-6-153W-Q2 3122 3"-CH-226-153W-Q2 3"-CH-13-153W-Q2 3125 4"-CH-72-1503-Q2 3131 4"-CH-76-1503-Q2 3"-CH-71-1503-Q2 3"-CH-75-1503-Q2 3"-CH-80-1503-Q2 3"-CH-69-1503-Q2 3031 3"-CH-70-1503-Q2 3"-CH-73-1503-Q2 3"-CH-74-1503-Q2 4"-CH-7 2 -1503-Q 2 4"-CH-76-1503-Q2 3"-CH-12 6-150 2-Q1 3035 3"-SI-81-1503-Q1&Q2 900, Safety Injection (SI) 3004 3"-SI-140-1503-QL 902, 9 r3004 J 'i I i_

2 of 4 Revision 1

BEAVER VALLEY POWER STATION, UNIT 1 TABLE B-3 (Cont)

Problem System Line Number No.

6"-SI-J4-1502-Q1 3006 6"-SI-74-1502-Q1 3"-SI-60-1503-Q2 3124 3"-SI-57-1503-Q1&Q2 900 3"-SI-130-1503-Q1&Q2 313, 902 3"-SI-134-1503-Q2 922 3"-SI-81-1503-Q2 3120 3"-SI-56-1503-Q3 3052 3"-SI-60-1503-Q3 3"-SI-133-1503-Q3 4"-SI-75-1503-Q3 3"-SI-134-1503-Q1 3"-SI-31-153W-Q2 3127 3"-SI-145-153W-Q2 3"-SI-35-152-Q3 965 Residual Heat Removal 6"-RH-14-152-Q2 3012 System (RH)

Reactor Coolant (RC) 3"-RC-13-1502-Q1 6530 3"-RC-23-1502-Q1 3"-RC-33-1502-Q1 3"-RC-160-153W-Q2 2"-RC-54-1502-Q1 220 3"-RC-160-153W 917 4"-RC-112-152-Q3 360 3"-RC-160-153W 917 3"-RC-160-153W-Q2 3021 Component Cooling (CC) 6"-CC-512-151-Q3 914 4"-CC-487-151-Q1 910 4 3/ t ! 12g7 I

3 of 4 Revision '

BEAVER VALLEY POWER STATION, UNIT 1 TABLE B-3 (Cont)

Problem System Line Nueber No.

4"-CC-525-151-Q3 3"-CC-235-151-Q3 921 3"-CC-466-151-Q2 3"-CC-523-151-Q3 Diesel Generator oil Line 3"-OL-46-151-Q3 650 (OL)

Primary Grade Water 3"-PG-5-152 917 (PG)

Quench Spray 2"-QS-29-152 315X (QS) 6"-QS-30-153B-Q3 840 6"-QS-31-153B-Q3 6"-QS-16-152 139 8 4"-QS-8-152 341B Neutron Shield Tank 6"-NSL-2-152-Q3 312 Cooling (NSL) 541 130 4 of 4 Revision 1

BEAVER VALLEY PCWER STATION, UNIT 1 LIST OF TIGURES Fieura Title 4-1 Main Coolant System Sh. 1 4-2 Main Coolant System Sh. 2 6.3-1 Safety Injection System Sh.1 6.3-2 Safety Injection System Sh. 2 6.4-1 Contain=ent Depressurization System 9.1-1 Charging and Volume Control System Sh. 1 9.3-1 Residual Heat Re= oval System 9.4-3 Component Cooling Water System Sh. 3 9.4-4 Component Cooling Water System Sh. 4 9.9-1A River Water System e

9.9-1B Intake Structure 10.3-1 Main Steam System 10.3-5 Teedvater System RM-53A Emergency Diesel Generator Fuel and Air System 54i i3l C-1 Revision 1

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BEAVER VALLEY F0WER STATION, UNIT 1 plant are subject to regulatory body approval, so this combination of requirements gover ,d seismic design of 331.1 piping on nuclear plants. As discussed previously, in all except the very early plants, a seismic ground motion be in the form of ground spectra and appropriate acceleration levels would specified. This motion would be applied to the buildings and amplifications of the ground motion at various levels throughout the buildings would be computed in the form of floor response spectra. It is the latter that were used as design bases for nuclear piping. The qualification of large piping systems of safety class categories is nearly always done by means of a computer analysis. A dynamic analytical model of the piping system is derived in which the mass of the system is concentrated at a finite number of mass points and the flexibility of the system is represented by springs connerting the masses. System damping is included as viscous damping, normally with highly conservative numerical values of 0.5 or 1 percent of critical J.iping. The completed model is then analyred for tha appropriate seismic spectral motion on the computer. Usually, one amplified floor response spectrum is used as an input acceleration at every point of support or connection to the building. simplification This can be an important conservatism especially for piping systems traversing piping different vertical levels or differ __. buildings. The model of the system is passed through the computer several times to account for all directions of motion and both the operating and desits basis earthquakes. Inertia. forces are developed first vibration, then the contributiens of each modefor all directions within each mode of total force. are combined to obtain the A current controversy lies in the fact that force combinacions within each mode were in some cases combined algebraically so that some loads would subtract from the total. such a way that subtraction could not occur, which isThe the alternative would be to combi approach is used. case if an SRSS Effects of the inertial forces building displacements, gravity are combined (veight) with effects from relative effects, and internal / external pressure loadings on the pipe. When load combinations are complete, bending moments and stresses in the piping system are computed according to B31.1 equations. Basically, twice the maximum limited to shearing 1.2 S stress in the pipe due to bending and tension is computed and h for the 03E and 1.8 Sh for the D3E in a manner very comparable to ASME III today. S as provided by the Code, in the hot h is the tabulated value of allevable stress condition. In 331.1, Sh is based on the lover of 5/8 Yield Strength except certain austenitic materials or 1/4 Ultimate Strength at operating temperature, temperatures are permitted S values up h at operating toughness and ductility to 90 of percent of yield strength because of the greater these materials. are the These values of alicuable stress lowest in use for any piping in the United States. nuclear piping has higher allevables, as does 331.3 RefineryASME III Class 1 and Chemical Plant Piping. 331.4 and 331.8 for Gas and 011 Transmission piping respectively permit allowable stresses up to 72 percent of the ultimate H-6 541 147

BEAVER VALLEY POWER STATION, UNIT 1 deflection. Extrapolating the curve of Figure H-3 to 0.5 inch deflection yields 10 percent damping. As plasticity develops in the piping even in small amounts, damping ratios of 10 percent and higher are definitely to be expected. In fact, there is a major project underway at the present'r to develop seismic restraints based on cyclic plasticity of the supports. The essential quality of the relationship between damping, acceleration level, and damage is that damage to piping does not increase proportionately with input acceleration levels and this is due in large part to increases in damping levels as deflections increase.

8. CONCLUSIONS AND IMPLICATIONS FOR MODERN NUCLEAR PLANTS The evolution of seismic design methods in nuclear power plants has been reviewed together with the development of the piping codes. It was shown that nuclear plants that meet the older 331.1 code will more than likely also satisfy the new nuclear codes that have better quantified conservatism.

Available data on the actual seismic performance of power piping systems were reviewed. It was shown that operating power plants do indeed have very high levels of seismic capability. Of the several plants that sustained severe ground motion from 0.2 to 0.6 g, there were no failures of velded stael power piping. Considering the magnitudes of the earthquakes and the variability of the design practices, this is an excellent record and can only have been made possible by the natural resillsney of power piping. The probable reasons for this natural resiliency were discussed next. It is believed that the main reasons are: first, the substantial conservatism of the Code for Power Piping, 331.1, including the provisions for materials, fabrication, and construction; second, that design of piping for thermal expansion provides inherent seismic capability; and third, that damping increases very rapidly with deflection levels. The large damping factors prevent buildup of seis=ic disturbances in resonant systems. It is believed these reasons explain the remarkable performance of piping systems in earthquakes. Based upon the foregoing observations, it is very improbable that piping-related safety problems would occur in nuclear plants in the eastern United States due to seismic disturbances. These plants have maximum ground motions of 0.15 g; they have been designed by dynamic analysis; and all safety piping systems have been specifically scrutinized. Contrast this situation with say the Kern County plant where 0.25 g was actually experienced and explicit analysis was perforned o _f on the steam and feed lines; or the ENALUP plant which was probably designed statically and experienced perhaps 0.6 g. The contrast is simply too great; piping failures of nuclear safety systems should not result from earthquakes in the United States.

9. REPERENCES

1. Cloud, R.L. et al. Editors, Pressure Vessels and Piping, Design and Analysis, Amer. Soc. of Mech. Engrs. N.Y., N.Y., 1972.

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BEAVER VALLEY POWER STATI0h UNIT 1

2. Biot, M.A., Analytical and Experimental Methods in Engineering Seismology.

Trans ASCE 108 p 365-408, 1942.

3. Biggs, J.M., Introduction to Structural Dynamics, McGraw Hill Book Co.,

1964 4 Berkowitz, L., Seismic Analysis of Primary Piping Systems for Nuclear Generating Systems, Reactor and Fuel Processing Technology, Argonne Nati Lab, Fall, 1969. ..

5. Marki, A.R.C., Tatigue Tests of Piping Components. Trans ASME V 74 1952.

6. Brock, J.E., Expansion and Tiexibility, Chap. 4, Piping Handbook, 5th Ed.,

King and Crocker (Eds.) McGraw Hill Publishing Co., 1967.

7. Marki, A.C.R., Piping Flexibility Analysis, Trans ASME, 1955, p 419.

8. Criteria of the ASME Boiler and Pressure Vessel Code for Design by Analysis, Amer. Soc. of Mech. Engrs., 1979 (to be published).

9. Swiger, W.T., personal communication, May 1979.

10. Sviger, W.T., Notes on Plants Designed by Stone & Webster Which Have Experienced Large Earthquakes, 1979, Unpublished.

11. How Nuclear Piping Code Rules Will Influr nce Piping Design Today and Tomorrow, Heating, Piping, and Air Conditioni'.g, June, 1970, p 69.

12. The Great Alaska Earthquake of 1964 Engineering, National Academy of Sciences, Washington, D.C., 1973.

L_,, 13. San Fernando, California, Earthquake of February 9, 1971, Leonard Murphy, Sci. Coord, U.S. Dept. of Coma., NOAA, Washington, D.C., 1973. f

14. Snyder, Arthur I., Damage to Mechanical Equipment as a Result of the Feb. 9, 1971 Earthquake in San Fernando, California, Seismic Design and Analysis, Amer. Soc. of Mech. Engrs., 1971.

15. Managua, Nicaragua Earthquake of Dec. 23, 1972, Earthquake Engineering Research Inst., Nov., 1973.

16. Bohm, George J. Damping for Dynamic Analysis of Reactor Coolant Loop Systems, Optical Meeting on Reactor Safety, Salt Lake City, Utah, March 1973, Conf-730304 Available NTIS.

17. Bush, Spencer, 3atte11e Northwest Laboratories, Personal Communication.

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