ML18136A113
| ML18136A113 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 10/16/1979 |
| From: | Neighbors D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910290456 | |
| Download: ML18136A113 (23) | |
Text
Docket Nos. 50-280/281/339 LICENSEE:
Virgin~a Electric.and Power Company (VEPCO)
FACILITY:
Surry Power Station, Unit Nos. 1 and 2 and North Arma Unit 2 e
SUBt.1ECT:
SUMMARY
OF MEETING HF.LO ON SEPTEMBER 5 ~ 1979, TO DISCUSS SEISMIC ANALYSES FOR Sl.lRRY UNITS 1 AND 2, STEAM GEN!:Rti.TOR REPAIR FOR SURRY UNIT 2.AND LICENSING OF NORTM ANNA 2 The meeting was held 1:lith--the Commom.::ealth of Virginia State Corporation Commiss*ion (SCC) at its request to obtain information related to the subject topfcs. A list of attendees is attached (Attachment l).
Surry Seismic Analyses
.r.i. chronology of events 1P.ad1ng up to the shutdown -orders of Mar~ch 13, 1979, was presented by Mr. Russell. This cht'onology is presented.in. provides details of the modif'ications' required for Surry Unit 1 compared with modifications required for the, three other ola'nts for which the reanalyses are complete.* As can:be-,seen in this attach-ment, modifications were more extensive for Surry than the-other*
facilities.
- - provides a tabulation of -plants using algebraic summation techniques. Reanalysis of p1ping systems at facilities other than the five shutdm*m under the-orders has not resulted in overstress.
However.
some piping supports required modifications to torrect deficiencies in the as-built riondition.at several facilities. [Our letter dated September 18, 1979, to the Virginia State Corporation Commission provided samples of our evaluations for facilities using ulhebraic summation.] In each of these cases the licensee was able to demonstrate that stresses were v.'ithin allowable limits. -
Steam Generator Repair Pro~,ram The SCC 1s interest in this area was the NRC imposed requirements which t
- ,i e
n ri.
- i3.-..,c11.,s ~-~.C?W.~... t.~~-.!1J.i.1.~.t.QD.(.:.~.. Rf... tJm N.RC...r:~Y.i.e}1.. iL cJ..:th~.J:Jn:t~.. t!1.. r~P.~1.f................................
p ogram began.
Page 2 of Att chment 5 pro ides the NRC mposed aURNAMl::JI-DATE;;;>
- 7.1..10.~.. o. ~b.......................................... :............... :....................................................
ll&C PORM 318 (9*76) NRCM 0240 u.a. OOVERNMl!l:NT PRINTING OP'Flc~: 1071 - 2*a - 7e;g
- require:ment,s when thE.! l icen~e amendments were issued on January 20,
- 1979.. In th'e S,taff's opinH>n, the onl_y, NRC raised action wfiich rn;:iy have caused* som:e delay' was.if:mutual agreement with VEPCO to remove a portion of* a steam generator_**support for inspection.
Only about a week ~ouJd be *attributed to_,this-1nspectfon and is not siqnificant
.consi,de,~ing that the steam generator repair completio'n 'is *five to* six
~onths behind the original schedule. also provides the steps in the removal and storage of the steam-generators.
f//
- North Anna 2 Licensing_
- 1.
'",/: -
I With respect to-the licensing status of North Anna Unit 2~ we advised the Commonwea 1th that there were several outstanding. non-Three Mi 1 e Island related 1icensinq issues which must be resolved before the issuance of an oneratinq license.
We also advised that our Office of Inspection and Ei1forcement has identified four construction related iteins which must be completed before fuel loading can be.Pei~mitted.
Attachments:
As stated cc ~;/attachments:
) See next page I
I Qdgina1 Signed B~
D. Nei9hbors, Project Manager
.Operating Reactors Branch 1n Division of Operating Reactors OFl'ICE.
DOR.:.ORB/tl......... DOR.:.ORB#.1.......................................................... ************************* *********************
llURNAM** DNeighbor.s:.sh.1\\S.ch11,1.er.u;.6!.r. *....* '....................................................................................................
DAT!t.
....J.9./.....!.?.<:)...... 1-.., ** l.Q/..... /.7..~........ s.*.* **** ***** ********** ** **.* ****** * * * *.... * * * * **.......................,.....,.., *.*,.,.. *******
m.c PORM 318 (9-76) NRCM 0240 U.U, GOVERNMENT PRINTING OP"FICE: I 9715.. 201
- 7.9
e Commonwealth of Virqinia State Corporation Commission J. Urttine C. Crofton
- W. Stephens L. Ivey NR.C D. Neighbors W. Russell D. Eisenhut L. Shao A. Lee O. 0. Liaw
.. 1. M. Cutchin P. Ke 11ogg A. Iberdt
, E. Christenbury J. Reisland J. Henderson VEPCO H. Spencer A. Parrish R. Gary (Hunton & l*Ji11iams)
Stone & Webster rt Gi ambusso W. Leland L. Senn C. Grochmaii LIST OF ATTENDEES SE~TEMBER 5; 1979 e
ATTACMMEf.JT l
......... _* ********.......... *~*-********************** ~......................... ~................ **:***.............. *********************
........ t**.......... ******* **t**.......... ********t***********************t-*********........
- t**........ :..... *-i---******...........
DATIi:..
NllC FORM 318 (9-76) MlCM 0240 u.m. GOV.!!RNMKNT PRINTING OP"!"ICr:!:: 1971 - 2*D - 709
MEETING
SUMMARY
DISTRIBUTION:
Oocqket~~-
50-201...J w=-33g--
ORB#l Reading NRR Readinq H. R. Denton E.G. Case D. Ei senhut ft Vollmer
- 13. Grimes W. Gamr:1ill J. Mi 11 er ~
L. Shao T. Carter D. Crutchfield D. Zienann V. Noonan Seismic Review Group A. Schwencer T.
- Ippo1 i to R. Reid, G. *Lainas P. Check F. Pagano*
G. Kniqhton*
D. Neighbors C. Parrish OELD 0I&E(3)
R. Fraley, ACRS{16)
W. Russell f\\. Lee B. D. L ia\\V
,J. Cutchin P. Kellogg A. Iberdt E. Christenbury
~J. Reisland J. Henderson
,J. H. Buchanan TERA rns.,,,,.111 ccp
?9lo~o~~""b
~*****.. ~ _.............. :...... ~*-*"********************~*********************** *"* :****************************************************:**************
om*** \\*........,........ t-***.............. *t********************** t-**............... t******'************ t****.............
DAT!!:')>
MC FORM 318 (9-76) NRCM 0240 U.3. GOVERNMl!NT PRINTING OP'~ICE: 1971! - 2458
- 7S9
e ATTACHMENT 2 CHRONOLOGY TABLE PIPE STRESS ANALYSIS ISSUE Note:
This chronology represents 1nfonnatfon available to the NRC as of March 17. 1979. It does not necessarily fully or accurately.
reflect.ihe actual sequence of events which occurred prior to the March 8; 1979 meeting.
10/2/78 Stone and Webster notified the Beaver Valley Unft 1 Station Manager*of,n error discovered in the orfgfnal hand calculated stress analysis of.some safety injection lines. The error was discovered while evaluating the impact of correcting the weight 10/13/78 10/23/78 10/26/78 on 14.:safety fnjectfon system check Ytlves. Since this was. technically. a deviation frcxn the Final Safety Analysis Report, ft was to be referred to the Stat i,on Safety Committee.
To that end. the Stat ion Superintendent asked for more specific fnfonnation on the*error*.
At a ~eetfng held at the Beaver Valley site between Duquesne Lfght Company and Stone and We~ster repre-
. sentatives, additional 1nfonnation on the error was provided but more specifics were requested by OLC*.
Stone and Webster provided*DLC more fnfonnation.
The Station Superintendent asked for additional clarification and was told Stone and Webster personnel would be at the site the next week*.
During a site visit, Stone and Webster fnfonned DLC that one safety injection line would actually be significantly overstressed.
OLC then made a prompt telephone notification to Region I of the Office of Inspection and Enforcement*.
- These entries provided from memory by Duquesne Light Company representative on March 17. 1979.
10/26/78 10/27/78 10/27 /78 10/31-11 /3/78 11/9/78 e
e A-2.
Prompt report LER 78-053/0lP to NRC Region I vta telecon from Duquesne Light Company. Reported tnfonnation received from Stone and Webster that hand calculation errors resulted tn stress levels above ANSIB 31.1, 1967 but only tn one case of six flow paths
- Daily Report by Region I to IIE headquarters included as I reportable _occurrence - inadequate p1p1ng supports during review of safety 1nject1on pipe stress analysis by the A/E (S&W). several points on the 6-inch and smaller piping were found to be inadequately supported. In the event of safety injection system operation during I DBE, 5 points could exceed the code allowable stress. A desfgn change for safety injection piping supports will be accomplished prior to unit startup in mid-November.
Written interim LER submitted by Duquesne Light Company.
DLC characterized the errors reported by Stone and Webster as resulting frcrn a hand calculation method of analysis.
IE Inspection 50-334/78 Region I followup on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report.
Inspector raised a number* of questions including: What assurance can be given to show that the calculational error applies only to the six points in question? To only the Safety Injection system? To only the Beaver Valley facility?
Second interim LER submitted'by Duquesne Light Company indicates.. that the original report was erroneous. The line stresses were thought to have been hand calculated only. when in fact they were subsequently computer calculated and found acceptable.
DLC also indicated that 11 full report on the situation was in preparation by Ston~ and Webster.
- ~*-,,
?
11114-17 na 11/16/78 11 /30/78 12/01/78 12/04/78 12/05/78 12/06/78 e
A-2
- 3 *.
IE Inspection 50-334/78 Reg1on*I inspectors followup but no 1nfonnatfon available. onsite.
Region I Daily Report 1ndfcated I rereview by A/E found that the previously reported condition was erroneous and that no inadequately supported piping existed, a full report of the situation fs being prepared by the A/E and. a followup to the LER will be submitted by the Licensee to NRC.
Followup calls to s1te by the IE inspector attempting to seek additional fnfonnation.
Followup calls to site by the IE inspector attempting to seek additional fnfonnation.
Followup calls to site by the IE inspector attempting to seek additional 1nfonnation.
Followup calls to site by the IE inspector attempting
- to seek additional infonnation.
LER 78-53/0lT-O was submitted to NRC by licensee.
Conclusion was that *corrective action has been
.reviewed, approved and satisfactorily completed*.
The report based on fnfonnation supplied by Stone and W~bster *attributes the pipe overstress to diff-erences between stresses analyzed by PSTRESS code and those done by the chart method. It mentions differences between PSlRESS and NUPIPE codes in force summation but does not elaborate on them.
It concludes that PSTRESS used methods acceptable for Beaver Valley Unit 1 generation plants. It states that Reg. Guide 1.92 issued in December 1974 established for facilities docketed after April 1975 more conservative techniques for intra-modal combinations of generalized loadings. The report states that analysis showed that only one safety injection system pipe required modification -
the addition of one snubber and the redesign of one support. The attachment to this LER provided additional historical information as follows:
e A-2
-~-
Duquesne Light Campany reported 1n 1n attachment to the December 6, 1178 LER 78-53/0lT-O that to generate data needed for fnstallatfon of a net positive suction head modification to the Beaver Valley Unit 1 safety fnjectfon slstem, they (Stone and Webster) decided to *code 1n the sfx inch SI lines into a currently used computer program (NUPIPE).
DLC indicated original design used the PSTRESS code.
No results of an analysis at this stage were reported by DLC to NRC.
Subsequent to the above activity the attachment states the Beaver Valley Power Station was notified by a vendor that check valves in SI system were actually heavier than used 1n design at construction stage.
This increased weight was used as input to the above NUPIPE model and found not to *affect* the piping design.
The Architect Engineer (Stone and Webster) also concluded that the hanger designs need not be changed as a result of using the correct (heavier) wefght for these valves. However errors were safd to have been discovered in the hand calcu-lation method. It was detenn1ned that piping analysis showed local overstress at several anchors but
- no overstress 1n *the pipe* alone.
Per attachment to LER 78-53/0lT-O, a more thorough evaluation was fnftiated to detennine ff *any other annulus piping* originally designed by 'the chart (hand calculation) meth~d was overstressed.
Per attachment to LER 78-53/0lT-O, licensee found that SI lines had been *as-built* reviewed in 1974 and that two of the six lines had been (at that time) coded tnto PSTRESS (not just hand calculation method).
The PSTRESS code was re-run using the correct valve weights and resulted in acceptable ptpe stresses.
Al so per attachment to LER-78-53/0.lT-O, licensee states *rhe models run in PSTRESS and NUPIPE are geometrically similar; however, the mass distribution and support stiffness are different. Further, the method of force summation (intra-modal) is different.
NUPIPE utilizes*more conservative techniques for intra-modal combinations of generalized loadings *
- f
12/11/7~
12/12/78.
12/14/78 e
A-2
- These newer techniques arose following establishment of Beaver Valley Unit No. 1 design criteria. In December, 1974, the USNRC published Regulatory Guide 1.92, applicable to facilities docketed after April, 1975, which required the use of the more conservative combinations. The PSTRESS methods used were accepted dynamic analysis techniques for Beaver Valley Unit 1 generation plants.
and 1s the basis for all computerized Category I pipe stress analysis perfonned".
(It 1s NRC understanding that results were unsatisfactory on two of three lines. but snubber and support modifications on one line reduced the overstress on the second line such that no modifications on that line were necessary.)
The pre December 61 1978 review of annulus seismic piping was 11m1ted to lines that had been previously analyzed using the hand calculation method (2-1/2 1nch to 6 inch lines). 103 lines were identified, 55 were reviewed and found acceptable. Licensee noted. that PSTRESS results were still available for 48 of the 103 lines from the 1974 as built review and were *acceptable*.
Licensee notes its Engineering Department 1s *continuing a review of the architect-engineer findings*.
Followup calls to site by the IE inspector to seek additional infonnation.
Region I IE inspector telephoned NRR Licensing Project Manager to obtain a contact for infonnal discussion of technical questions.
Region I Daily Report -
Further review of in-containment SI system piping supports identified one line requiring support modification, attributed to an error in original design calculations.
Regional inspector was telephoned by NRR individual who was designated as contact. Preliminary technical discussion was held about potential problems.
/
12/18-20./78 12122ns 1/23/79 About 2/2/79 2/2/79 2/5/79 3/1 /79 e
A-2.
IE.Inspection S0-334n8 RegfoniI followup on 12/6 LER.
During this inspection, the inspector reviewed the detailed report submitted to the licensee by A/E and discussed the results of that review with representatives of the licensee and A/E.
\\
\\
Region I inspector discussed with NRR 1nd1v1dua1s via telep~one questions he had as a result of discussions he had with S&W on 12/1B-2ons. The NRC individuals involved determined that there was I possible problem.
Region I mailed to IE Headquarters 1 11emorandum request f ng that information be forwarded to NRR for review. The memo defined concerns to include:
- 1. Reconciliation of the differing analysis results to assure that the design methods ~sed a~e
- neither incorrect nor unconservatfve.
- 2.
The need for further 11censee; review of piping potentially affected by any incorrect or nonconservative calculation.
The IE Inspe*ctor provided copy of the 01/1~n9 memorandum to Licensing Project Manager.
Discussion between IE inspector and NRR project manager determined that a fonnal transfer of lead responsibility between I&E and NRR had not been made of the 01/18/79 memorandum to NRR.
A fonnal request for DOR's Engineering Branch support (TAC form) was prepared by the project manager.
IE f ns'pector was fnfonned by IE: HQ that telephone discussion had established that NRR was working on the problem and that a formal transfer of lead to NRR would be made.
Durf ng a conference ca 11 to DLC and S&W. a. canputer run was requested for DOR review.
Since S&W corporate policy was not to provide S!.!ch proprietary data, a meeting was set up for S&W to bring 1n a computer run for DOR review at Bethesda
- 3/8/79 3/8/79 3/9/79 3/9/79 3/10/79 e
A-2 A*technical meeting was held between DLC, S&W, and the NRC staff to discuss and review the PIPESTRESS 1nd NUPIPE codes. The NRC approached the review with the belief that the two codes were acceptable and that some modeling or input problem created the results in question. It was revealed that the PIPESTRESS code used an algebraic SUT11T1at1on of seismic loads which 1n the absence of a detailed tfme history analysis, gave unconservative results fn the seismic stresses. Management was fnnediately_
fnfonned and a management level meeting arranged with DLC and S&W.
A management 1eve1 meeting was held with DLC and S&W to arrange for inwnediate review of the Beaver Valley pipe stress analyses.
Commftments were requested of S&W to identify the systems and plants involved, the inadequacies expected and the reanalysis to confinn safe operation.
No definitive infonnation was available at that time.
DLC was requested to have its plant safety canmittee review the situation.
Numerous staff meetings were held at Bethesda to scope the problem with respect to the effects ff a seismic
- event were to occur.
Te,1 econs were made to S&W on the schedule of commitments for further fnfonnation on Beaver Valley. The other utilities identified by S&W as having plants with the same problem were notified. These plants and utilities were:
Fitzpatrick, Power Authority of the State of New York; Maine Yankee, Maine Yankee Atomic Power Company; Surry 1 and 2, Virginia Electric and Power Company.
The Chairman was advised. Three staff members were sent to Boston to provide ir.rnediate review and analysis of results.
DLC sent eight people to Boston to a 1ssi st 1n expedft ing the review.
In view of the problems and with the Offsite Safety Review Cormnittee concurrence, the Beaver Valley ijnit l was placed in hot standby for the weekend by DLC to await further analyses fra:1 S&W.
Staff meetings continued as pieces of fnfo.nnation were fed back fran Boston. The I&E Duty Officers were advised of actions. The NSSS vendors for the plants were contacted to assure no -Other codes for
3/11 /79 3/12/79 3/13/79 e
A-2 pf-pe stress during that period used the same algebraic approach. A DOR Assistant Director was sent to Boston to provide management review and coordination.
S&W' s computer was dedicated.fu1J time to these stress calculations and extended work hours for data reduction was instituted for S&W staff. NRC options were explored and draft materf als developed to support appropriate action based on the technical results becoming availab1e on Beaver Valley
- Early S&W reanalysis results on Beaver Valley runs indicated problems with pipes as well (originally thought only supports). Licensees'-top management was contacted to assure action underway by 111 plants to identify inadequacies and obtain reanalyses of stresses in all affected safety systems.
Additional fnfonnatfon fran DOR staff in Boston confirmed pipe stresses above all~wable and unaccept-able.
Arrangements were made to brief the Commission on this matter. All the licensees were notified of a pending decision.
In view of t~e safety significance of this matter as discussed above, the Director of the Office of Nuclear Reactor Regulation ~reposed to the Commission that the public health and safety requires that the present suspension of operation of the facility should be continued:
(1) until such time as the piping systems for all safety systems.have been reanalyzed for earthquake events to demonstrate confonnance with General Design Criterion No. 2 using a piping analysis canputer code which does not contain the error discussed above, and (2) ff such reanalysis indicates that there are ccrnponents which deviate frcrn applicable ASME Code requirements, until such deviations are rectified. The Commission concurred 1n the NRR Di rector's decision. *.
Prior to the NRC final decision to order the plants shutdown, the Beaver Valley Offsite Safety Review Committee recommended the facility be p1 aced in cold shutdown based on the data and analysis recieved fran S&W.
The DLC ordered the plant shutdown.
3/14/79.
3/16-17 /79 A-2
-*9 -
\\
The.1 f censees confirmed by tel econ that the Orders were received 1nd provided times when each facility would be 1n cold shutdown., All f1cf11t1es will be at or below 200°F by 10:40 p.m. on March 15, 1979 in confonnance with the Order.
Subsequently all affected licensees were notified by telephone that the Orders were executed and that a copy would be transmitted by facsimile
- Meetings were held with Stone and Webster with the Utilities to discuss acceptable methods of analysis for interim and long tenn fixes of the piping and supports.
Maine Yankee Beaver Valley Fitzpatrick Surry 1 Total 19 19 0
0 0
0 116 114*
2 2
0 3
96 91 05 1
4 9
63 44 19 4
15 64 294 268 26 7
19 76
..;,,,-------.-/
7 7
2 0
1060 1052 8
0 3069 2251 107 27 2
8 80 0
0 0
0 1
4 (107) 711 5to15 FOUR PLANT TOTALS THREEPLANTTOTALS
- includes 4 problems previously analyzed for DBE & Waterhammer against 2.4 Sh vice 1.8 Sh and 1 river water branch line as discussed in SER.
Note:
~Inside Containment
~Outside Containment
- -*:*.. ***.:-*:*.*.. ~.:.:;*.*:::*:
)::,
-I i!
n :c 3:
rr,
- z
-I W*
e
PL/\\NT
- Beaver Valley 1 Brunswick 1,2*
Cook 1, 2 Cooper Fitz pa trick Ginna Indian Point 2 Indian Point 3*
Maine Yankee
SUMMARY
OF PLANTS USING I\\LGEBRAIC SUMMATION*
SHUTDOWN REQUIRED ORDER OTHER Yes No No No Yes No No No Yes Voluntary No No No No No EXTENT OF SYSTEMS ANALYZED USING ALGEBRAIC SUMMATION.
TECHNIQUE.
Extensive Extensive Main Reactor Coolant loop and some lines in-side containment SRV lines Extensive Main Steam and RHR Lines l O lines Extensive 19 lines (Initially thought to be extensive)
AITACHMENT 4 COMMENTS Complete and Order Terminated 8/8/79 Reanalysis in progress. Analysis to date indicates modifications will no~
be required.
Complete Under staff review Order pennitting startup issued 8/14/:_t Complete Reanalysis in progress. Unit is shut
_ f~r refueling. Operation for 5 *weeks refueling permitted based upon prelin analysis Reanalysis. in progress. Analysis t-indicates modifications will not be, quired Complete and Order terminated 5/24/7!
Millstone 1 No No*
2 systems (Control Rod Complete Drive Exhaust and CU2 By-f pass) -
- J_*.
"' ~mr. St;iff h;i-; rr.vicw('rl hasis for opcriltion l
,,,,.., *,*,nr;ilinn\\1i1S
- r-* *-- -----,*----~----~_..~-~* ** ---~------.*.r.K....,_::,_*.. --
- PLANT I,,
- \\
Mill stone 2 I
'1 I
j L
_Nine *Mlle Pt l Pilgrim
. Pt. Beach 1,2 Robinson 2 Salem 1
. Surry 1, 2
- Turkey *Pt. 3,4 Zion l, 2
SUMMARY
OF PLANTS IJS lrHi ALGU3R/\\l L '.'lUi,li*1i\\ U UiJ
- (Continued
- SHUTDOWN REQUIRED..
- A-4 ORDER OTHER EXTENT OF SYSTEMS ANALYZED usrn*G*
ALGEBRAIC *SUMMATION "TECHNIQUE COMMENTS.
No No.
No No No Yes No*.
No
. "No No Tech Spec No No Extended Refuel.
No
. No 6 systems {Volume Control Complete Tank Chan9ing Byp~s~,
Nitrogen Addition, Charging*,
Diesel Generator Exhaust, RCP Top Root Valve Instru-ment, SI and Containment Spray Test Line) 7 systems (Reactor Recir-Complete culation, Shutdown Cooling, Emergency Condenser Returns, Reactor*Cleanup, Reactor Drain, Reactor Feedwater CRD).
Recirculation and Main Steam Complete lines 2 CCW and 2SW lines in radwaste system*
Complete Complete I.
Main Reactor Coolant Loop Extensive Unit shutdown for refueling
- e Ex tens i've Order permitting operation of Surr) issued 8/22/79.
- Surry 2 shutdown for steam generator repair.
Main Reactor Coolant Loop Main Reactor Coolant Loop.
Complete Complete
~
,_ ---- -~*-
.,-~,-.~-. *-=*...
--****- ~-::.:.-- --.:.-
A-4
SUMMARY
OF PLANTS USING ALGEBRAIC SUMMATION (CONTINUED
- j.
j I
- PLANT (Under Construction}
Salem 2 Forked River
! WNP 1, 4
- 1 I.
I!
I i
SIIUTDOWN REQUIRED ORDER OTHE-R EXTENT OF SYSTEMS ANALYZED USING ALGEBRAIC SUMMATION TECHNIQUE Extensive (Reactor Coolant System excluded)
Containment Spray ASHE Code Class 1 Reactor Coolant System Branch Lines COMMENTS Reanalyses and implementation of any required modifications prior to criticality.
e Reanalyses and implementation of any required modifications prior to receipt of operating license.
Reanalyses and implementation of a~y required modifications prior to receipt of operating license.
e
e
'TACHMENT 5 STEAM GENERATOR REPAIR CHRONOLOGY HIGHLIGHTS VEPCO SUBMITS PROGRAM NRC ISSUES SAFETY EVALUATION
... '.. '~
NRC ISSUES LICENSE AMENDMENT
\\
~
WITH ENVIRONMENTAL IMPACT.
APPRAISAL SURRY UNIT 2 BEGINS REPAIR PROGRAM AUGUST 17, 1977 DECEMBER 15, 1978 JANUARY 20, 1979 FEBRUARY 4, 1979
A-5 NRC LICENSE REQUIREMENTS
- 1.
ALL FUEL SHALL BE REMOVED FROM THE REACTOR PRESSURE VESSEL AND STORED IN THE SPENT FUEL POOL,
- 2.
THE TEMPORARY CONTAINMENT AND VENTILATION SYsrEMS.
SHALL BE OPERATING FOR ALL CUTTING AND GRINDING OPERATIONS INVOLVING COMPONENTS WITH REMOVABLE RADIOACTIVE CONTAMINATION >2200 DPM PER 100 CM2,
- 3.
THE HEALTH PHYSICS PROGRAM AND PROCEDURES WHICH HAVE BEEN ESTABLISHED FOR THE STEAM GENERATOR REPAIR PROGRAM SHALL BE IMPLEMENTED, 4,
PROGRESS REPORTS SHALL BE PROVIDED AT 60 DAY INTERVALS FROM THE START OF THE REPAIR PROGRAM AND DUE 30 DAYS AFTER CLOSE OF THE INTERVAL WITH A FINAL REPORT PROVIDED WITHIN 60 DAYS AFTER COMPLETION OF THE REPAIR, THESE REPORTS WILL INCLUDE:
- A
SUMMARY
OF THE OCCUPATIONAL EXPOSURE EXPENDED TO DATE USING THE FORMAT AND DETAIL OF TABLE 5,3-1 OF THE REPORT ENTITLED "STEAM GENERATOR REPAIR PROGRAM,"
AN EVALUATION OF THE EFFECTIVENESS OF DOSE REDUCTION TECHNIQUES AS SPECIFIED IN CHAPTER 6 OF THE REPORT ENTITLED, "STEAM GENERATOR REPAIR PROGRAMS" IN REDUCING OCCUPATIONAL EXPOSURES, AN ESTIMATE OF RADIOACTIVITY RELEASED IN BOTH LIQUID AND GASEOUS EFFLUENTS, AN ESTIMATE OF THE SOLID RADIOACTIVE WASTE GENERATED DURING THE REPAIR EFFORT INCLUDING VOLUME AND RADIOACTIVE CONTENT,
- 5.
SIXTY DAYS PRIOR TO FUEL LOADING, THE PROGRAM FOR PREOPERATIONAL TESTING AND STARTUP SHALL BE SUBMITTED FOR NRC REVIEW,
e e
A-5 STEAM GENERATOR CUT LOCATION I
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1-
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CUT
SCHEMATIC OF STEAM GENERATOR-LOWER ASSEMBLY REMOVAL ANO INSTALLATION SURRY POWER STATION - UNIT NO. 1 AND 2 I UPENDING I DEVICE I I POLAR CRANE CRANE:'"
WALL
,.-CONTAINMENT WALL J: :*.
GROUND LEVEL ~
HATCH PLATFORM
):::,
I u,
STEAM GENERATOR TRANSPORT ROUTES SURRY POWER STATION UNIT NO. 1 AND 2 I
'.\\',
SURRY POWER STATION ~
a QIJ LEGEND:
.. NEW STEAM GENERATOR
< OLD STEAM GENERATOR
. STEAM GENERATOR STORAGE FACILITY CONT.
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