ML20053D752
| ML20053D752 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/21/1982 |
| From: | Parker W DUKE POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| REF-SSINS-6820 IEB-79-02, IEB-79-2, NUDOCS 8206070288 | |
| Download: ML20053D752 (19) | |
Text
m-Duxe POWER COMI%NY Powen 13citnswo 422 Socin Cuuacu STurer, CIIAH1DTTE, N. C. asa4a May 21, 1982 wi w 4w o. Pannen.sn.
VIC F PRES +0fMT TELEPMONE; Amen 704 STEAM PAODucitOM 373-4083 Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Fbrietta Street, Suite 3100 Atlanta, Georgia 30303 Re: RII:JP0 McGuire Nuclear Station Docket Nos. 50-369 and 50-370 IE Bulletin 79-02
Dear Mr. O'Reilly:
Attached is Revision 5 of Duke Power Company's response to IE Bulletin 79-02.
Note that only pages 1 & 5 of the response and pages 14, 28-30, and 46-51 of the anchor bolt safety factor analysis have changed. These revised pages should be inserted in the previous submittal dated December 8, 1980.(Revision 4). With the submittal of this information, all known outstanding items on Unit 2 have been resolved.
Should you have additional questions regarding this matter, please advise.
V truly yours, b,
u>
William O. Parker, Jr.
PBN/j fw Attachment cc: Director Mr. P. R. Bemis
- Division of Reactor Operations Inspection Senior Resident Inspector-NRC Office of Inspection and Enforcement McGuire Nuclear Station U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Director Mr. W. P. Ang-Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Region II Washington, D. C.
20555 101 Marietta Street, Suite 3100 Atlanta, Georgia' 30303 8206070298 820521 PDR ADOCK 05000369 0
((/l
y y r g m J: 5 MCGUIRE NUCLEAR STATION Responses to USNRC IE Bulletin 79-02, Revision 5 Original:
July 2, 1979 Revision 1:
January 7, 1980 Revision 2:
July 24, 1980 Revision 3:
December 1,1980 Revision 4:
December 8, 1980 Revision 5:
May 14, 1982 McGuire Nuclear Station is in the later stages of construction of Unit #2 with commercial operation of Unit #1.
All pipe supports have been erected in Unit #1 and a large number have been erected in Unit #2.
The following is a summary, by item, of the extent and manner in which Duke Power Company intends to satisfy Actions 1 through 9 of the IE Bulletin 79-02, Revision 2.
Response 1:
Duke Power Company will account 1or base plate flexibility in the calculation of expansion anchor bolt loads for all Seismic Category I pipe support base plates using either a conservative hand calculation method which has been verified by non-linear finite element analysis or a specific non-linear finite element analysis for a particular base plate.
The models and boundary conditions, including appropriate load displacement character-istics of the anchors, used for the finite element analyses, are based on Duke studies and on work performed by Teledyne Engineering Services which was sponsored by a group cf thirteen (13) utilities formed to respond to generic items of IE Bulletit. 79-02.
All l
expansion anchor support plates designed prior to implementing these analysis methods have been reanalyzed accordingly and have been modified if required to comply with allowable anchor bolt loadings.
Response 2:
The minimum factors of safety, between the expansion anchor bolt design load and the bolt ultimate capacity determined from static load test, used in Duke's design of pipe supports, are as follows:
Normal Conditions
-4 Upset Conditions
-4 Faulted Conditions *
-4 These factors of safety are for wedge type and sleeve type expanison anchors.
Some shell type anchors were used in the early stages of McGuire construction.
Use of shell type anchors for Nuclear Safety Related applications was discontinued in February 1975.
Duke Power Company has identified all pipe supports using shell type anchors and the design of these supports has been reviewed to assure that a minimum factor of safety of five (5) is maintained.
" *For work completed prior to November 1,1980, the 95% confidence level criteria, presented in the statistical analysis enclosed with this response, applies to faulted loading combinations.
The design criteria for all designs after that date require a minimum safety factor of 4 for faulted conditions.
Revision 5
,3,
p.
(- ympymy;4g concluded that this test sample provides reasonable and adequate assurance of proper plate bolt hole size for wedge and sleeve type expansion anchors.
The supplemental Self-Drill Inspection Program implemented under MCS-1196.02-00-0003 identified 55 of 191 plate holes inspected as oversized.
This oversizing was determined to be due to the self-drill anchor installation procedures. All oversized holes have been reviewed and modification made where required.
In order to address the question of the relationship of cyclic /
load carrying capacity to installation procedure (anchor pre-load), the tests referred to in Response 3, performed by Teledyne Engineering Services and sponsored by the group of thirteen (13) utilities, have been performed on anchors installed in accordance with manufacturer's recommended installation procedures and have no more preload than is provided by the use of these procedures.
Based on Duke's understanding of the behavior of expansion anchors and on cyclic testing which has been performed, Duke Power Company is confident that the anchors will perform adequately.
Some pipe supports with anchor bolts were physically inaccessible for inspection under the provisions of this bulletin.
These have been independently assessed to verify that they are not reasonably accessible.
Duke has concluded that there is reasonable and adequate assurance that these supports will perform adequately.
This conclusion is based on the results of inspections of other supports and the long history of documentation associated with pipe supports.
Specifically, 115 supports with and without anchor bolts were installed under Construction Procedure 308 and inspected under Construction Procedure 503 or QA Procedure M-52.
It is Duke's position that this documentation assures the proper installation of the anchor bolts on inaccessible supports and satisfies the requirements of IEB 79-02.
Response 5:
Nuclear Safety Related/ seismic pipe supports are prohibited from being attached to block (masonry) walls using concrete expansion anchors.
In response to Revision 2 of IE Bulletin 79-02, Duke Power Company has conducted an on-site confirmatory review at McGuire Units #1 and #2 of Nuclear Safety Related/ seismic pipe supports to assure that ao such installations exist.
Results of this review have confirmed that there are no such installations i
of this type at McGuire Nuclear Station Units #1 and #2.
Response 6:
The expansion anchor installation and inspection procedures utilized at McGuire Nuclear Station and described in Response 4 apply to all expansion anchors installed in Nuclear Safety Related pipe supports.
Each expansion anchor is inspected regardless of the physical configuration of the steel members
. Revision 5
______._.....____/N.___
--,ns,_
i Duke Power Company McGUIRE NUCLEAR STATION ANCHOR BOLT SAFETY FACTOR ANALYSIS FOR USNRC I & E BULLETIN 79-02 eaw I
DUKE POWER l
mQ AUGUST 8,1980 REVISED DECEMBER I,1980 REVISED DECEMBER 8,1980 REVISED MAY I4,1982 i
_ j
MqqqgqQniqqy;
.- m :,, -
1
%; 1 3
8.0 CASE-BY-CASE REVIEW 0F SUP' ORTS / RESTRAINTS-P McGuire Support / Restraint Design Group personnel conducted several reviews of supports with anchor bolts to verify compliance with IE Bulletin 79-02.
Prior to receipt of the Bulletin, base plate flexibility was not considered in the design.
Supports at McGuire are generally classed as rigorous or alternate depending on the type of piping analysis employed.
Alternate analysis supports comprise approximately 6,513 of 15,276 safety re-lated supports in Unit I and 4,686 of 9,817 in Unit 2.
A screening of these alternate supports identified 549 which potentially would not have the Bulletin design safety factor in the normal / upset load combination if base plate flexibility were considered in Unit 1 and none in Unit 2.
A detailed review of each of the support / restraints was conducted accouting for base plate flexibility.
Modifications were specified as required to upgrade the support to meet this design safety factor criteria.
This work has been completed.
Rigorous analysis supports comprise anproximately 8,763 or 15,276 safety re-lated supports in Unit 1 and 5,131 of 9,817 in Unit 2.
These were also screened to account for base plate f,lexibility consideratinns, and 1,952 supports potentially did not have the required design safety factor in Unit I and 46 in Unit 2.
A detailed review of each of these support / restraints was conducted accounting for base plate flexibility. Modifications were specified as required to upgrade the suppart/ restraint to meet this design safety factor criteria.
This work has been ccmpleted.
In response to concerns expressed by V. S. Nuclear Regulatory Commission Region II inspectors regarding the McGuire design criteria of FSp = 2.125, McGuire Support / Restraint Design Group reviewed a sample of 1,410 rigorous supports to estimate the overall impact of upgrading Unit 1 to a faulted safety factor of 4, holding all else constant.
The importance of concrete strength was examined by reviewing a 3000 psi nominal strength case and a 5000 psi nominal strangth case.
[ presents the projections derived from this analysis.
Note that 604 rigorous and 98 alternate supports, in Unit 1, were identified as havina a strong potential for not complying with a faulted condition safety factar of 4 (5000 psi concrete).
A detailed review of standard design practices at Revision 5
%S112i49 TdC7j The minimum sample size required in using a binomial distribution assumption is 74.
Large sample sizes are permitted and have correspondingly larger allowabic number of anchor bolts with FS(4 to satisfy the hypothesis.
These values are readily obtainable from a table of confidence limits for propr-
/
tions applicable to Binomial, Poisson, and Hypargeometric Distributions.
The Binomial distribution function and inverse sampling tedr[iques assume a large population (N) compared with the sample, (n).
Some systems have fewer anchor bolts than the minimum sample size and will-he analyzed absolutcly, i.e., the total population (fi) analyzed.
11.4 RESUUS The results for the 39 Nuclear Safety Related systems are presented in Attach-ment 13.
11.4.1 Unit 1 1
All rigorousiv analyzed systems passed the acceptance criteria. Alternate Analysis systems which did not meet the criteria are NF (ice condenser refrigeration), VE (annulus ventilation), VG (diesel generator starting air), WS (solid waste), WZ (ground water drainage), and YM (demineralized water).
11.4.2
_ Unit 2 The rigorously analyzed systems which did not meet the criteria is II (Incore Instrumentation).
There was only one support in the system utilizing anchor bolts and it was already identified for redesign due to changes in the design loads. All Alternate Analysis systems passed the acceptance criteria.
The small number of systems which did not meet the criteria and the fact that time and cost trade-offs did not permit analytical credit for the full range of conservatisms existing in the original design demonstrate clearly that evision 5.
~
L.C N' l:C:221
- 1cfl2k'Lli.
anchor bolt safety factors are generally far in excess of that required by the plant design.riteria employed for ficGuire Nuclear Station.
Many of the anchor bolts in the sample which did not have a minimum safety factor of 4 had calculated safety factors of 3.5 or greater.
As demonstrated by Attachment 6, there are no anchor bol'ts with a safety factor less than 2.4.
1 1
1
(
l
/
1 Revision 5 (Entire Page Revised)
d.c22fCC.iC3 N 2% 6 11.5 IMPACT ON FUEL LOADING AND FULL POWER OPERATION Based on Duke's commitment in Revision 4 to the IEB 79-02 response, there is no impact of completing repairs on fuel loading or full power operation.
All repairs required to meet the requirements of IEB 79-02 were completed prior to fuel loading in Unit 1 and are completed in Unit 2.
11.6 PLAN OF CORRECTIVE ACTION All anchor bolts on each of the systems which did not meet the criteria have been screened, analyzed in detail, and repaired as necessary to achieve a condition which meets the acceptance criteria, i.e. no more than 5% of the anchor bolts on that system with a safety factor less than 4.
This program was similar to the hanger baseplate review which was done to consider baseplate flexibility.
Also, we expeditiously corrected all bolts identified as having a safety factor less than 4.
A summary of these activities and the status is presented below.
ACTIVITY STATUS i
Unit 1 l
Upgrade to flect 95% C.L. - Systems NF, VE, VG Complete WS, WZ & YM Correct Bolts Identified v/ FS<4 Comple te Unit 2 Upgrade to Meet 95% C.L.
- System II Complete Correct Bolts identified w/ FS(4 Complete Revision 5 (Entire Page Revised) 4 l J
. r - a <.- ~ s. c.
-... o;; y 3qaw y r g g;q ATTACHME!!T 13 Unit 1 Rigorous Systems At.tu al Allowed #
Hypothesis System flame w/FS(4 w/FS<4 Accepted
- l BB Steam Generator Blowdown 2
4 Yes l
CA Auxiliary Feedwater 3
3 Yes
(
0 Yes FW Refueling Water 0
0 Yes II Incore Instrumentation 0
0 Yes KC Component Cool 4
16 Yes KD Diesel Genera tor Engine Cooling 0
0 Yes Water KF Spent Fuel Cooling 2
2 Yes LD Diesel Generator Engine Lubricating 0 0
Yes Oil NB Boron Recycle 0
7 Yes NC Reactor Coolant 0
2 Yes ND Residual Heat Removal 0
0 Yes NI Safety Injection 12 12 Yes NM Nuclear Sampling 0
0 Yes NR Boron Thermal Receneration 0
0 Yes NS Containment Spray 0
0 Yes Rtvision 5
f-MX5Kiku"T*2y's."
~ ~ ~ ',
'~
ATTACHf1Ef4T 13 (Cont'd)
Unit 1 Rigorous Systems (Cont'd)
Actual Allowed #
Hypothesis System flame w/FS(4 w/FS<4 Accepted
- flV Chemical & Volume Control 0
0 Yes Rf1 fluclear Service Water 4
6 Yes l
RV Containment Ventilation Cooling 1
9 Yes Water SA Auxiliary Steam 0
0 Yes SM Main Steam 0
0 Yes SV Main Steam Vent 0
0 Yes VQ Containment Pressure Centrol 0
0 Yes VX Containment Air Return Exchanqe 0
0 Yes WL Liquid Waste Recycle 2
5 Yes YC Chilled Water 0
36 Yes e Revision 5 m
~'
^
3 m:5L M 2 M n
e ATTACHMENT 13 (Cont'd)
Unit 1 Alternate Analysis Systems Actual Allowed #
Hypothesis System Name w/FS<4 w/FS44 Accepted
- FD Diesel Generator Engine Fuel Oil 0
0 Yes NF Ice Condenser Refrigeration 13 1
No RF Fire Protection 0
0 Yes VB Breathing Air 1
1 Yes VE Annulus Ventilation 6
0 No VG Diesel Generator 16 4
No VI Instrument Air 0
0 Yes VS Station Air 0
2 Yes WE Equiprhent Decontamination 0
1 Yes WG Waste Gas 0
0 Yes WS Nuclear Solid Waste Disposal 8
0 No WZ Groundwater Drainage 40 1
No YM Demineralized Water 6
1 No j
. Revision 5
v.< gyg{gg},gy 7
,~..
' t. '.
y ATTACHMElli 13 UNIT 2 Rigorous Systems Ac tual Allowed #
Hypothesi System flame w/FS(4 w/FS<4 Accepted BB Steam Generator Blowdown 0
0 Yes CA Auxiliary Feedwater 0
0 Yes FW Refueling Water 0
0 Yes II Incore Instrumentation 8
0 No KC Component Cool 0
0 Yes KD Diesel Generator Engine Cooling 0
0 Yes Water KF Spent Fuel Cooling 0
0 Yes LD Diesel Generator Engine 0
0 Yes Lubricating Oil flB Boron Recycle 0
0 Yes flC Reactor Coolant 0
0 Yes ND Residual Heat Removal 0
0 Yes til Safety Injection 0
0 Yes l
l NM fluclear Sampling 0
0 Yes fir Boron Thennal Regeneration 0
0 Yes flS Containment Spray 0
0 Yes
. Revision 5 (New Page)
.,_____________________.___.__________.________.___..__.___.___________.._____m_..__
M '7 N ATTACNMENT 13 (cont'd)
UNIT 2 Rigorous Systems (cont'd) l l
l Actual Allowed #
Hypothesi System Name 4/FS<4 w/FS(4 Accepted NV Chemical & Volume Control 0
0 Yes RN Nuclear Servic Water 0
0 Yes RV Containment Ver, lation Cooling 0
0 Yes Water SA Auxiliary Steam
~
0 0
Yes SM Main Steam 0
0 Yes l
SV Main Steam Vent 0
0 Yes I
VQ Containment Pressure Control 0
1 Yes VX Containment Air Return Exchange 0
0 Yes WE Equipment Decontamination 0
0 Yes WL Liquid Waste Recycle 0
0 Yes e Revision 5 (New Page)
nWE9 WON $WW$12M ATTACHMENT 13 (Cont'd)
UNIT 2 Alternate Analysis Systems 1
Ac tual Allowed #
Hypothesi!
System Name w/FS<4 w/ FS< 4 Accepted' FD Diesel Generator Engine Fuel Oil 0
0 Yes NF Ice Condenser Refrigeration 0
0 Yes RF Fire Protection 0
0 Yes VB Breathing Air 0
0 Yes VE Annulus Ventilation 0
0 Yes VG Diesel Generator 0
0 Yes VI Instrument Air 0
0 Yes VS Station Air 0
0 Yes WG Waste Gas (Included in Unit 1)
WS Nuclear Solid Waste Disposal (Included in Unit 1)
WZ Groundwater Drainage 0
0 Yes YC Chilled Water 0
0 Yes YM Demineralized Water 0
0 Yes s
- The acceptance criteria requires a 95% confidence level that no more than 5% of the anchor bolts on a system have a safety factor less than 4.
. Revision (New Page)5
_ _.. _ _. _. _ _ _ _ _ _ _. _ - -. - - -