ML19207B086
| ML19207B086 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/08/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19207B080 | List: |
| References | |
| TAC-11431, NUDOCS 7908230357 | |
| Download: ML19207B086 (16) | |
Text
f(y no) n( (k NUCLEAR REGULATORY COMMISSION n
UNITED STATES
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g W ASHINGTC N, D. C. %555
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o,%.v f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR DUQUESNE LIGHT C0f9ANY BEAVER VALLEY POWER STATION, UNIT NO.1 DOCKET NO. 50-334_
INTRODUCTION the Connission issued an Order to Show Cause to the On March 13, 1979, Duquesne Light Company (the licensee) requiring that Beaver Valley Power Station, Unit No.1 (the facility) be placed i.. cold shutdown and the licensee to show cause:
Why the licensee should not reanalyze the facility (1) piping systens for seismic loads on all potentially affected safety systems using an appropriate piping analysis computer code which does not combine loads algebraically; Why the licensae should not make any modifications (2) to the facility piping systems indicated by such reanalysis to be necessary; and Why f acility operation should not be suspended pending (3) such reanal,ysis and conpletion of any required modifications.
The Stone and Webster (S&W) PSTRESS/ SHOCK 2 computer code for pipe stress analyses sums earthquake loadings algebraically and is unacceptable for reasons set forth in the March 13, 1979 Order to Show Cause. This code was used in the seismic analyses of safety and non-safety related systems at the facility.
31, 1979, stated that The licensee's response to the Order, dated March they will reanalyze the affected facility piping systems using an appropriate piping analysis computer code which does not conbine loads algebraically.
Further, they stated that they will make any appropriate nodifications to the affected facil piping systems which they determine to The licensee requested be necessary based or < _sults of the -analysis.
that, upon conpletion of the reanalysis of and any necessary nodifications to the affected piping systens required to assure safe shutdown and accident mitigation capability of the Engiraered Safety Features and the Energency Core Cooling System (ECCS), the facility be pernitted to resume operation pending completion of reanalysis of the balance of the affected piping systems and any necessary modifications of the In support of this request, the renaining affected piping systens.
_13 1itensee provided infornation as attachments to letters dated April 10,
\\
790 8 2 3 03$1
, 19 and 25, May 23, June 11 and 19, and July 11, 18, 23, and 27, 1979 and Stone and Webster has provided information as attachments to letters and June 4 dated March 22 and 30, April 3, 6,13 and 27, May 18,1979, and 18, 1979.
DISCUSSION 1n this section of the Safety Evaluation, the actions conducted by theThe licensee and its conclusions regarding those actions are discussed.
NRC staff's Safety Esaluation of these actions and conclusions is set forth in the Evaluation section of this Safety Evaluation.
Systems The licensee has identified 184 pipe stress problems that used SHOCK Of these 184 problems, 63 were check runs of hand calculations.
2.These 63 problems are listed in the licensee's submittal in Appendix The static analysis nethod, i.e., hand calculations, B, Tabl e B-2.
is discussed and evaluated later in this Safety Evaluation Report (SER).
The remaining 121 pipe stress problems identified by the licensee for which SH0CK 2 was the calculation of record are in the following systems:
Reactor Coolant System Safety Injection System Quench Spray System Recirculation Spray System Charging and Volume Control System Residual Heat Removal System Component Cooling Water System River Water System Main Steam System Main Feedwater System Diesel Generator Exhaust Fuel Pool Cooling and Purification System The licensee has approached the reanalysis effort in a two phase program nanely: (1) systems, or portions of systems, required for plant operation that are acceptable based on the current stress reanalysis results described in this Safety Evaluation, and (2) systems, or portions of systens, that are not currently required for operation and will be addressed in the licensee's long term effort.
Of the above listed systens that used SH0CK 2, the licensee has stated that all of these systens, with the exception of the Fuel Pool Cooling and Purification Systen (FPCDS), and a portion the River Water System (RUS),
have been reviewed and found to be acceptable for operation.
m Rc'nalysis Methods and Results 19 the response spectra modal analysis The piping was reanalyzed u.
The piping was modeled as three dimensional lumped mass technique.
systems and included considerations of eccentric masses at valves and appropriate flexibility and stress intensification factors (SIF).
The resultant stresses and loads from the reanalysis were used to The computer evaluate piping, supports, nozzles, and penetrations.
The codes used to perform the reanalyses were NUPIPE-SW or SH0CK 3. The acceptability of these codes is discussed later in this SER.
floor response spectra used as input in the reanalyses included the original amplified response spectra ( ARS), as specified in the licensee's Final Safety Analysis Report (F5AR), and ARS developed using SSI methodology current soil-structure interaction (SSI) techniques.The peaks on the is discussed in greater detail later in this SER.
original ARS and new SSI-ARS were broadened +25". on the frequency to account for variations in material properties t.nd approximations in model ing.
27, 1979, show that with the addition Reanalysis results as of July of three supports, pipe stresses for lli out of a total of 116 affected f r the Design Basis problems are within their allowable value of 1.8 SThe total stresses for 26 of these h
Earthquake (DBE) loading case.
do not include stresses due to DBE seismic anchor movements (discussed Ninety problems indicate stresses lower later in this Safety Evaluation).
allowable for the Operating Basis Earthquake (OBE) loading than the 1.2 STwo of the problems do not include stresses due to OBE seismic h
condition.
The discharge lines of the quench spray pumps and part anchor movements.
of the recirculation spray piping, both inside and outside containment, were seismically analyzed by fEDIPE for DBE and water hammer loads to an These lines allowable of 2.4 Sh previous to the present reanalysis effort.
The contain four of the five problems that show stresses above 1.8 S.
h 1icensee states that they will reanalyze these usin] an allowable value The other problem with a calculated stress greater than 1.8 S h of 1.8 S.
There are two unreinforced h
is Problem No.122 in the River Water System.
branch connections on a segment of piping inside the turbine building that show an overstress condition.
Nozzle and penetration loads have been re-evaluated based on the results Of a total cf 131 nozzles,117 have been evaluated of the piping reanalysis.
and found to be acceptable fer both the OBE and DBE, and the remaining 14 are acceptable for the DBE.
All 58 penetrations contained in the affected problems have been evaluated and found acceptable for the OBE and DBE.
As nentioned above, three additional p!:e supports were determined to be Reanalysis required in order to maintain pipe strerses within allowable values.
The results also indicated that seven existing supports required modification.
problem nunber, system, support designation, and reason for the addition / modi-fication are discussed belcw:
Problem No. 833, Reaccor Coolant System - Vertical snubber added.
As-buil t dif ferences ninor. The additional snubber was required First, there was significant load reduction or for two reasons:
offset due to the algebraic summation perfomed in the original SH0CK 2 analysis. Second, a nore conservative stress intensification factor was applied in the reanalysis.
Prablem No. 217, Compcaent Cooling Water System - Addition of one This additional support was support, consisting of two snubbers.
required for the same reasons the snubber was added to problem 833.
The increase in stresses when the loads were combined by the square root of the sum of the squares (SRSS) is the primary reason that this suppcrt had to be added.
Problem No. 270, Component Cooling Water System - H-56 beii.g removed and new support being added adjacent to this location. One of the lugs to which H-56 is attached is overstressed locally due to dead-weight alone.
The r.ew support will eliminate the local overstress condition by utilizing a pipe clamp. The original design for H-56 was not adequate.
Problem No. 653B, Reactor Coolant System - Supports H-30, H-31, and Stiffener plates were added to these H-107 required modification.
supports and additional welding was done on the snubber bracket of H-107.
These modifications are a result of changes made to the supports during plant construction, (i.e., the as-built support details were not accounted for originally).
..L Problem No. 653B, Reactor Coolant System - H-38 was modified to remove one direction of restraint. Originally the support was a two-directional restraint acting in both the North-South (N-S)
I Reanalysis showed the N-S members and East-West (E-W) directions.
The N-S restraint was 2 moved and the problem to be overstressed.
rerun with acceptable results. The problem was reanalyzed with NUPIPE In the and modeling changes resulted in higher loads et the support.
original SH0CK 2 analysis the degrees of freedom were restrained and, therefore, lower than actual support loads were indicated.
Problem No.123, River Water System - H-32 angle members required Increased stiffener plates due to increase in upward vertical load.
load resulted from seismic anchor movement (SAM) case which was inadvertently ommitted in the original analysis.
P,roblen No.123, River Water System - H-33 required removal of lateral This support was installed as a 3-way restraint although restraint.
The re-the original piping analysis called for a 2-way restraint.
analysis showed that the lateral loads due to SAM overloaded the support.
Problem No.123, River Water System - H-309 required additional stiffener plates to the structural steel. The old and new loads were approximately the sane. however the original design was not adequate.
IE Bulletin 79-02 dated March 8,1979, revised June 21, 1979, on "Pipu r"pport Base Plate Using Concrete Expansion Anchor Bolts" provides direction on the re--
This reanalysis effort on pipe qualification of base plates and anchor bolts.
stress and support integrity interfaces with Bulletin 79-02 at the base plate /
The licensee has stated that if results indicate new supports anchor bolts.
are needed or existing suppcrts require modification, the base plates and anchor bolts shall be designed /e/aluated incorporating I&E Bulletin 79-02 criteria. Additionally, field inspections will be performM on those existing base plates being modified in order to ensure bolt integrity.
Including the seven requireo modifications and the three additions, there are 1063 supports on lines within the scope of the reanalysis Of these, 677 (including the seven modified and three added) effort.
have been evaluated and fcund acceptable based on FSAR criteria. Of the remaining 386, 384 are acceptable when the SAM 1 cad is renoved from the DBE loading condition. The remaining two have not, as of August 1, been accepted. However, the licensee believes that there is sufficient
'f n
analytical information available for these remaining two supports to exercise engineering judgment in determining that they will be found acceptable using the same criterion applied to the 384 already found acceptable.
Several snubbers have been found acceptable for a one time load equal to the DBE load.
Several of the total of 1063 supports have been evaluated for the DBE loading condition only. The licensee has stated that if a seismic event occurs which results in an acceleration greater than 0.01 g, the plant will be shut down for inspection of there supports and piping systems which have not been shown to be fully acceptable for the OBE The facility accelerometers and recording start at a setpoint case.
of 0.01 g.
The licensee has committed to checking the seismic instru-mentation for proper operation prior to startup.
The FSAR states that break locations have been postulated for the main steam and feedwater systems inside and outside containment. The licensee states that the reanalysis results show that the highest intermediate stress points occur in those areas where the lines are fully restrained by existing ' pipe whip restraints and, therefore, no additional restraints are required.
F_ield Verification of As-Built Conditions The licensee states that field verified piping fabricator isometric drawings provide the basis for program inputs for the pipe stress analyses.
Beginning ir Septenber 1974, and completed prior to facility startup, pine stress analysts and pipe support designers walked down all Category I (seisnic) piping systems and checked for piping configuration, support location and type. The results of this effort were documented and became part of the permanent plant record.
Licensee personnel verified the accuracy of a portion of these piping isometric drawings during March and April of 1979, subsequent to ti.e Order to Show Cause.
Verification of Computer Codes and Analysis Methods In accordance with the letter of April 2,1979 frou V. Stello to the licensae, the licensee's Architect-Engineer, Stone and Webster (SaW),
has submitted documentation on the conputer codes NUPIPE-SW and PSTRESS/
SHOCK 3 which are being used in the reanalysis of the Beaver Valley plant.
'h
,~
NVPIPE-SW SAW has stated that NUPIPE-SW calculates intramodal and intermodal responses A review of the according tu the provision in Regulatory Guide 1.92.
code listing by the staff has confirmed this statement. An option also anodal combination consisting exi;ts for users which specifies an intt of the addition of the absolute value of the responses due to the vertical earthquake couponent and the square root of the sum of the squares (SRSS) combination of the responses due to the two horizontal earthquake components.
Additional documentation has also been submitted by the originators of this code (Quadrex) providing detailed information on the methods of modal combination.
S&W has solved three benchmark piping problems provided by the NRC and NUPIPE-SW solutions show acceptable agreement with the benchmark solutions.
In addition, SSW provided a confirmatory problem (No.101) to the Brookhaven A comparison of National Laboratory (BNL) for confirmatory solution.
the solutions demonstrates good agreement (within about 10%).
PSTRESS/SH0C U S&W has stated that PSTRESS/ SHOCK 3 calculates the intramodal responses by adding the absolute value of the responses due to the vertical earth-quake excitation to the (SRSS) combination of the responses due to the two horizontal earthquake components. The intermodal components are calculated by the SRSS method. A review of the code listing has confirmed these statements.
S&W has also solved three benchmark piping problems provided by the NRC with this code and its solutions show acceptable agreement with the benchmark solutions.
In addition, a comparison of the S&W and BNL solutions of the confirmatory problem also demonstrate good agreements (within 10%).
Static Analysis Much of the 6 inch and smaller Category I piping at Beaver Valley Unit 1 was The methods were intended to keep analyzed using simplified static methods.
the fundamental piping frequencies out of the range of the fundamental struc-Calculations tur al frequency by establishing span lengths between supports.
Tabulations relating various spans, were based on simpie beam formulations.
nominal pipe sizes, and acceleration levels to actual pipe stress levels were
' A,
. The acceleration applied to the piping was provided for use by the analyst.
dependent upon where the piping fundamental frequency was rela structural frequency. Support loadit:gs were based on standarized loadings component earthquake.
Nozzle locis were calculated enveloping the various loading conditions.
based on similar, simplified methods.
Piping two inches and below wu shown on the piping drawings " diagrammatical (i.e., without detailed dimensions). The stress engineers located supports during the installation process working at the site with erection isometric
, sketches.
Small bore piping analyzed by a simplified static method were subjected to The results demonstrated the applicability of a NUPIPE dynanic analysis.
the method and standardized support loads.
Soil Structure Interaction The amplified floor response spectra (ARS) for three levels in the containment; i.e., base mat, operating floor and spring line, were conputed using the The nulti-layered elastic half space method and the finite element methods.
results of tnese analyses were compared for frequency and acceleration of the The elastic half-space method gave acceleration values floor response spectra.
which were larger than the finite element method for the operating floor and the-The finite element method gave accelerations slightly higher than spring line.
the elastic half-space method for the containment base nat in the frequency Since no piping systems would use the base mat spectra for range of interest.
analysis, it was agreed that the elastic half-space method would be used for The time history used for tnis comparision was the original reeval uation.
design time history used in the original design of the >lant along with the original damping values.
The same floor re';ponse spectra were generated for the Regulatory Guide 1.60 requirements anchored at 0.125 g along with the Reguletory Guide 1.61 damping The values for comparison with the original earthquake input requirements.
time history and the damping values are considered as a consistent set of design The comparison of the FSAR design requirements and the Reg. ';uide parameters.
1.60 and 1.61 set of values show that the responses are very consistent ano that the original FSAR design requirements would be ac' equate.
A study of the effcc.:ts of the variation of the soil properties was undertaken.
The response spectra for the three locations in the containment building were computed for five variations of the soil properties. Variation one was the computed strain dependent properties using the best estimate variation of the in situ properties as input to computer code SIMKE; two used the in situ properties plus 50% as input to the computer code SFMKE; variation three used the in situ properties ninus 50% as input to the computer code SHAKE; variation four considered the first iteration value of the computer code SHAKE using in situ properties as input; and variation five used the measured values (low strain) of the soil properties. This study indicated that variations in the soil oro n
The peaks of the amplified floor response spectra in the amplitude.
This peak broadening would are broadened by +25% on the frequency.
The enveloping procedure envelope the variations in frequency of each peak.
also accounts for the variation in amplitude by using the maximum amplitude of the variation one and two.
Because the soil shear noduli used in the generation of ARS depend upon the level of strain induced by carthquake motion, the ARS are not in direct Therefore, an investigation proportion to the maximum ground acceleration.
of the effects of earthquakes smaller than the DBE was also undertaken.
For the purpose of this study, ARS's were computed for various average strain compatible shear noduli, each due to a peak horizontal ground acceleration ranging from 0.125 to 0.01 g.
The licensee has provided the resulting family of ARS's at the operating floor which show the DBE spectrum This demonstrated to envelope the other spectra due to smaller earthquakes.
that the effects of DBE are not exceeded by those of smaller earthquakes and that the stresses in piping due to the DBE are not exceeded by those due to snaller earthquakes.
The conputer codes used in the re-analysis for the soil structure interaction were:
1.
SHAKE 2.
PLAXLY 3.
REFUND 4.
KINACT 5.
FRIDAY EVALUATION Systems We concur with the licensee's evaluation that the FPCPS, and those portions of th7 RWS that have either not yet been reanalyzed or do not meet the acceptance criteria are not necessary for operation.
The FPCPS does not peform any accident mitigation function nor is it The function required to achieve or maintain a safe shutdown condition.in the fuel pool.
of the FPCPS is necessary only if there is spent fuel Since Beaver Valley Unit No.1 has not completed its first nuclear fuel The function perfonned cycle there is no spent fuel in the fuel pool.
The licensee by the FPCPS therefore is not required for interir operation.
has committed to make any modifications to this system necessary as a result of the reanalysis prior to placing any spent fuel in the fuel pool.
There are two outstanding items to be completed on the RWS during the long These involve a portion of the RWS in the Intake Structure term effort.
that has not been analyzed and a portion of the RWS discharge piping in the turbine building that has an overstress condition.
The portion of the RWS that has not been reanalyzed is the portion of the Raw Water Pump discharge line (30"-WR-175-151-Q ) that runs under the floor 3
in the forebay of the intake structure.
The Raw Water Pumps supply cooling water to the turbine plant and are not required for accident mitigation Additionally, failure of this or to achieve and maintain safe shutdown.
line would not affect the operation of the River Water Pumps also located in the Intake Structure which are necessary for safe shutdown and accident The discharge line of the RWS (30" "R-17-151-Q ), has been 3
mitigation.
found to have an overstress condition at two unreinforced branch connections.
The RWS discharges to the main condenser discharge tunnel in the +urbine The overstressed branch connections are located in the turbine building.
building and failure of the RWS discharge line at this location would not The RWS affect the function of any safety related systems or equipment.
at this point is downstream from the coolant supply to vital safety components and since the RWS is a once through system, failure of this portion caused We concur with the by a seismic event is clearly not a safety concern.
licensee's determinatien that this overstress condition is acceptable for However, tf e licensee has committed to modify these branch operation.
connections prior to startup following the refueling cutage.
As a part of the continuing effort the licensee will reanalyze the discharge piping of the Quench Spray and Recirculation Spray Pumps and their associated This piping was spray distribution headers for the OBE loading condition.
analyzed with NUPIPE for the DBE plus water hammer loads, however SHOCK 2
. is the calculation of record for the OBE case. The suction piping for both the Quench Spray and Recirculation Spray System for whico SHOCK 2 was the calculation of record have been reanalyzed and found acceptable for operation. The discharge piping for both spray systems is acceptable for From operation and will be reanalyzed for the OBE in the lorg term.
a,ystems consideration, we find the licensee's evaluation acceptable ad sufficient to permit operation.
Reanalysis Methods and Results The three dimensional lumped mass response spectra modal analysis technique employed in the reanalysis is an acceptable method.
The three components of The earthquake response have been acceptably combined by the SRSS method.
analyses also considered eccentric masses at valves, (including correct weights-of VELAN 6 inch check valves, as stated in the licensee's response to I&E Bulletin 79-04), appropriate fl.aibility and stress intensification factors, and support flexibility.
Static Analysis In addition to the dynamic analysis (computer analysis) technique, we have also reviewed the static analy'.; method used for 6 inch and smaller piping.
Conservative weights had previously been assumed for the VELAN 3 inch check valves. The methods of equivalent static onalyses employed are similar to the procedure described in Section 3.7.2 of the Standard Review Plan and are acceptable.
Results of the pipe stress reanal;s;s show that, after the addition of three supports and the modification of seven others, stresses in all but five piping problems are below the allowable for the DRE loading case.
In accordance with the FSAR, the allowable is taken from the 1967 version of the ANSI B31.1.0 Code including addenda up to and including June 30, 1971. Additionally, DBE seismic anchor movement effects have been neglected for some piping problems and many supports.
Consideration of only the inertial pertion of the DBE load, i.e., neglecting DBE seismic anchor movement effects, is in accordance with Section III of the ASME Code, to which nuclear power plant piping is designed today, consistent with current practice and, therefore, acceptable. The licensee has committed to shut down the facility if a seismic event occurs which results in accelerations greater than an acceleration level er 0.01 g, the setpoint of the facility's acceleroneters, and inspect those piping systems and supports which have not been shown to be fullv acceptable for the OBE case (ground acceleration of 0.06 ).
This commitm.nt essentially resets the OBE for the plant at 9
1/6 its previous value and assures that no degradt. tion of piping, supports or nozzles will occur which night affect their capability to withstand the DBE.
The staff finds the 0.1 g for shutdown and inspection to be an acceptably conservative level for resumption of operation and until the OGE reanalysis is completed.
The accelerometer alarn is annunciated in the
^
,,~
Therefore, we find the evaluation of the facility capability control room.
to withstand a DBE acceptable for resumption of operation.
are on the Quench Four of the five problems with stresses exceeding 1.8 S h However, all stresses are below 2.4 S '
h and Recircul ation Spray Systems.
'tne currently accepted stress allowable for this loading condition for new Additionally, the licensee has committed to reanalyze these problems pl ants.
Since the stresses on the problems are using an allowable of 1.8 S.
b currently based on the old ARS, we 5elieve that reanalysis with the new SSI-ARS will result in stresses below 1.8 S
- h is Problem No.122 in The fifth problem showing stresses above 1.8 SThe two overstressed branch conn h
the River Water System.
the turbine building and we found this condition acceptable from a systems All stresses in the remainder consideration previously in this evaluation.
allowable value.
of this piping are belos their 1.8 Sh At the request of the NRC, INEL/EG&G* performed aud '
pipe stress calculations The on five Beaver Valley problems using the NUPIPE-II conputer code.
A direct results indicate all pipe stresses to be within allowable values.
comparison between the SH0CK 3/NUPIPE-SW stresses calculated by the licensee' Further, the results consultant and the EG&G audit results was not madc.
of the audit calculations indicate that seismic stresses may be significantly altered depending on support stiffnesses used and which method of seismic If piping natural response combination (algebraic vs. SRSS) is employed.
frequencies are close to the natural frequencies of the building, relatively small (a.g.,10-15%) shifts in piping frequencies can result in significant increases in accelerations. These frequency shifts uay occur when support stiffness is varied. The problems analyzed with NUPIPE-SW incorporated 5 - 107 lb/in) and, therefore, realistic support stiffness values (e.g.,10 For these problems the calculated frequencies are approximately correct.
analyzed with SH0CK 3, a tabulation of 10 and 15% frequency shifts and corresponding acceleradons indicate that, when the SSI-ARS is considered, the current pipe stress reanalysis results are "asonable or concervative.
The licensee has identified three pipe stress problems whose results are based on the original ARS and have natural frequencies in an area of the new SSI-ARS that is not enveloped by the old ARS.
However, results of a detailed examination of the current stress level to allowable value
- Idaho Nuclear Engineering Lab /EG8E (consultant to the NRC).
Y indicate sufficient margin available if the accelerations increased to those corresponding to the new SSI-ARS.
Based on the above evaluat'on, we find the piping stresses resulting from
.the reanalyses acceptable.
Results of the re-evaluation of all 131 nozzles and 58 penetrations are acceptable.
The support evaluation indicates all but two are acceptable, following the nodifications required on seven.
The licensee believes that, based on engineering judgment, both will be acceptable upon fu-ther evaluation using the DBE inertia load only and neglecting the DBE seismic anchor novement l oad. This criterion is acceptable to the staff as previously The licensee has committed that, prior to stated in this evaluation.
resumption of operation, these two supports will be determined acceptable or they will be modified to make them acceptable. We believe this commitment adequately addresses the acceptability of these two supports.
Some hydraulic snubbers h:ve been found acceptable for a one time load corresponding to the DBE ioad. The basis for their acceptability is an April 11,1979 letter from R. J. Masterson of ITT Grinnell Corporation, manufacturers of the snubbers, to M. Pedell of SSW.
Prior to Cycin 2 operation, the 1Sensee will have to quantify the loading and corresponding acceleration level that the snubber could be subjected to and remain within FSAR acceptance Niteria and revise the facility technical specifications to reflect this condition.
If this load or the acceleration le.el is exceeded, the snubber will be tested for operability pricr to continuing operation With this commitment, therefore, we find this criterion or returning to power.
acceptable.
The licensee may replace snubbers qualified by this criterion to nake them fully conform to FSAR criteria.
Other than the two supports and the snubbers discussed above, for the DBE case all remaining supports are in accordance with original design criteria, AISC Code and WRC Bulletin 107 for local stresses, and are acceptable.
In addition, the licensee has also committed to make any modifications to supports, excluding hydraulic snubbers themselves, discussed above, required to meet FSAR acceptance criteria for both the OBE and DFE loading Also, prior to return to power for the start of Cycle 2 operation, cases.
the licensee has conm1tted, by letter dated July 23, 1979, to complete the' seismic reanalysis of all safety related piping using the NUPIPE-SW computer code and the new SSI-ARS.
All piping stresses, support loads, and nozzle ar.d penetration loadings will be evaluated for both the OBE and DBE load conditions, based on their respective acceptance criteria.
All acceptance criteria will be in accordance with the FSAR or exceptions acceptable to the NRC staff, discussed above.
The use of the NUPIPE-SW computer code and the SSI-ARS has been found acceptable by the NRC, as l'
4 -
d evaluated later.
Further, the use of this computer code and these response spectra curves will adequately address the potential problems due to support flexibility and the new SSl-ARS not being enveloped completely by the old ARS.
Results of the evaluation of the affect the reanalysis has on the FSAR pipe break criteria show that no new whip restraints are required.
There-fore, we find that the reanalysis has not changed the pipe break protection of the facility.
Field Verification of As-Built Conditions An April 18-21, 1979, inspection of Beaver Valley 1 by NRC inspectors from the Office of ISE resulted in no items of noncompliance being identified within the scope of the inspection.
The inspection results are discussed in a May 25,1979, letter from R. Carlson of I&E to C. Dunn of DLC.
The inspectors exanined for accuracy the as-built safety related pipe supports and pipe system drawings.
Based on the information on the subject provided by the licensee, as discussed previously, ar.d (4a the results of the laE inspection, we believe that the reanalyses accurately reflect the as-built condition of the plant.
Verification of Computer Codes As discussed previously, the staff's review of the NUPIPE-SW and PSTRESS/
SNCK 3 computer code listings confirn that the codes calculate intranodal and intermodal responses as stated by S&W.
Also, solutions to the bench-mark and confirmatory problems demonstrate good agreement with the bench-mark and BNL confirmatory solutions.
Based on these considerations we find the use of these codes acceptable for seismic anaysis by response spectrum techniques.
Soil Structure Interaction The soil-structure interaction (SSI) analysis for the Beaver Valley Power Station, Unit No.1, has been reviewed against the current staff positions.
As discussed previously in this SER, the staff required studies 1) conparing ARS generated using the FSAR time history and canping values 2nd Regulatory Guide 1.60 and 1.61 requirements, 2) of the effects of varying the soil properties, and 3) investigating the effects of earth-quakes smaller than the DBE.
Based on the results of these studics, we conclude that the nethod used to develop the new SSI-ARS is acceptable.
The computer codes ined to develop the SSI-ARS were SHAKE, PLAXLY, REFUND, KI MCT, and FRIDAY.
The computer code SHAKE is a public domain program and was used to conpute only the strain dependent properties of the supporting soil under the structures.
Because this code was only used to compute soil properties no further verification is necessary.
PLAXLY is a proprietary code and was qualified by comparison to the existing public domain computer code FLUSH.
Anplified response spectra for the containment operating floor computed by both : odes were compared. The conputer code REFUND computes the f equency dependent conpliance functions for a multi-layered elastic half-space. This code is a proprietary code and was qualified by comparing the results of a sample problem with the results published in the literature. KINACT is a proprietary code and is used to compute translation and rotation time history at the base of the
. structure from the design time history applied at the free ground surface.
This code was qualified by comparing the results of a sanple problem to the results of the computer code PLAXLY.
The computer code FRIDAY uses the results of REFUND and KINACT to compute the floor response spectra The code for each mass point in the mathematical model of the structure.
is a proprietary program and was qualified by comparing the results of a sample problem with the results of the public domain program STARDYNE.
The comparisons of the results for the above codes were favorable and are, therefore, acceptable by the current acceptance criteria.
To verify that the licensee's proposed 125% peak broadening of the anplified response spectra was conservative, the staff conducted an independent study of the variations in soil properties which were used in the dynanic analyses.
First the staff checked the validity of the average soil properties selected The staff by the licensee and confirmed that the values were appropriate.
then conducted a parametric study using the computer code SHAKE with The variations of 150% from the best estimates of in situ soil properties.
results of this study indicated that a variation of +50% for the input shear The modulus would cover the uncertainties in the in situ soil properties.
lower -50% variation in properties was not considered representative of the soil s at the pl ant site.
It was also determined that the establishment of the actual lower variation bound was not necessary because the amplified response spectra of the best estimate properties and the +50% variation were shown to essentially envelope the spectra curve of the -50% variation in the frequency range important in pipe stress analysis.
Based on staff studies and a review of the licensee's work, the staff concluded that the proposed 125% peak broadening was reasonably conservative with one exception. Design ground motions in the free-field at foundations level were previously established by the applicant by calculating the site response due to a number of earthquakes, then enveloping the calculated site response with an assumed site independent response.
This procedure resulted in design motions with frequency dependent conservatisns, with minimum conservatisms occurring at the natural frequency of the soil deposit over-lying the rock.
In an effort to add conservatisn in the natural period range of the foundation soils, the staff required at least a 50% increase in spectral acceleration above the response curve which was developed using the best estimate soil properties. The natural periods of the foundation
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l soils was estinated to range from 0.4 sec. to 0.55 sec.
The :taff's re-quirement essentially caused a 20% increase in the amplified response spectra above the peak broadened spectra in the natural period range of the foundation soils.
Based on the above, and since the SSI-ARS used tcok into account the staff's reconnendation to increase the spectral accelerations by 20% in the period range of 0.4 to 0.55 sec., we find acceptable the +25 peak broadening.
CONCLUSI0tl Based on the above discussion and evaluation,
,e conclude that Beaver Valley Power Station, Unit No.1, nay resume power operation. This conclusion is based on the required modifications to seven supports, tiie addition of three others, and the twc supports not yet found acceptable are determined acceptable or modifications to make then acceptable being completed prior to startup.
Date:
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