ML19207B998
| ML19207B998 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/25/1979 |
| From: | Gray J NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Mccollom K, Mark Miller, Paxton H AFFILIATION NOT ASSIGNED, Atomic Safety and Licensing Board Panel, OKLAHOMA STATE UNIV., STILLWATER, OK |
| Shared Package | |
| ML19207B999 | List: |
| References | |
| NUDOCS 7909060290 | |
| Download: ML19207B998 (3) | |
Text
-
M NRC PUBLIC Doce,ug;g,
'9, UfJITED STATES S%..
NUCLEAR REGULATORY COMMISSION
$fn'1 hcl'%
E WASHINGTON, D. C. 20555
- c..<.w h / l 9
w4; o
o July 25, 1979 Marshall E. Miller, Esq., Chairman Dr. Kenneth A. McCollom, Dean Atomic Safety and Licensing Board Division of Engineering, U.S. Nuclear Regulatory Commission Architecture & Technology Washingon, D.C.
20555 Oklahoma State University Stillwater, Oklahoma 74074 Dr. Hugh C. Paxton g i m,, %
1229 41st Street
\\
Los Alamos, New Mexico 87544
[\\
M
['f Q
~
s "t f k $.'
A 3d In the Matter of
~'
/[7 5)
Portland General Electric Company, et al.
a (Trojan Nuclear Plant)
D.
- M [
Docket No. 50-344 d
c*
' (Control Building) w ; --.N Gentlemen:
Enclosed for your information is IE Bulletin 79-14, issued on July 2, 1979, and Revision 1 to that Bulletin, issued on July 18, 1979.*
This Bulletin, as re-vised, deals with the accuracy of information used for the seismic analyses of safety-related piping and piping systems at nuclear power plants licensed by the NRC.
h" nile the information sought by the Bulletin is not related to the Trojan Control Building design deficiencies, it may have some relevance to Phase II of the referenced proceeding as indicated more fully below.
In accordance with the guidelines set forth in Vircinia Electric & Power Company (North Anna Nuclear Power Station, Units 152), ALAB-551, Slip Op. pp. 10, 11 (June 26, 1979), the following should be noted with regard to this board noti-fication matter:
(1)
IE Bulletin 79-14, as revised, is self-explanatory as to the precise nature of this board notification matter.
- Briefly, problems have been discovered at a number of operating plants in which the currently existing as-built configurations of safety-related piping systems differ from the original de-sign documents.
As a res. ult, the configurations assumed for the original seismic analyses based on design documents at the licensing stages for these facilities could differ significantly from, and be non-conservative relative to, the actual as-built facility. Accordingly, IE Bulletin 79-14, as revised, requires licensees and permittees to inspect safety-related piping systems, identify any instances in which currently existing as-built piping systems do not r.
~.
W,..P (gh
- A corrected version of Revision 1, dated July 19, 1979, is also enclosed.
19 090 6 u 270 P conform to seismic analysis input information set forth in design documents, evaluate non-conformances, and take corrective action.
(2) This board notification matter applies to all licensed power reactors, including Trojan.
(3) As to the relevance of this board notification matter to the issues remaining before the Licensing Board in the referenced proceeding, the general issues remaining betore the Board are whether the proposed Control Building modifications are ade-quate from a safety standpoint and whether they may be per-formed while the facility operates (The matter of interim operation is no longer before the Licensing Board since the decision on inters operation is now final agency action, the Appeal Board having affirmed that decision and the Commission having determined not to review it within the prescribed time period.
The question as to whether the seismic qualification analyses for interim operation are impacted in any way by the matters raised in IE Bulletin 79-14, as revised, is being actively pursued, and will be resolved, by the Staff).
The determination ns to tae adequacy of the proposed modifications and as to the acceptability of operation during the modifi-cation work is dependent, in part, upon an accurate analysis of the seismic qualification of safety-related piping systems in the Control Building complex both during the period when modification work is being performed and after the modificatiens are in-place.
Such an analysis, to be accurate and fully appli-cable and valid, must be substantially based on the actual, as-built configuration of safety-related piping systems in the
~
Control Duilding Complex.
(4)
Because of the potential relevance of this Bulletin to Phase II, as set forth above, the Licensee was requested by the Staff in a telephone communication on July 12, 1979, to pro-vide a preliminary assessment of the impact of this Bulletin on the Control Building Complex on an expedited basis. The Staff was aware of the fact that extensive investigation and reanalyses of safety-related piping systems in the Control Building Complex have been performed and that much of this work was based on as-built piping systems in the Complex.
Enclosed is the Licensee's July 17, 1979 letter to R. H.
Engelkin, Director of Region V of the NRC's Office of Inspec-tion and Enforcemcnt confirming that:
a)
Licensee investigations in response to IE Bulletins 79-02, 79-04 and 79-07 have revealed no significant discrepancies between design and as-built piping con-figurations; b)
Extensive reevaluation and modifications to piping in
- g.,t--u.,p.
the Complex in conjunction with the Control Building
'} A
- l b I t '
. proceeding have identified no as-built conditions which affect the operability of safety-related piping; c)
An inspection of safety-related piping in the as-built condition in conjunction with resolution of pipe-pene-tration grouting anocolics reported to the Licensing Board at the Prehearing Conference in this proceeding on August 14, 1978 identified no as-built conditions which affect operability of safety-related piping; d)
Prior to initial facility operation, as-built configu-rations of safety-related piping were confirmed to be consistent with design and engineering evaluations of that piping and any discrepancies were evaluated and resolved; e)
As-built drawings incorporating any changes to piping systems are maintained by the Licensee and any differences between as-built conditions and design assumptions are reviewed and resolved.
In view of all of this, the Staff believes that any significant differences between design analyses and as-built configurations of safety-related piping in the Control Building Complex would be promptly identified and resolved.
In addition, past investigations and existing design change rechanisme give reasonable assurance at this time that no design /as-built discrepancies exist for the Control Building Comolex that would adversely affect the public health and safety in the event of an earthquake.
Portions of the results of the Licensee's inspection conducted pursuant to IE Bulletin 79-14 are to be submitted within 60 and 120 days of the date of the Bulletin.
In the event that significant discrepancies between design and as-built safety-related piping configurations in the Control Building Complex are discovered, the Staff will promptly take any actions found necessary to assure facility safety and inform the Licensing Board.
Sincerel,
f n,-
=
K.u?-l')b*W/!M
[osep'hR. Gray Counsel for NRC Staff
Enclosures:
As stated cc w/ enclosures:
Service List
- r. q, - < * -w -
%f.% e E
- 9 $ fL4
, t ' *
- t o u,1 s
UNITED STATES d
[ T['p,.. f,i NUCLE AR REGULATORY COMMISSION 3.aAi? ~,' 9)
C nECICN V S /eW h 1993 N. CAllFOR*:l A COULEVARD SUITC 202, WAlf.U T CRE E K PLAZ A
't gJ 9
WALNUT CREE K, CALIFCRNI A 94596 July 2,1979 Docket tio. 50-344 Portland General Electric Ccapany 121 S. W. Salmon Street Portland, Oregon 97204 Attention:
Mr. Charles Goodwin Assistant Vice President Gentlemen:
The enclosed Bulletin 79-14 is forwarded to you for action.
Written respsnses are required.
If you desire additioniti information regarding this matter, please contact this office.
Sincerely, n(Z75,~.40.-
R. H. Engelken Director
Enclosures:
1.
IE Bulletin !!c. 79-14 with Appendix A 2.
Listing of IE Bulletins Issued in Last 12 Months cc w/ enclosures:
B. Withers, PGE F. Gaidos, PGE c.
r - c...,,,
s]>. s) % / :
Ut1ITED STATES fiUCLEAR REGULATORY COMMISSIOil WASHINGTO1, D.C.
20555 July 2,1979 IE Bulletin fio. 79-14
~
SEISMIC A!iALYSES FOR AS-BUILT SAFETY-RELATED PIPIrlG SYSTEMS Description of Circumstances:
Recently two issues were identified which can cause seismic analysis of safety-related piping systems to yield nonconservative results.
One issue involved algebraic summation of loads in some seismic analyses.
This was addressed in show cause orders for Beaver Valley, Fitzpatrici:, Maine Yankee and Surry.
It was also addressed in IE Bulletin 79-07 which was sent to all power reactor licensees.
The other issue involves the accuracy of the information input for seismic analyses.
In this regard, several potentially,unconservative factors were discovered and subsequently add: essed in IE Bulletin 79-02 (pipe supports) and 79-04 (valve weights).
During resolution of these concerns, inspection by IE and by licensees of the as-built configuration of several piping systems revealed a number of nonconformances to design documents which could potentially affect the validity of seismic analyses.
"Onconformances are identified in Appendix A to this bulletin.
Because apparently significant non-conformances to design documents have occurred in a number of plants, this issue is generic.
The staff has determined, where design specifications and drawings are used to obtain input information for seismic analysis of safety-related piping systems, that it is essential for these documents to reflect as-built con-figura tions. Where subsequent use, damage or modifications affect the con-dition or configuration of safety-related piping systems as described in documents from which seismic an("si-input information was cbtained, the licensee must consider the need
- e-eva'luate the seismic analyses to con-sider the as-built configuration.
IE Bulletin No. 79-14 July 2, 1979 Page 2 of 3 Action to be taken by Licensees and Permit Holders:
All power reactor facility licensees and construction permit holders are requested.to verify, unless varified to an equivalent degree within the last 12 months, that the seismic analysis applies to the actual configura-tion of safety-related piping systems.
The safety related piping includes Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Design Classification" P.evision 1, dated August 1,1973 or as defined in the applicable FSAR.
For older plants, where Seismic Category I require-ments did not exist at the time of licensing, it must be shown that the actual configuration of these safety-related systems meets dasign require-ments.
Specifically, each licensee is requested to:
1.
Identify inspection elements to be used in verifying that the seismic analysis input information conforms to the actual configuration of safety-related systems.
For each safety-related systen, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic analyses.
Also submit a descriptien cf the seismic analysis input information which is contained in each document.
. Identify syste7s or partions of systems which are planned to be inspected during each sequential inspection identified in Itams 2 and 3. Subnit all of this ir. formation within 30 days of the date of this bulletin.
2.
For portions of systems which are nornally accessible *, inspect one system in each set of redundant systems and all nonredundant systems for con-formance to the seismic analysis input infornation set forth in design documents.
Include in the inspection: pipe run gecmetry; suoport and restraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachements; and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).
Within 60 days of the date of this bulletin, submit a description of the results of this inspection.
Where nonconformances are found which affect operability of any system, the licensee will expedite ccmpletion of the inspection described ic Item 3.
- Normally accessible refers to those areas of the plant which can be entered during reactor operation.
9 (J.j ? *TWQ s.* A s ) *.* I LI
IE Bulletin !!o. 79-14 July 2, 1979 Page 3 of 3 3.
In accordance with Item 2, inspect all other normally accessible safety-related systems and all normally inaccessible safety-related systems.
.Within 120 days of the date of this bulletin, submit a description of
. the results of this inspection.
4 If nonconformances are identified:
A.
Evaluate the effect of the nonconfomance upon system operability under specified earthquake loadings and ccmply with applicable action statements in your technical specifications including prcmpt report-ing.
B.
Submit an evaluation of identified nonconformances on the validity of piping and support analyses as described in the Final Safety Analysis Report (FSAR) or other NRC approved documents. Uhere you determine that reanalysis is necessary, submit your schedule for: (i) completing the reanalysis, (ii) ccmcarisons of the results to FSAR or other I!RC approved acceptance criteria and (iii) submitting descrip-tions of the results of reanalysis.
C.
In lieu of 8, submit a schedule for correcting nonconfoming systems so that they confom to the design documents. Also subnit a descrip-tion of the work required to establish'confomance.
D.
Revise dccuments to reflect the as-built conditions in plant, and describe measures which are in effect which crovide assurance that future ecdifications of piping systems, including their supports, will be reflected in a timely manner in design documents and the seismic analysis.
Facilities holding a construction pemit shall inspect safety-related systems in accordance with It2ms 2 and 3 and report the results within 120 days.
Reports shall be submitted to the Regional Director with copie; to the Director of the Office of Inspection and Enforcement and the Director of the Division of Operating Reacturs, Office of fluclear Reactor Regulation, Washington, D.C.
20555.
Approved by GA0 (R0072); clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for generic problems.
kht f un)I S UNSO l-I
APPEllDIX A PLAliTS WITH SIGilIFICA'!T DIFFERE! ICES BETWEEll ORIGINAL DESIGN AND AS-BUILT C0liDITION OF PIPIriG SYSTEMS Plant Difference Remarks Surry 1 Mislocated supports.
As built condition Wrong Support Type.
caused majority of pipe Different Pipe Run
.overstress problems, not Geometry.
algebraic summation.
Beaver Valley liot specifically identified.
As built condition resulted Licensee reported "as-bui'.t in both pipe and support conditions differ signifi-overstress.
cantly from orginal design."
Fitzpatrick IE inspection identified Licensee is using as differences similar to built configuration
- Surry, for reanalysis.
Pilgrim Snubber sizing wrong.
Plant shutdown to restore Snubber pipe attachment origiral design conditicn.
i welds and snubber support assembly ncnconformances.
Brunswick 1 and 2 Pipe supports undersize.
Both units shutdown to restore original design condi tion.
Ginna Pipe supports not built Supports were repaired to original design.
during refueling outage.
15t. Lucie Missing seismic supports.
Install corrected Supports on wrong piping.
supports before start up frcm rerueling.
11ine Mile Point Missing seismic supports.
. Installed supports before startup from refueling.
Indian Point 3 Support location and Licensee performing as suppart construction built verification to be deviations.
completed by July 1.
Davis-Besse Gussets missing from main Supports would be over-Steam Line Supports.
strassed.
Repairs will be completed prior to start-up.
Sj.,' M8_I.
IE Bulletin fio. 79-14 Enclosure July 2,.1979 Page 1 of 4 LISTI!;G CF IE BULLETIIiS ISSUED lti LAST TWELVE V.0MTHS Bulletin Subject Date Issued Issued To No.
79-13 Cracking In Feedwater 6/25/79 All PWRs with an System Piping OL for action. All BWRs with a CP for i nformation.
79-02 Pipe Support Base Plate 6/21/79 All Power Reactor (Rev. 1)
Designs Using Concrete Facilities with an Expansicn Anchor Bolts OL or a CP 79-12 Short Period Scrams at 5/31/79 All GE BUR Facilities BWR Facilities with an CL 79-11 Faulty Overcurrent Trip 5/22/79 All Pcwer Reactor Device in Circui*. Breaners Facilities with cn for Engineered Safety OL or a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Stacistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Pcwer Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-03 Events Relevant to BWR 4/14/79 All BWR Power Reactcr Reactors Identified During Facilities with an OL Three Mile Island incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 3)UC32
Encicsure
'IE Bulletin fio. 79-14 Page 2 of 4 July 2, 1979 LISTING 0F IE BULLETItiS ISSUED Ill LAST TWELVE MC'iTHS Bulletin Subject
_ece Issued Issued To No.79-06B Review of Operational 4/14/79 All Cctbustion Engineer-Errors and System Mis-1ng Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Inciernt Operating Licensee 79-06A Review of Operational 4/18i79 As. Pressurized Water (Rev 1, Errors and System Mis-Power Reactor Facilities alionments Identified of Westinghouse Design During the Three Mile with an OL Island Ircident 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile ~
with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an alignments Identified OL except C&W facilities During the Three Mile Island Incident 79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three tiile Island Facilities with an CL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three 11ile Island Facilities with an CL and CP 935283
IE Bulletin No. 79-14 Enclosure July 2, 1979 Page 3 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS.
Bulletin Subject Date Issued Issued To No.
79-04 Incorrect Weights for 3/30/79 All Pswer Reactor
~~
Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 78-123 Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Strel Pipe Spcois OL or CP Manufactured by Youngs: cwa Welding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/79 All Power Reactor Designs Using Concre:e Facilities with an Expansion Anchor Bolts OL cr CP 79-01A Environmental Qualification 6/6/79 All Pcwer Reactor of Class IE Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP 78-14 Deterioration of Buna-N 12/19/78 All GE EWR facilities Ccmponent In ASCO with an CL or CP Solenoids 4
IE Bulletin No. 79-14 Enclosure July 2, 1979 Page 4 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE tt0NTHS Bulletin Subject Date Issued Issued To No.
78-13 Failures in Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges specific licensees Models 7050, 70503, 7051, with the subject 7051B, 7060, 70603, 7051 Kay-Ray, Inc.
and 7061B gauges78-12A Atypicai Meld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-11 Exanination of Mark I 7/21/78 BWR Pcwer Reactor Containment Torus Welds Facilities for action:
Peach Ect:cm 2 and 3, Quad Cities 1 and 2, Hatch 1, "anticello and Vermont Yankee e
Y
{ ', " g II a y e P.w. :
.f'p a n i c,%,
o U.*JITED S TATEs
? t,. 7 NUCLEAR REGULATORY COT.1T,1!SSION
^-
d.,%d,:,d,[l_ n E
RECIOil V j
1990 rJ. CAllFOGr;I A COUt.CVAHD k 'g,.j,_ f.,
o
'i!/ *I SUIT E 702, WAL.' JUT CRE EK PLAZA
~7.,,,, *[
WALNUT CREEK CALIFORNIA 94505 July 18, 1979 Docket f!o. 50-344 Portland General Electric Ccmpany 121 S. W. Salmon Street Portland, Oregon 97204
' Atten tion: Mr. Charles Goodwin Assistant Vice President Gentlemen:
IE Bulletin !!o. 79-14 is revised to limit the scope of work required.
The changes are indicated on the enclosed replacement page for the bulletin.
If you desire additional informaticn regarding this matter, please contact this office.
Sincerely,
/f e
/
a /
I[.\\,6 299", g[ g
.:.. Engelecen Director
Enclosure:
IE Bulletin ?!o. 79-14, Revision 1 cc w/ enclosure:
B. Withers, PGE F. C. Gaidos, PGE S
6
\\
8 i
i
?
e f, 4
9f 6
d j.34 0]_',
j 6
W IE Bulletin No. 79-14 July 18,1979 Revision 1 Page 2 of 3 Action to be taken by Licensees and Permit Holders:
All power reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree within the last 12 months, that the seismic analysis applies to the acutal configura-tion of safety-related piping systems.
The safety related piping includes Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Design Classification" Revision 1, dated August 1,1973 or as defined in the applicable FSAR. The action items that follow acoly to all safety related pioina 21/2-inches in diameter and greater and to seismic Category I ciping, regardless of size wnich was dynamically analyzed by comouter.
For older plants, unere Seismic Category I requirements did not exist at the time of licensing, it must be shown that the actual configuration of MLf safety-related systems, utilizing oiping 21/2 inches in diameter and greater, meets design requirements.
Specifically, each licensee is requested to:
1.
Identify inspection elements to be used in verifying that the seismic analysis input information conforms to the actual configuration of safety-related systems.
For each safety-related system, submit a list of desien documents, including title, identification number, revisicn, and date, ~
which were sources of input information for the seismic analyses. Also contained in each dccument.
Identify systems or portions of systems which are planned to be inspected during each sequential inspection identified in Items 2 and 3.
Submit all of this information with 30 days of the date of this bulletin.
2.
For portions of systems which are normally accessible *, inspect one system in each set of redundant systems and all r.onredundant systems Tor con-
.formance to the seismic analysis input information set forth in design documents.
Include in the inspection; pipe run gecmetry; support and restraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachments; and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).
Within C0 days of the date of this bulletin, submit a' description of the results of this inspection.
Where nonconformances are found which affect operability of any system, the licensee will expedite completion of the inspection described in~ Item 3.
- Normally accessible refers to those areas of the plant which can be entered
'during reactor operation.
o,,
syp
.y.r..
!p-UNITED sT ATEs s u o,
[ fp, NUCLEAR REGULATORY COMMISSION
- J,[,M,7,g g[ )j REGIOrj V g.gf.j 0, la t.4.
- .4 / 8 1930 TJ. CAllFORr;I A COULC'/ARD N[
5tJITC 202 c/A LNUT CT< C E K PLAZ A
'/
WALNUT CP.EE K. CA LIF ORNI A 94536
- e,,*
July 19, 1979 Docket tio. 50-344 Portland General Electric Ccapany 121 S. U. Salmon Street Portland, Oregon 97204 Attention:
f1r. Charles Goodwin Assistant Vice President Gentlemen:
Attached is a corrected copy of Revision 1 to IE Bulletin ilo. 79-14 mailed to you on July IS,1979.
A sentence was inadvertently omitted from the previous copy.
Sincerely, M
f(fl d-~
' ~
R. H. Engelken Director
Enclosure:
IE Bulletin ilo. 79-14, Revisicn 1 - Corrected Copy cc w/ enclosure:
B. Withers, PGE F. C. Gaidos, PGE
I IE Bulletin No. 79-14 July 18, 1979 Revision 1 Page 2 of 3 Action to be taken by Licensees and Permit Holders:
All pcwer reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree within the last 12 months, that the seismic analysis applies to the actual configura-tion of safety-related piping systems.
The safety related piping includes Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Design Classification" Revision 1, dated August 1,1973 or as defined in the applicable FSAR.
The action items that folicw acoly to all safety related pioina 21/2-inches in diaceter and creater and to seismic Category I cioina, regardiess of size which was dynamically analyzed by comouter. For older plants, wnere Seismic Category I requirements did not exist at the time of licensing, it must be shown that the actual configuration of fMM safety-related systems, utilizina oicing 21/2 inches in diameter and greater, meets design requirements.
Specifically, each licensee is requested to:
1.
Identify inspection elements to be used in verifying that the seismic analysis input information conforms to the actual configuration of safety-related systems.
For each safety-related system, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic ar.alyses.
Also O
submit a descriotion of the seismic ar.alysis input information which is 4
contained in each dccument.
Identify systems or portions of systems which are planned to be inspected during each sequential inspection identified in Items 2 and 3. Submit all of this infccmation within 30 days of the date of this bulletin.
2.
For portions of systems which are normally accassible*, inspect one system in each set of redundant systems and all nonredundant systems for con-formance to the seismic analysis input information set forth in design documents.
Include in the inspection: pipe run geometry; support and restraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachements; and valve and valve ccerator locations and weights (excluding those covered in IE Bull'etin 79-04).
Within 60 days of the date of this bulletin, submit a description of the results of this inspection. Where nonconforma c es are fcund which affect operability of any system, the licensee will expedite completion of the inspection described in Item 3.
%crmally accessible refers to those areas of the plant which can be entered during reactor operation.
C0RRECTED C0PY 378M