ML18130A559
| ML18130A559 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/06/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18130A555 | List: |
| References | |
| NUDOCS 7909180244 | |
| Download: ML18130A559 (16) | |
Text
e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FACILITY OPERATING LICENSE NO~ DPR-32 VIRGINIA ELECTRIC AND POWER COMPANY SORRY POWtR STATION, UNIT NO. l DOCKET NO~ 50-280
Introduction On March 13, 1979; the Commission issued an Order to Show Cause to Virginia Electric and Power Company (licensee) requiring that Surry Power Station, Unit 1 (facility) be placed in cold shutdown and the licensee show cause:
(1)
Why the licensee should not reanalyze the facility piping systems for seismic loads on all potentially affected safety systems using an appropriate piping analysis computer code which does not combine loads algebraically; (2)
Why the licensee should not make any modifications to the facility piping systems indicated by such reanalysis to be necessary; and (3)
Why facility operation should not be suspended pend-ing such reanalysis and completion of any required modifications.
The licensee's response to the Order, dated April 2, 1979, stated that it is reanalyzing all potentially affected safety systems for seismic loads using an appropriate piping analysis method.
The licensee requested that the Order be modified or rescinded such that the facility could be restarted based_9n the resuJts of having analyzed all of the piping systems including Hcrzzles and penetrations which previously used SHOCK 2, all piping supports inside containment, and a portion of the piping supports outside containment. In support of this request the licensee provided information by letters dated March 30, April 2, 23, 24, 27, May 2, 22, 24, 30, June 4, 8, 12, 15, 19, 25, and August 1, 1979.
The licensee indicated that piping restraints in 19 problems needed to be modified based on its reanalysis to date.
Discussion 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 certain safety and nonsafety related systems at the facility. The licensee has identified the seismically analyzed (Seismic Category I) systems at the facility including those analyzed with SHOCK 2.
It has also identified the other methods of seismic analysis used for other Seismic Category I systems.
Further-more, the licensee has reported the results of the reanalyses of SHOCK 2 safety systems and has provided support for the acceptability of the analysis methods used on the remaining Seismic Category I systems.
We have evaluated the results of the seismic reanalyses and all the methods of pipe stress analysis previously utilized and used in the reanalyses for the facility.
Evaluation
- 1. Systems Portions of the following systems were identified by the licensee as having been analyzed with SHOCK 2.
Pressurizer Spray & Relief Low Head Safety Injection High Head Safety Injection Containment and Recirculation Spray Residual Heat Removal Component Cooling Water Service Water Main Steam High Pressure Steam Feedwater Auxi l i ary,.Seedwater **
Containment Vacumm Fi re Protectfon Diesel Muffler Exhaust The licensee has reanalyzed all 63 pipe stress problems originally analyzed by SHOCK 2.
In addition, the licensee reanalyzed 12 SHOCK 0 problems and 6 problems which were originally.done by hand calculation.
The reanalysis of those later 18 problems was not required by the March 13, 1979 Show Cause Order.
All supports inside containment were reanalyzed and modifications will be completed prior to startup.
A portion of the supports ouside containment have been analyzed and the remainder will be reanalyzed within sixty (60) days of the date of the Order allowing startup.
3 -
Of the 63 SHOCK 2 problems reanalyzed, 19 required hardware modification***
to bring the pipe stresses within allowables. These modifications consisted of 63 added, modified, or deleted supports. Also, modifi-cations to supports on 4 problems were necessary to reduce nozzle and penetration loads to acceptable levels.
Most of these modifications are due to differences between as-built and original design, except in Problems 508~ 517, 727 and 526C where the overstress condition was attributed, 1n part, to the incorrect use of intramodal combinations in the original seismic analysis. Support modifications for these problems are typically as follows:
(1)
Problem 743 - Low Head Safety Injection (LHSI) System.
Involves a shim to close a gap between a restraint and pipe.
(2)
Problem 548A -;Containment and Recirculation Spray (CRS)
System.
Install snubber.
(3) Problem 745 - CRS System.
Modify and replace restraints.
(4) Problem 746 - High Pressure Steam System.
Modify and add restraints.
(5) Problem 508 - Residual Heat Removal (RHR) System.
Add three vertical and one lateral restraints.
(6) Problem 540 - RHR System.
Remove rod hangers.
Anchor was added.
(7) Problem 744/754 - CRS System.
Modify *and replace restraints.
(8) Problem 562 - CRS System.
Change a rod hanger to a spring hanger and add a vertical restraint in place of a spring hanger.
(9) Problem 727 - LHSI System.
Install Snubber.
( 10) Problem 735 - HHSI System.
Install two restraints.
e
- (11)
Problems 766 - Component Cooling Water (CCW) System.
Install two horizontal restraints and remove one axial restraint.
(12) Problem 481/507 - CCW System.
Install two lateral restraints and two guides.
(13) Problem 480/488 - CCW System.
Same as problem 481/507.
(14) Problem 509 - CCW System.
Install one vertical restraint.
(15) Problem 605 - CCW System.
Install four lateral restraints.
(16)
Problem 526C - CCW System.
Change spring hanger to vertical restraint. Install lateral restraint.
(17)
Problem 2527/2529 - CCW System.
Remove an anchor.
Add supports on heat exchanger to make it behave as an anchor.
( 18) Problem 527C - CCW System.
Remove an anchor.
heat exchanger to make it behave as an anchor.
restraint *
- --;::.:.--=-..-~ **
Shim supports on Add vertical (19) Problem 517 - CCW System.
Add vertical support.
Add lateral restraints.
Add a snubber.
Add a vertical restraint.
Most of the problems were reviewed in detail by the NRC staff during meetings with the 1 icensee on June 21 and 22, 1979, and July 18 and 19, 1979.
- 2.
Soil Structure Interaction By letter dated March 30, 1979, the licensee stated its intent to use soil structure interaction amplified response spectra (SSI-ARS) in reanalyzing the piping systems for those cases where the origin~
amplified response spectra did not give satisfactory results. The licensee requested our review and stated that this approach was similar to that approved for Surry 3 and 4.
The NRC approved in concept the use of SSI-ARS by letter dated April 13, 1979.
Based upon review of the licensee's information submitted by letter dated May 24, 1979 as discussed below and our independent analyses, we informed the licensee by letter dated May 25, 1979 that SSI-ARS was acceptable.
The amplified floor response spectra (ARS) for three levels in the containment~ base mat, operating floor and spring line were computed using the multi-layered elastic half space method and the finite element methods.
The results of these analyses were compared for frequency and acceleration of the floor response spectra. The elastic half-space method gave acceleration values which were larger than the finite element method for the operating floor and the spring line. The finite element method gave accelerations slightly higher than the elastic half-space method for the containment base mat.
Since no piping systems are located at and would not use the base mat spectra for analysis, it was concluded the elastic half-space method would be used for the reevaluation because that would be conservative. The time history used for this comparison was the original design time history used in the original design of the plant along with the original damping values.
The same floor response spectra were generated for the Regulatory Guide 1.60 requirements anchored at 0.15 g along with the Regulatory Guide 1.61 damping values for comparison with the original earthquake input requirements.
The time history and the damping values are considered as a consist,ent set of design parameters~ The comparison of the FSAR design requirements and the Reg. Guide 1.60 and 1.61 set of values show that the responses are very consistent and that the original FSAR design requirements would be adequate.
A study of the effects of the variation of the soil properties was undertaken.
Tbe response spectra for the three locations in the containment builaing were computed for five (5) variations of the soil properties. Variation one considered the computed strain dependent properties using the best estimate of the in situ properties as input to computer code SHAKE; variation two used the in situ properties plus 50% as input to*the computer code SHAKE; variation three used the in situ properties minus 50% as input to the computer code SHAKE; variation four considered the first iteration value of the computer code SHAKE using the in situ properties as input; and variation five used the measured values (low strain) of the soil properties. This study indicated that the response of the structure to the variations in the soil properties is essentially limited to the amplitude of the floor response spectra. The peaks of the floor response spectra occur at the same place, but have different values of acceleration.
For comparison purposes, two curves were constructed using the ratio of variation five to variation one response spectra, and the ratio of variation two to variation one response spectra, plotted for the operation floor and the spring line of the containment. These ratios were plotted for all values of period from 0.0 to 0.8 sec.
e e The ratios of variation two to variation one for the operating floor range from 1.05 at.14 sec. to 1.75 at 0.19 sec. and for the spring line 1.0 at 0.12 sec. to 1.09 at 0.29 sec. The ratios of variation five to variation one for the operating floor range from 1.1 at 0.10 s_ec. to 2.55 at 0.33 sec. and for the spring 1 ine from 1.2 at 0.1 sec. to 2.95 at 0.29 sec. After considering the variation
- of the measured in situ*soil properties and accounting for uncertainty
.in the computer code SHAKE 1 s prediction of strain. dependent soil properties, we judged that the value of the floor response spectra acceleration would be not greater than 50% more than the floor response
. spectra acceleration calculated using the variation one of the soil properties. This being the case and since the licensee had already finished a large number of stress computations for the piping system using the response spectra based on the soil properties determined in variation one, it was determined that an increase of the values of the*response spectra already used in piping stress calculations by a factor of 1.50 would be acceptable. This increase in the a~celeration value for the floor response spectra would result in a*conservative re-analysis.
To further verify that this increase (1.5) is conservative, the staff conducted an independent study of the variation of soil properties used in the dynamii:.. analyses L First the staff confirmed the adequacy of the average soil.properties selected by the licensee and then considered parametric studies of these properties. The results of this effort indicated that a variation of +25 percent for the input shear modulus (G max) would accommodate-uncertainties in the in situ soil properties. The results of this variation appear to bound the possible range in soil properties based on staff experience with other site studies.
Therefore, the licensee's studies for +50 percent and the increase (1.5 factor) in the response spectra are conservative.
Because the soil shear moduli used in the generation of ARS depend upon the level of_ strain induced by earthquake motion~ the ARS are not in direct proportion to the maximum ground acceleration. Therefore, an investigation of the effects of earthquakes smaller than the DBE was also undertaken.
e For the purpose of this study, ARS's were computed for various average strain compatible shear moduli, each due to a peak horizontal ground acceleration ranging from 0.15 to 0.05 g.
The, licensee has provided the resulting family of ARS at the operating floor which show the DBE spectrum to envelope the other spectra due to smaller earthquakes. This demonstrated that the effects of DBE are not exceede? by those of smaller earthquakes.
Therefore, based on its review of the above infonnation the staff concludes that the stresses in piping due to the DBE are not exceeded by those due to smaller earthquakes.
The computer codes used in the re-analysis for the soil structure interaction were:
- 1.
SHAKE
- 2.
PLAXL Y
- 3.
REFUND
- 4.
KINACT
- 5.
FRIDAY The computer co_g_~ SHAKE is __ a public domain program and was used to compute only the strain***ctependent properties of the supporting soil under the structures. Because this code was only used to compute soil properties no further verification is necessary.
The computer code PLAXLY is a proprietary code and was qualified by
.comparison to the existing public domain computer code FLUSH.
Amplified response spectra for the containment operating floor computed by both codes were compared.
The computer code REFUND computes the frequency dependent compliance 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.
The computer code KINACT is a proprietary code and is used to compute the 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 sample 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 for each mass point in the mathematical model of the structure. The code is a proprietary program and was qualified by comparing the results of a sample problem with the results of the public domain program STARDYNEo The comparisons of the results for the above codes were favorable and are, therefo:e, acceptable by the current acceptance criteria.
- 3. Verification of Analysis Methods We have reviewed the acceptability of the analytical methods which are currently a basis for the facility piping design.
The licensee has identified the following computer codes/analysis methods as applicable:
PSTRESS/SHOCK O (In,itial 3 Versions of SHOCK l).
Static Analysis Methods NUPIPE PSTRESS/SHOCK 0 This code was u~~d for 12 safety related system problems and although it did not algeb-raically sum responses, the code was not equivalent to current practice~* The licensee, therefore, reanalyzed these systems with the NUPIPE code.
Static Analysis Methods used for design of the p1 ping at the Surry Pl ant not subjected to computer seismic analysis were based on simple beam formulations*
which, in essence, controlled seismic stress levels through use of pre-established seismic spans. These simple beam formulations were utilized to calculate maximum allowable spans based upon an assumed acceleration factor of 1.5 times the peak acceleration obtained from the response spectra.
In calculating the maximum span lengths, it was conservatively assumed that a longitudinal pressure stress of 4,000 psi and a maximum deadweight stress of 1,500 psi were present in the pipe. This combined value of 5,500 psi was subtracted from the allowable stress (1.8 Sb for pressure and deadweight and seismic) to obtain a seismic allowable stress.
Calculating maximum spans by this procedure results in maximum allowable***
spans greater than the deadweight spans recommended in ANSI B31.l. Thus deadweight governs and provides a greater number of supports resulting in closely spaced restraints.
To minimize effects of concentrated weights, restraints were placed as required at valves and other concen-trated masses.
For Surry Unit 1: piping 6 inches in diameter and smaller was generally analyzed using the simplified static method, with the option of utilizing more rigorous methods available to the analyst.
Piping 2 inches and below was shown on the piping drawings diagrammatically (i.e., without detailed dimensions).
The stress engineers located supports during the installation process working at the site with erection isometric sketches.
As described above, the stress analysis was performed by assuming many simple supported straight beams, the spans of which are governed by dead load spacing requirements of ANSI B31.l. The piping fundamental frequencies associated with these maximum allowable spans (9.7 to 13.6 cycles per second) are not in resonance with the building in which they are lqcated (2 to 8 cycles per second). The method of equivalent statrc--*analysis outlined in this procedure has been compared with the NRC's Stan~ard Review Plan 3.7.2 and is found to be acceptable.
NUPIPE In accordance with the letter of April 2, 1979 from V. Stello to the licensee, the licensee's Architect-Engineer,. Stone and Webster (S&W) has submitted documentation on the computer code NUPIPE which is being used in the reanalysis of the Surry plant.
S&~1 has stated that this code cal cul ates i ntramodal and intermodal responses according to the provision in Regulatory Guide 1.92.
A review of the code listing by the staff has confirmed this statement.
The option used by the licensee specifies an intramodal combination consisting of the addition of the absolute value of the responses due to the vertical earthquake component and the root-mean-square 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.
The licensee has solved three NRC benchmark p1p1ng problems and its solutions show acceptable agreement with the benchmark solutions.
In addition, it provided a confirmatory problem (No. 323A) to the Brookhaven National Lab for confirmatory solution. A comparison of the solutions demonstrates good agreement (within about 10%).
Based on these considerations we find the use of this code acceptable for seismic analysis by response spectrum techniques.
- 4. Reanalysis Method~ and R~sults The safety related piping systems at the Surry 1 nuclear plant have been reviewed to determine the method of analyses. Sixty three (63) computer stress problems of safety related piping have been identified where the analysis used the computer code SHOCK 2 which used an algebraic intramodal summation of responses to earthquake loadings. The problems where an algebraic intramodal response combination technique was used in the design have been reevaluated using acceptable m*ethods.
The reevaluation included a dynamic computer analysis using NUPIPE programs~ which incorporated a lumped mass response spectra modal analysis technique.
The floor response spectra used in the reanalysis included the original amplified respon{e",*spectra specified in the FSAR.
In some cases~
piping was reanalyzed utilizing ARS that were developed using SSI techniques. The peaks* in the amplified floor response spectra were broadened by +15% in accordance with Regulatory Guide 1.122 to account for variation-in material properties and approximations in modeling.
The piping systems were modeled as three dimensional lumped mass systems which included considerations of eccentric masses at valves and appropriate flexibility and stress intensification factors.
The dynamic analysis procedures meet the criteria specified in the plant FSAR and are acceptable. The resultant stresses and loads from the reanalysis were used to evaluate piping, supports, nozzles, and penetrations.
All of the 63 SHOCK 2 pipe stress problems have been reanalyzed and verified by Stone and Webster Engineering Assurance and the licensee's Quality Assurance Program. This together with all of the 6 original hand calculations and all of the 12 SHOCK O problems reanalyzed completed the entire scope of piping stress reanalysis.
Based on our review of the computer codes being used for reanalysis, independant check analysis performed by the staff and a review of modeling methods used by the licensee, we find acceptable the procedures and methods used in reanalyzing these problems.
11 -
The reanalysis included those SHOCK 2 problems involving the reactor coolant system boundary and the supports associated with those problems.
Since the reactor coolant system boundary is inside containment and all of the supports which must be modified will be modified prior to startup, there is no potential for a loss-of-coolant accident in the event of a DBE.
In the reanalysis~ the new total stress, at the point of maximum total stress in the pipe, and new seismic stress, at the same point, were taken from the NUPIPE computer runs with the seismic inertial stress magnified by a factor of 1.5 for runs using the SSI-ARS, as discussed in Section 2.
Of the 63 problems 56 used the SSI-ARS and 7 used the original ARS.
The stresses after the 1.5 magnification for the runs using SSI-ARS are below the allowable stresses. A pipe stress reevaluation summary provided by the licensee was reviewed by the staff to confirm that for the pipe stress problems reanalyzed, the total stress values were all below the allo,iable stress. Included i-n this summary was a 1 isting of the original total, original seismic, new total, new seismic~ and the allowable stresses for the pipe stress problems that have been reanalyzed.
At the request of the NRC, its consultant~ EG&G performed audit pipe stress calculations of five Surry 1 problems using the NUPIPE computer code.
The results of the EG&G audit compare favorably with the results of the licensee's results.
The piping support designs for affected system piping were inspected by the licensee to verify the location, orientation, support clearances, and support type.
Any deviations that were identified are incorporated into piping reanalyses. These piping systems were also verified by the NRC Office of Inspection and Enforcement.
The pipe supports were reevaluated in cases where the original support design loading was exceeded as a result of piping reanalysi~.
In cases where the original support capacity was exceeded, the support reevaluation has included the consideration of base plate flexibility
e and a verification of actual field construction of the support.
Where concrete expansion anchor bolts were used, their capacities, without compromising the originally committed safety margin, were also included in the reevaluation.
There are 1005 supports on lines origtnally analyzed by SHOCK 2; of these, all 487 supports are located inside containment and have been evaluated and subject to modification as identified in Table 4-2 of the licensee's August 1, 1979 submittal, are acceptable.
Sixty-six of these 487 supports were identified to require modifications.
Five supports outside containment are currently identified to require modification. There are approximately 400 supports remaining to be evaluated. During the reanalysis it was determined that the majority of the support modifications arose as a result of the 11 as-built 11 supports being deviated from the original design.
Only four of them can be qualified as due to inadequate, original ~eismic.analysis
~ncorporating algebraic summation technique.
A NRC staff evaluation of pipe supports inside containment which required modification compared the percentage of the originally calculated support load with respect to the ultimate capacity of the supports.
A5-"'of July 31 ~ 1979 results indicate 4 supports did not have at least-a factor of safety of 2 to ultimate. Ten other supports based upon p,rel iminary cal cul at ions may not have a factor of safety to ultimate of 2.
The licensee is continuing analysis of these problems, however, these 14 and the remainder of the 66 supports discussed above will be modified prior to startup.
Based on the results to date, we expect other supports may be found that will not have a minimum factor of safety of 2 to ultimate.
However, if support reanalysis indicates this we will require the licensee to inform the NRC of the results of reanalysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that the affected system be considered inoperable as specified in the facility Technical Specifications until the necessary modifications are implemented or a reanalysis assuming support failure is completed~
13 -
Loads on attached equipment nozzles and penetrations were checked and verified to be either below the initial allowable values or were evaluated and determined to be acceptable. Confirmation of the results of reanalysis will be obtained from the equipment manufacturers where necessary.
The design and analysis of the supports and attached equipment are in accordance with the criteria specified in the plant FSAR.
The pipe break criteria of the FSAR were reviewed in connection with the possible effect of changes of the high stress point resulting from the reanalyses.
Results of the evaluation of the effect the reanalysis has on the FSAR pipe break criteria show that no new whip restraints are required.
Therefore, we find that the reanalysis has not cha~ged the pipe break protection.
The piping systems and supports were designed to the allowable limits of ANSI 831.1 for the gross properties and to the limits of ANSI B31.7 Appendix F for local stress considerations per the FSAR criteria.
The safety related piping systems supports and attached equipment, where the original analysis used an algebraic intramodal summation technique, have been, or are to be reanalyzed with acceptable methods.
The procedures used in the support reanalyses and their results have been reviewed against the criteria in the FSAR and found acceptable.
- 5.
Conclusion The licensee has demonstrated that SHOCK 2 is the only method of analysis used for the facility's safety related systems which combines seismic loads algebraically. Safety related piping systems analyzed with SHOCK 2 have been reanalyzed with an acceptable dynamic code.
Results of the reanalysis indicated that the pipe stress and equipment 1 oads, after necessary modifications, will be acceptable when compared with the FSAR a11owab1es and the manufacturer's specified load criteria.
J
- The reevaluation of pipe stress indicated that modifications in only four problems were necessary as a result of the algebraic summation problem.
These modifications are identified in Section 1. The Licensee will complete all modifications inside containment prior to start of plant operation. Evaluation of the supports and schedule for completion of necessary modifications outside of containment will be cor.ipl_eted within sixty (60) days of the date of the Order.
Further, in those cases where reanalysis exceeds code allowable, the staff requires that the criteria used to determine whether a factor of safety of 2 to ultimate does exist be linear elastic analysis techniques or no more than twice the rated load for snubbers.
Use of detailed finite element an~ysis for evaluation of local stresses due to integral attachment is acceptable. Supports outside containment which exceed 50 percent of ultimate capacity (or twice rated load for snubbers) will be considered as inoperable as defined in the Technical Specifications.
We reviewed the analysis techniques which are currently the bases for the facility's piping design.
We have detemined that the application of these techniques, at Surry l, assures that safety related systems will withstand the design basis earthquake. Although the reanalysis of supports outside containment is not complete, there is reasonable assurance that the facility can operate during the interim period until the reanalysis and any required modifications are completed without endangertng the health and safety of the public. This assurance is based on the following factors:
(1)
All safety system piping outside containment which was originally seismically analyzed with the SHOCK 2 program has been reanalyzed and, subject to modification, is acceptable.
(2) All of.the affected safety systems inside containment have been reanalyzed (piping, supports, nozzles, and penetrations) and were found either acceptable as presently designed or will be modified as identified in this SER prior to startup.
(3)
The review of 487 supports inside containment identified only 4 calculated support loads exceeding 50 percent of ultimate capacity (10 other supports are still under review, however all supports are being modified).
It is therefore, reasonable to expect that few supports outside containment would exceed 50 percent of ultimate capacity.
J
j~
1.,
( 4)
(5)
( 6)
(7)
(8) e
- Confirmation of input data through "as-buil t 11 verification provid,es assurance that analytical results are correct and significant 11 as-built 11 deficiencies repaired.
The licensee has committed to complete analysis and any necessary modifications for the 46 remaining supports associated with high and low head safety injection, containment and recirculation spray and auxiliary feedwater systems prior to startup assuring that ECcs*systems and systems necessary for maintaining hot shutdown will be capable of withstanding a design basis earthquake.
The licensee has committed to complete all the support reanalysis outside containment within sixty (60) days of the date of the Order.
The probability of an earthquake exceeding the design basis earthquake durin'g the sixty (60) day period that the remaining support analysis is being completed is small and the licensee has committed to shut down the facility in the event of an earthquake which exceeds 0.01 g acceleration and inspect.all piping, penetrations, supports and nozzles which have not been reanalyzed for both OBE and DBE.
The NRC will require prompt notification and either resolution by reanalysis of the piping system assuming a failed support or modification of the affected support, if reanalysis of a support indicates loading in excess of 50 percent of ultimate capacity (or snubber loading greater than twice rated capacity).
Based on the above, we conclude that the licensee has shown cause why Surry l can be operated during completion of reanalyses required by the Show Cause Order of March 13, 1979.
Date: