IR 05000455/1986027: Difference between revisions

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{{Adams
{{Adams
| number = ML20209F053
| number = ML20197H382
| issue date = 09/05/1986
| issue date = 05/12/1986
| title = Insp Rept 50-455/86-27 on 860819 & 20.No Noncompliance Noted.Major Areas Inspected:Procedures,Specs & Results of Facility Structural Integrity Test of Containment Structure
| title = Ack Receipt of 860124 & 0409 Ltrs Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-455/86-27. NRC Concurs w/7 of 10 Items Cited in 860124 Response.Encl 2 Addresses Remaining Items
| author name = Danielson D, Norton F
| author name = Warnick R
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| addressee name =  
| addressee name = Reed C
| addressee affiliation =  
| addressee affiliation = COMMONWEALTH EDISON CO.
| docket = 05000455
| docket = 05000455
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-455-86-27, NUDOCS 8609120033
| document report number = NUDOCS 8605190157
| package number = ML20209F048
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 6
| page count = 5
}}
}}


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U.S. NUCLEAR REGULATORY COMMISSION
MAY 121986
/
[3 I
Docket No. 50-454" Commonwealth Edison Company ATTN:
Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:
Thank you for your letters dated January 24, and April 9,1986, informing us of the steps you have taken in response to the results of the NRC Construction Assessment Team (CAT) inspection forwarded in Inspection Report No. 50-455/85027 by letter dated November 13, 1985 and the Notice of Violation forwarded by our letter dated December 12, 1985.


==REGION III==
Your January 24, 1986 response was forwarded to the Office of Inspection and Enforcement who lead the CAT inspection.
Report No. 50-455/86027(DRS)
Docket No. 50-455 License No. CPPR-131 Licensee:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name:
Byron Station, Unit 2 Inspection At:
Byron Site, Byron, IL Inspection Conducted: August 19 and 20, 1986 Wh e Inspector:
F. Norton
.
Date b
Approved By:
D. H. Danielson, Chief Materials and Processes Section Date Inspection Summary Inspection on August 19 and 20, 1986 (Report No. 50-455/86027(DRS))
Areas Inspected:
Routine announced inspection of procedures, specifications, and results of the Unit 2 structural integrity test of the containment structure.


Results:
Your response has been reviewed and we are in agreement on seven of the ten individual items used as examples of violations. The specific items and our findings are summarized in Enclosure 1.
No items of noncompliance were identified.


8609120033 860705
Our comments concerning the remaining violations are provided in Enclosure 2.
,
PDR ADOCK 0500


.
We will examine these matters during subsequent inspections.
.
DETAILS 1.


Persons Contacted Commonwealth Edison Company (Ceco)
Your cooperation with us is appreciated.
*E. Martin, Quality Assurance Superintendent
*R. Klingler, Project Quality Control Supervisor
*R. Guse, Structural Engineer
*E. Briette, Quality Assurance Engineer U.S. Nuclear Regulatory Commission J. Hinds Jr., Senior Resident Inspector
*P. Brochman, Resident Inspector
* Denotes those who attended the exit interview.


2.
-


Containment Structural Integrity Test - Unit 2 This inspection addressed the structural integrity test performed on the Unit 2 containment structure.
Sincerely,
"Orighrd cit ed hi 2. F. tia nic M R. F. Warnick, Chief Reactor Projects Branch 1 Enclosures: As stated See Attached Distribution RIII RIII RIII R
RIII RIII RI LQ k
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Muff tt rjieIson Pea ison P sePiello Learch/bls


Test procedures, specifications, quality records, and test results were reviewed.
9
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General The Unit 2 structural integrity testing pressurization was commenced at 0800 on May 23, 1985, and depressurization to 0.0 psig was accomplished at 0700 on May 25, 1985. The engineering firm of Wiss, Janney, Elstner Associates, Incorporated (WJE) of Northbrook, Illinois was retained by Commonwealth Edison Company to install the prescribed instrumentation, monitor the response of the instruments, conduct crack surveys prior to and during the testing, and report on the results.
o
-


The location of test instrumentation was planned by Sargent & Lundy Engineers, Chicago, Illinois.
.
Commonwealth Edison


The work was conducted in accordance with Sargent & Lundy Specification No. F/L-2922.
MAy { g jggg Distribution cc w/ enclosure:
D. L. Farrar, Director of Nuclear Licensing V. I. Schlosser, Project Manager Gunner Sorensen, Site Project Superintendent R. E. Querio, Plant Manager DCS/RSB (RIDS)
Licensing Fee Management Branch Resident Inspector, RIII Byron Resident Inspector, RIII Braidwood Phyllis Dunton, Attorney General's Office, Environmental Control Division D. W. Cassel, Jr., Esq.


All installations were performed or supervised by WJE personnel.
Diane Chavez, DAARE/ SAFE Steve Lewis, ELD L. Olshan, NRR LPM H. S. faylor, Quality Assurance Division
 
.
That part of the work normal to their skills (routine installation of electrical lead wire, etc.), was performed by tradesmen.
 
b.


Objective of the Structural Testing The instrumentation and subsequent structural integrity test was performed to accomplish the following:
._
Measure and record the structural response of the primary
_
*
-_
containment under design pressure loading to 50 psig.
.. _ _ - -.
 
_ _,.
Verify that the measured response fell within the predicted
_
*
__
design limitations and tolerances.
 
l
 
i i
l
 
y
.
.
.
.
.-
i'
. ENCLOSURE 1 l
SUMMARY a
.
.
Demonstrate that the structural integrity of the primary
1.
*
containment structure is maintained under the 1.15 times design internal pressure load.


-c.
Agree Disagree
;
Violation No.


Pretest and Post Test Examination Prior to and following the structural integrity test, accessible portions of the exterior structure were surveyed for cracks.
With Licensee With Licensee
'
la - (Butt splices)
X*


Accessible portions of the liner were surveyed to detect excessive deformation.
lb - (Structural steel bolted X
connections)
'
.
;
Ic - (CEA)
X 2 - (W radiographs)
X i


Observations were made from all accessible walkways, floors, roofs, and available scaffolding.
-_(Tanks / heat exchangers)
X
- (Vendor radiographs)
X*


In addition to the over -
]
all inspection, ten areas were chosen for detailed crack monitoring during the test.
- (Component fasteners)
 
X 3a - (Electrical mounting X
Each area measured approximately 49 square feet.
deficicacies)
 
3b - (Electrical separation)
Crack widths observed prior to, during, and after the pressurization were measured using 6X comparators.
X
 
.
The crack widths were recorded only if they exceeded 0.01 inch.
3c - (MOV wiring)
 
X i
d.
Total
 
Displacements Gross deformation measurements were obtained at the following nominal levels:
0, 10, 20, 30, 40, 50, 57.5, 50, 40, 30, 20, 10, and-0 psig.
 
At each pressure increment, pressurization /
depressurization was halted for one hour before data was obtained.
 
These data were immediately reduced and printed out.
 
The printout was reviewed by the attending Sargent & Lundy engineer prior to continuing to pressurize'or depressurize to the next increment.
 
At each specified pressure level, a series of deflection measurements were made at selected locations as outlined below:
Radial displacements of the cylinder on four azimuths at four
*
elevations between the base slab and dome springline and at dome to cylinder transitions.
 
Radial displacements of the containment wall adjacent to the
*
equipment hatch at 12 points,'four equally spaced on each of three concentric circles, and the change in diameter of the equipment hatch in the horizontal and vertical directions.
 
i Vertical displacements of the cylinder at the top relative to
*
the base at four azimuths.
* Vertical deflections of the dome of the containment near the l
apex and at two other locations between the apex and the
;
springline on one azimuth.


3
!
*Not an example of a violation
!
!
Change in diameter of the equipment hatch barrel.
i
* Meters designated D1 through D8 measured the change in radius i
,
between the inner face of the cylinder and. interior reference structures.


Meters designated D9 through D14 recorded the change j
.
i
i
]
!
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.
l
,
;
,


3 i
- - - -
*
- - - - -.
-..
- - - - - -
. - - - - - - -
- - -. - - -.
. - -
- -.
- -
- -
-
-
.. -
- -
 
- -
.
.
-
. - -
-
. -.
.
.
in diameter of the cylinder above the operating floor.
- -.
. ~ -.
-.
-. - -.
.. - - - -
- - - - -
- -
- -


Meters 021 through 032 measured the change in radius cf the cylinder around tha equipment hatch. Meters V1 through V4 recorded the change in elevation between the operating floor and the dome-wall springline.
_ _.
_.


Meters V9, V10, and V11 measured the change in distance between points on the dome and the operating floor.
.. _
.
__ _ __ _
:
,
t
.
l ENCLOSURE 2
,
VIOLATION Ic As stated in the NRC CAT inspection report, the CEA qualification report and site inspection procedures specify an embedded depth (Le) to be measured from the concrete surface to the bottom of the expansion ring. However, in the installed condition Le cannot be physically measured or derived indirectly.


Meters T1 and T2 recorded the change in diameter of the equipment hatch barrel.
!
Only if it is assumed that the expansion ring remains stationary as the back
!
of the anchor does not move all the way to the ring or the expansion ring also moves towards the concrete surface as the bolt is tightened and the anchor is measured (below the concrete surface) to be a distance of exactly the minimum embedded depth, the actual Le (measured to the expansion ring) would be less than the minimum embedded depth specified by the qualification report.


Predicted displacement values are documented in Table 3.8-6 of the FSAR Amendment 44.
i Tables 1 and 2 summarize some instances in which it appears that anchor bolts


In general, the measured deflections are lower than the predicted values. At two locations, the measured values exceeded predicted values.
may not meet the embedded depth specified by the qualification report. Tables 1
.l and 2 reflect a (minus) 1/16 inch installation tolerance and physical measure-


The predicted response of D2 was 0.08 inches, and the measured response was 0.082 inch which is 2.5 percent greater. Meters D2 through 04 were all located at elevation 384 feet 6 inches.
ments of expansion anchor bolts made by the NRC CAT. The maximum deviation from i
the specified embedded depth (including installation tolerances) is 11/16 inch.


Allowing for some rounding out at this elevation, they average 0.074 inch which is 7.5 percent less than the predicted value.
We still consider that the anchor bolt qualification requirements have not been adequately translated into appropriate installation and inspection procedures.


The predicted response of D6 was 0.20 inches, and the measured response was 0.229 inches, which is 14.5 percent greater.
;
i l


The average of all measurements at this elevation (gages D5 through 08) is 0.178 inches which is 11 percent less than the predicted value.
4


Following depressurization, deflection recovery was monitored for 24 hours.
4 a
e
,
)
;
.- - - - -,,
_.. - - _.... ~ - ---
,
,
.-. -, - -
-
,en.,
,,..,,., -,,
-
,, ___
.
.
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.
Table 1*
Maximum Maximum Calculated Difference CEA CEA Between Projections Length Wedge and Hanger CEA UT CEA Above Below Bottom Minimum No.


In a majority of cases, the recovery exceeded 90 percent of maximum deflection.
Diameter Length Concrete Concrete of CEA Installed Le Traveler 8601 1/4 3 1/4 2 1/4


In all but seven cases, the recovery within 24 hours after depressurization was 80 percent or more.
3/8 5/8 WS-2 3/8


Recovery at D5 was 79.5 percent.
1 7/8 3 1/8 1/2 2 5/8 WS-35 3/8


At 013 and D14, recovery was 70.1 and 59.0 percent respectively.
1 5/8 3 3/8 1/2 2 7/8 WS-50 3/8


At V1 through V4, the average recovery was 69.8 percent with only slight variation between gages.
2 1/4 2 3/4 1/2 2 1/4 2CV130345 3/4


The average recovery of all gages was 87.0 percent.
3 9/16 6 7/16 3/4 5 11/16 2FW93E010X 5/8 8 1/2 3 1/2


e.
5/8 4 3/8
*All dimensions in inches TABLE 2*
Adjusted Difference in Qualifica. Le for 1/16" Minimum Installed and Hanger CEA Report Installation Installed Le Qual. Report No.


Cracking Ten areas were selected for detailed crack inspections.
Diameter Le Tolerance From Table 1 Adjusted Le Traveler 8601 1/4 3/4 11/16 5/8 1/16 WS-2 3/8


The inspection areas had one foot square reference grid markings on the surface. The pretest crack inspection revealed only minor shrinkage cracks.
2 15/16 2 5/8 5/16 WS-35 3/8


All cracks observed were less than 0.01 inches in width and 6 inches in length with the exception of the equipment hatch area at elevation 437 feet and azimuth 72 degrees which had one crack with 0.01 inch width.
2 15/16 2 7/8 1/16
,
WS-50 3/8


During pressurization, few new cracks were observed.
2 15/16 2 1/4 11/16 2CV130345 3/4


Most old cracks remained essentially unchanged during the test.
5 15/16 5 11/16 1/4 2FW93E010X 5/8


There were some extensions and expansions of existing cracks.
4 15/16 4 3/8 9/16
*All dimensions in inches


All cracks measured were between 0.002 inches and 0.01 inches in width.
_.


f.
_
 
o
Conclusions In most cases the measured deflections were less than predicted.
.
 
VIOLATION 2 (Westinghouse Radiographs / Construction Weakness 1)
The average measured recovery at all locations was 87.0 percent.
For the Unit 2 component cooling surge tank it has not been addressed how the ASME Code requirements are being met without the availability and existence of the radiographic film. An alternative is to reduce the joint efficiency from 100% to 75% which will then meet the ASME Section III rules. The use and acceptance of a lower joint efficiency would then have to be evaluated. The current response cannot be accepted because the ASME Code requirements have not been met.


The majority of the measurements appear to have essentially a linear response and good recovery.
VIOLATION 3b The Supplemental Safety Evaluation Report (SSER) transmitted to CECO in a memorandum from V. S. Noonan, NRR dated February 25, 1986, the NRC staff's evaluation of Byron /Braidwood electrical separation criteria.


This indicates that yielding of materials in the containment structure is not a concern. With the exception of D2 and 06, all displacements were less than predicted.
In summary the staff's conclusions were:
(1) Between safety-related and nonsafety-related raceway, the separation distances of 12" vertical and 3" horizontal is adequate, and (2) Between safety-related cables in free-air and nonsafety-related raceway and for the case of nonsafety related cable in free-air and safety-related raceway, contact is acceptable.


.
The NRC approved criteria should be the basis of QC inspection criteria.
s
.
The pretest cracks observed in the designated-areas were not significant in size or length, and are probably the result of thermal strains and/or drying shrinkage. These are believed to be surface cracks. No appreciable change in crack width was observed during pressurization to 57.5 psig. There was no sign of any structural distress at any time during or after the test. No visible signs of permanent damage to either the concrete or steel liner were detected.


The overall structural response demonstrated that the containment structure performed satisfactorily and maintained the desired structural integrity under design loading conditions.
However, your response maintains tnat the only raceway separation criteria necessary is 1" between raceway. The QC criteria remains in conflict with current FSAR and SSER statements.


3.
Installations which do not meet the separation criteria as defined in the SSER require identification and evaluation.


Exit Interview The inspector met with licensee representatives (denoted under Persons Contacted) on August 20, 1986. The inspector summarized the purpose and findings as reported herein. The inspector also discussed the likely informational content of the inspector's report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents / processes as proprietary.
We maintain that you had not established and still do not have inspection procedures which verify conformance to the FSAR/SSER criteria for electrical raceway separation. This violation remains valid.


,
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Revision as of 12:49, 10 December 2024

Ack Receipt of 860124 & 0409 Ltrs Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-455/86-27. NRC Concurs w/7 of 10 Items Cited in 860124 Response.Encl 2 Addresses Remaining Items
ML20197H382
Person / Time
Site: Byron Constellation icon.png
Issue date: 05/12/1986
From: Warnick R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To: Reed C
COMMONWEALTH EDISON CO.
References
NUDOCS 8605190157
Download: ML20197H382 (6)
v • d • e


Text

od 7?? 6

.

.

MAY 121986

/

[3 I

Docket No. 50-454" Commonwealth Edison Company ATTN:

Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:

Thank you for your letters dated January 24, and April 9,1986, informing us of the steps you have taken in response to the results of the NRC Construction Assessment Team (CAT) inspection forwarded in Inspection Report No. 50-455/85027 by letter dated November 13, 1985 and the Notice of Violation forwarded by our letter dated December 12, 1985.

Your January 24, 1986 response was forwarded to the Office of Inspection and Enforcement who lead the CAT inspection.

Your response has been reviewed and we are in agreement on seven of the ten individual items used as examples of violations. The specific items and our findings are summarized in Enclosure 1.

Our comments concerning the remaining violations are provided in Enclosure 2.

We will examine these matters during subsequent inspections.

Your cooperation with us is appreciated.

-

Sincerely,

"Orighrd cit ed hi 2. F. tia nic M R. F. Warnick, Chief Reactor Projects Branch 1 Enclosures: As stated See Attached Distribution RIII RIII RIII R

RIII RIII RI LQ k

< A ONd

.j W A Da,677//

M ~g RIII j

h F rfey Warnick G n

Muff tt rjieIson Pea ison P sePiello Learch/bls

9

"

f czIb 5[3/Bb 5/2/ 4 flt h

/

\\

IM"I8$8E SESMss

'

ip

o

-

.

Commonwealth Edison

MAy { g jggg Distribution cc w/ enclosure:

D. L. Farrar, Director of Nuclear Licensing V. I. Schlosser, Project Manager Gunner Sorensen, Site Project Superintendent R. E. Querio, Plant Manager DCS/RSB (RIDS)

Licensing Fee Management Branch Resident Inspector, RIII Byron Resident Inspector, RIII Braidwood Phyllis Dunton, Attorney General's Office, Environmental Control Division D. W. Cassel, Jr., Esq.

Diane Chavez, DAARE/ SAFE Steve Lewis, ELD L. Olshan, NRR LPM H. S. faylor, Quality Assurance Division

.

._

_

-_

.. _ _ - -.

_ _,.

_

__

.

.

i'

. ENCLOSURE 1 l

SUMMARY a

.

1.

Agree Disagree

Violation No.

With Licensee With Licensee

'

la - (Butt splices)

X*

lb - (Structural steel bolted X

connections)

'

.

Ic - (CEA)

X 2 - (W radiographs)

X i

-_(Tanks / heat exchangers)

X

- (Vendor radiographs)

X*

]

- (Component fasteners)

X 3a - (Electrical mounting X

deficicacies)

3b - (Electrical separation)

X

.

3c - (MOV wiring)

X i

Total

3

!

  • Not an example of a violation

!

i

,

i

]

!

.

,

,

- - - -

- - - - -.

- - - - - -

. - - - - - - -

- - -. - - -.

. - -

- -.

- -

- -

-

- -

- -

.

-

. - -

-

. -.

.

- -.

. ~ -.

-.

-. - -.

.. - - - -

- - - - -

- -

- -

_ _.

_.

.. _

.

__ _ __ _

,

t

.

l ENCLOSURE 2

,

VIOLATION Ic As stated in the NRC CAT inspection report, the CEA qualification report and site inspection procedures specify an embedded depth (Le) to be measured from the concrete surface to the bottom of the expansion ring. However, in the installed condition Le cannot be physically measured or derived indirectly.

!

Only if it is assumed that the expansion ring remains stationary as the back

!

of the anchor does not move all the way to the ring or the expansion ring also moves towards the concrete surface as the bolt is tightened and the anchor is measured (below the concrete surface) to be a distance of exactly the minimum embedded depth, the actual Le (measured to the expansion ring) would be less than the minimum embedded depth specified by the qualification report.

i Tables 1 and 2 summarize some instances in which it appears that anchor bolts

may not meet the embedded depth specified by the qualification report. Tables 1

.l and 2 reflect a (minus) 1/16 inch installation tolerance and physical measure-

ments of expansion anchor bolts made by the NRC CAT. The maximum deviation from i

the specified embedded depth (including installation tolerances) is 11/16 inch.

We still consider that the anchor bolt qualification requirements have not been adequately translated into appropriate installation and inspection procedures.

i l

4

4 a

e

,

)

.- - - - -,,

_.. - - _.... ~ - ---

,

,

.-. -, - -

-

,en.,

,,..,,., -,,

-

,, ___

.

.

.

.

Table 1*

Maximum Maximum Calculated Difference CEA CEA Between Projections Length Wedge and Hanger CEA UT CEA Above Below Bottom Minimum No.

Diameter Length Concrete Concrete of CEA Installed Le Traveler 8601 1/4 3 1/4 2 1/4

3/8 5/8 WS-2 3/8

1 7/8 3 1/8 1/2 2 5/8 WS-35 3/8

1 5/8 3 3/8 1/2 2 7/8 WS-50 3/8

2 1/4 2 3/4 1/2 2 1/4 2CV130345 3/4

3 9/16 6 7/16 3/4 5 11/16 2FW93E010X 5/8 8 1/2 3 1/2

5/8 4 3/8

  • All dimensions in inches TABLE 2*

Adjusted Difference in Qualifica. Le for 1/16" Minimum Installed and Hanger CEA Report Installation Installed Le Qual. Report No.

Diameter Le Tolerance From Table 1 Adjusted Le Traveler 8601 1/4 3/4 11/16 5/8 1/16 WS-2 3/8

2 15/16 2 5/8 5/16 WS-35 3/8

2 15/16 2 7/8 1/16

,

WS-50 3/8

2 15/16 2 1/4 11/16 2CV130345 3/4

5 15/16 5 11/16 1/4 2FW93E010X 5/8

4 15/16 4 3/8 9/16

  • All dimensions in inches

_.

_

o

.

VIOLATION 2 (Westinghouse Radiographs / Construction Weakness 1)

For the Unit 2 component cooling surge tank it has not been addressed how the ASME Code requirements are being met without the availability and existence of the radiographic film. An alternative is to reduce the joint efficiency from 100% to 75% which will then meet the ASME Section III rules. The use and acceptance of a lower joint efficiency would then have to be evaluated. The current response cannot be accepted because the ASME Code requirements have not been met.

VIOLATION 3b The Supplemental Safety Evaluation Report (SSER) transmitted to CECO in a memorandum from V. S. Noonan, NRR dated February 25, 1986, the NRC staff's evaluation of Byron /Braidwood electrical separation criteria.

In summary the staff's conclusions were:

(1) Between safety-related and nonsafety-related raceway, the separation distances of 12" vertical and 3" horizontal is adequate, and (2) Between safety-related cables in free-air and nonsafety-related raceway and for the case of nonsafety related cable in free-air and safety-related raceway, contact is acceptable.

The NRC approved criteria should be the basis of QC inspection criteria.

However, your response maintains tnat the only raceway separation criteria necessary is 1" between raceway. The QC criteria remains in conflict with current FSAR and SSER statements.

Installations which do not meet the separation criteria as defined in the SSER require identification and evaluation.

We maintain that you had not established and still do not have inspection procedures which verify conformance to the FSAR/SSER criteria for electrical raceway separation. This violation remains valid.

l

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